ML20203H405
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| ML20203H405 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 04/04/1986 |
| From: | NRC |
| To: | |
| Shared Package | |
| ML20203H364 | List: |
| References | |
| FOIA-86-36 NUDOCS 8604290489 | |
| Download: ML20203H405 (227) | |
See also: IR 05000445/1985013
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CPRT OBJECTIVITY QUESTIONNAIRE
1. Kane
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2. Position on CPRT
3. Name of current employer
4. Data this questionnaire is completed
. 5. Do you hold any stock or other securities of Texas Utilities
Company? .
6. Have you ever been a director, officer, or employee of Texas
Utilities Company or any,of its subsidiaries *?
7. Have you ever entered. into a contract, 'whether oral or written,
with Texas Utilities Company or any of its subsidiaries or with
another person; fira, or corporation who has acted as a contractor
for the Comanche Peak project other than a centract relating to
your involvement in the current CPkT program?
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8. Would the answer to any of the above questions be "yes" if answered
- by any member of your immediata fanily (father, mother, spouse,
son, or daughter)?
9. Have you been promised any additional compensation or reward or
- anything of value by anyone, contingent upon the position you take
on any issue being considered by you in the CPRT program?
10. Do you know of any reason, whether inquired about in this
questionnaire or not, which would affect your ability to be
completely objective in performing any of the tasks assigned to you
under the Comanche Peak CPRT program?_
! ! 11. Would the answer to any of the above questions be "yes" if answered
by your current employer?_
12. If the answer to any one or more of the above questions was "yes"
then please fully explain each such "yes" answer, by number, on the
reverse side of this questionnaire. Attach additional sheets to
[ provide further information, if necessary. l I-
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Signature
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* Subsidiaries of Texas Utilities Company are Texas Utilities
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Electric Company, which has four divisions: Texas Power & Light
!
Company Texas Electric Service Company, Dallas Power & Light
Company, and Texas Utilities Cenerating Company; Texas Utilities
Mining Company; Texas Utilities Fuel Company; Basic Resources Inc.;
and Chaco Energy Company.
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B604290489 860409
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1. Allegation Categoryi ' Mechanical and Piping 11, Pipe Installation.
2. Allegation Number: AP-4,'AP-9 and AP-13
3. Characterization: It is alleged that:
AP-4
Two or three 100-ton jacks were used to " cold spring" the reactor coolant
syspemlines.
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AP-9
A stainless steel piping spool was heated during installation to achieve
proper fitup. The alleger indicated that the spool piece was probably
nonsafety-related and probably located in the Containment Spray System.
;AP-D
A 32-inch Main Steam line was forced into position by the polar crane and
3-ton come-alongs.
4. Assessment of Safety Significance:
AP-4
In assessing allegation AP-4, the NRC Technical Review Team (TRT) reviewed
the configuration of and the installation requirements for the Reactor
Coolant System (RCS) piping to assess whether " cold springing"* was likely
to be used during installation of the RCS piping. The TRT concluded that
the configuration and installation procedures for the RCS piping were such
as to preclude the need for springing of the piping. Brown & Root (B&R)
Drawing BRP-RC-1-520-001, " Reactor Coolant Loops, Layout and Details,"
Rev.10, July 1984, showed that each loop of the RCS piping consists
essentially of two straight pieces of pipe, the hot and cold legs, which
connect the steam generator and reactor coolant pump, respectively, to the
reactor pressure vessel and a U-shaped crossover leg which connects the
steam generator to the reactor coolant pump. Additionally Westinghouse
Specification PSPAS-01, " Assembly, Fitting and Welding Sequence for Primary
Loop Piping," Rev. O, May 26, 1977, indicated that both the hot and cold
legs were supplied as single lengths of pipe while the crossover leg was
supplied in two sections and that c ch leg was installed by fabricating
,
"The term cold springing, as used in Se' tionc 4.13 of G&H Specification 2323-MS-100,
Rev. O, March 1, 1976 through Rev. 8, July 5, 1984, refers to a controlled process
to reduce the effects of stresses due to thermal expansion in piping systems
and is consistent with the use of the term in subparagraph NB-3672.8 of the
ASME B&PVC and paragraph 119.9 of the ANSI B31.1 Power Piping standard. The
term springing as used in Section 4.7.6 of G&H Specification 2323-MS-100,
Rev. O through Rev. 8 refers to the uncontrolled practice of mechanically
deflecting piping at closure joints in piping systems to correct construction
misalignments.
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all welds in the leg simultaneously. Consequently, there was no oppor-
tunity for misalignment and no need for springing of the piping. f
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Additionally, figures 5.4-13 and 5.4-14 of the CPSES Final Safety Analysis
Report (FSAR) showed that the steam generator and pump in each loop were
supported vertically by pin-ended columns but that lateral support of the
steam generator and pump were provided by the attached RCS piping and
and auxiliary structures. Due to their method of support and the antici-
pated shortening of the attached piping due to weld shrinkage, measures
were taken to monitor and maintain the vertical positioning of the steam &
l generator and pump casing during hot and cold leg installation, respec- ,
tively (the pump internals and motor were not installed during RCS piping i
installation). TRT reviews of data packages for the welds in the Unit 1 ', ' ~
RCS piping indicated that vertical positioning of the steam generator was -l
maintained by the use of jacks. Jacks were not employed to maintain ver- ;
tical positioning of the less massive and more compact pump casing. {
Based on the results of the above described reviews, the TRT believed that '
the alleger mistakenly identified jacking operations for maintaining ver- 1
tical positioning of the steam generators during installation of the RCS
hot leg piping as springing.
The TRT also reviewed applicable specifications and construction proce-
dures and inspection instructions to evaluate the requirements for and the
.
construction practices related to springing and cold springing. The re- .
view determined that the requirements of all revisions of G&H Specifica- '
tion 2323-MS-100 " Piping Erection Specifications," were adequate to assure
that springing and cold springing were controlled such that the related
design requirements of the ASME B&PV Code and ANSI B31.1 Standard would be ,
satisfied.
The TRT review of B&R Procedure CP-CPM-6.9E, " Pipe Fabrication and Instal-
lation," determined that: (1) cold springing was not addressed in all is-
sues of the procedure and (2) springing requirements were addressed
adequately only in Rev. 1, May 23, 1980 through Rev. 7, May 14, 1984. A
similar TRT review of B&R Instruction QI-QAP-11.1-26, "ASME Pipe Fabrica-
tion and Installation Inspections and Requirements Prior to System / Subsystem
N-S Certification," determined that: (1) cold springing was not addressed
in all issues of the instruction and (2) springing requirements were ad-
dressed adequately only in Rev. 10, February 2, 1983 through Rev. 15,
April 18, 1984. The TRT reviews also determined that: (1) both B&R Proce- 3
dure CP-CPM-6.90 and B&R Instruction QI-QAP-11.1.26 used the term " cold
springing" to describe springing-related activities, and hence, (2) that i
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the use of the term " cold springing" in both B&R documents was inconsis- / ,
tent with the use of the term in the G&H Specification 2323-MS-100, the ~j
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ASME Code, Section III and the ANSI B31.1 standard.
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With respect to the failure of both B&R Procedure CP-CPM-6.90 and B&R In- i
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struction QI-QAP-11.1-26 to address cold springing requirements the TRT T
was informed by TUEC and/or B&R Engineering, construction. and QA/QC per-
sonnel that no piping systems had been cold sprung. Subsequent TRT dis- }g
cussions with G&H piping design engineering personnel confirmed that there
was no G&H requirement for cold springing of ,niping systems.
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However, the TRT found that the failure of B&R Procedure CP-CPM-6.90 and
B&R Instruction QI-QAP-11.1-26 to adequately address springing require-
ments of G&H Specification 2323-MS-200 prior to May 1980 and February 1983,
respectively, and the use of the term " cold springing" in a manner which
was inconsistent with the use of the term in G&H Specification 2323-MS-100
were in noncompliance with the requirements of B&R Procedure CP-CPM-6.1,
" Preparation & Control of Construction Procedures & Instructions," and B&R
Instruction CP-QAP-6.1, "Preparetion of QA Procedures and Instructions,"
respectively. B&R Procedure CP-CPM-6.1 required that the construction
procedures be reviewed for adequacy. Paragraph 3.4 of B&R Instruction
CP-QAP-6:1 required that QA Procedures and Instructions shall be reviewed
for technical content, Code and Regulatory requirements, grammatical cor-
rectness, clarity and for any conflict or redundancy with other documents.
In view of inadequate treatment of springing by B&R Procedure CP-CPM-6.90
and B&R Instruction QI-QAP-11.1-26, the TRT interviewed TUEC mechanical
engineering and QA/QC and B&R construction and QA/QC personnel to deter-
mine the actual practice for springing at CPSES. Although the B&R con-
struction and QA/QC personnel said that springing was prohibited, a TUEC
mechanical engineering (ME) supervisor said that authorization for spring-
ing was permitted in all piping systems on a case-by case basis. Criteria
for authorization were based on guidelines in an Appendix D to an uniden-
tified Bechtel Corporation Specification 10466-M-204. (The TUEC ME super-
visor had a copy of the Appendix only.) The TRT observed that Appendix 0
to the Bechtel specification stated that the springing guidelines provided
were intended to limit springing strdsses at nozzles to 1000 psi in nuclear
and non-nuclear carbon steel and stainless steel piping in the range of
Sch. 105 to Sch. 160. Although the use of guidelines may be acceptable
for the installation of Comanche Peak Steam Electric Station (CPSES) piping
systems, the TRT found that use of the Bechtel specification for installa-
tion of piping systems at CPSES was unauthorized and undocumented. This
was contrary to the requirements of Paragraph 33 of TUEC procedure CP-EP-4.0,
" Design Control," Rev. 3, July 11, 1982, which required that design inputs
on which final design were based were to be identified, documented and
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approved on a timely basis. The inputs were also required to be detailed
enough to provide a consistent basis for making design decisions, accom- .
i plishing design verification and evaluating design changes.
During subsequent interviews with the B&R QA personnel the TRT was in-
formed that springing in piping systems at closure joints was limited to
cases where gaps at the joint could be closed by hand. (The TRT notes
the discrepancy in the B&R QA/QC personnel's response.) During similar
subsequent interviews with B&R construction personnel, a general superin-
tendent claimed to have no knowledge of uncontrolled springing in piping
systems.
The TRT found that it was unreasonable to expect piping systems
to be installed without some springing.
AP-9
During its assessment of allegation AP-9, the TRT attempted to cuntact the
alleger on August 1, 1984, to obtain clarification and specific informa-
tion regarding the allegation, The alleger declined to meet with the TRT.
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The TRT also found that the allegation had been investigated by the NRC
Office of Investigations (01). The results of the NRC 01 investigation,
which included an interview with the alleger, were documented in 01 Report
50-445/84-006, March 7, 1984.
The TRT review of the 01 interview with the alleger found that the alleger
had implied that: (1) the alleged heating process was permitted to achieve
proper fitup in " normal" piping but was not permitted in stainless steel
piping and (2) during the alleged heating process heat was applied locally
by a " rosebud" (gas torch) at welds in the piping. .
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The TRT found that the' Containment Spray System was a safety-related, stain- 3
less steel piping system. Table 6.2.2.1 and Table 6.2.2.4 of the CPSES i
Final Safety Analysis Report (FSAR) indicated that the Containment Spray
System is a nuclear-safety related system and that Containment Spray Sys-
tem piping material was to be ASME B&PVC SA-312 or SA-358 stainless steel 4
piping. Consequently, due to the lack of identification of the stainless i
steel piping spool involved in the alleged incident, the complexity of the I
safety-related, stainless steel Containment Spray System Piping, and the U
large number and complexity of other safety-related and nonsafety-related
stainless steel piping systems, TRT investigations could not substantiate
the allegation. I..
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Other TRT investigations found that localized heating in some carbon steel s
piping was permitted but not permitted in stainless steel piping. Para- A.'
graph 3.12 of B&R Procedure CP-CPM 6.9E, " Pipe Fabrication and Installation," '
Rev. 1, July 23, 1980 through Rev. 7, May 14, 1984 required engineering -
approval for localized heating to obtain proper alignment during piping
installation. Paragraph 3.21.10 of B&R Procedure CP-CPM-6.9D, " Welding
and Related Processes" Rev. 1, September 23, 1980 through Rev. 6,
January 12, 1984 specified that localized heating for minor adjustments in
alignment were permitted on a case-by-case basis as approved by the Owner
but prohibited the use of such heating in stainless steel material or
carbon steel piping requiring Charpy impact values.
AP-13
During its investigation of allegation AP-13, the TRT found that the al-
leged incident was discussed in affidavits by the alleger submitted by
intervenor CASE to the CPSES Atomic Safety and Licensing Board (ASLB) or.
February 3, 1983 and November 28, 1983. The extent to which the alleger's
allegations were open in the CPSES operating license (0L) proceeding was
clarified by the ASLB Memorandum and Order (Clarification of Open Issues),
March 15, 1983. ;
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The TRT also found that the allegation had been investigated by NRC Office i
of Investigations (OI) and RIV. The OI investigation was documented in
OI Report A4-83-005, May 20, 1983; OI Memorandum, " Assistance to In- 3
spection Report No. A4-83-005," June 2, 1983; 01 Supplemental Report e
Q4-84-007, February 9, 1984. The 01 investigation included an interview
with the alleger on April 14, 1983 and interviews with 27 other persons
relating to the allegations. The NRC RIV investigation was documented in .
IR 50-445/83-27, September 29, 1983 and had concluded that:
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Although B&R personnel named by [the alleger] contradicted his '
allegation, the NRC inspector conducted an independent review
of the onsite documented records regarding this matter. !
It was observed by the NRC inspector that the specific 32-inch
stcam line mentioned by [the alleger) is, Loop 1, Line number
MS-1-R8-001-1302-2,
and the reactor building polar crane was
utilized in a vertical
this permanent piping.
lift to assist repositioning a section of
engineering
e
record of theThe licensee
specific linehas maintained a documented
movement. The NRC in-
sp' ctor noted that the movement of the line was necessary in
order that a large section of temporary piping (attached to the
steam generator feedwater nozzle and previously used for water
flushing) be removed and to relocate the permanent section'of
the main steam line that had " sagged" due to the weight of the
temporarily installed flushing pipe.
The record folder contains
meeting notes (memorandum) which reflect discussions with Westing-
house (NSS Supplier) and the cognizant A/E representatives prior
to the work activity, in addition to establishing engineering
limitations and acceptability. The line was moved on January 16,
1982 under the supervision of the field mechanical engineering
group, and was witnessed by an engineering representative who
observed the installation and use of the dynamometer (to regis- .
ter crane lifting loads) throughout the operation. The lift
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connections
file.
and applied forces were recorded and retained in the
The Itfting points were consistent with the hanger loca-
tions to simulate the permanent support system. The as-built
configuration
the line confirmed.was analyzed for stress and the acceptability of
In addition, the recent completion of the
" Reactor Hot Functional Test" did not reveal any undue stress
conditions. This allegation cannot be substantiated.
No violations or deviations were identified in this area of the
inspection.
The TRT investigation of the allegation was based on the numerous claims
by the alleger made during the OI interview on April 14, 1983, additional '
information obtained by the TRT during its September 19, 1984 interview
with the alleger, and other information contained in the alleger's
February 3,1983 and November 28, 1983 affidavits.
Based on its reviews
of these interviews and affidavits, the TRT found that the alleger made
the
line:following claims regarding the forced movement of the main steam (MS)
1.
The incident occurred in 1982, prior to the summer.
2.
The piping involved in the incident consisted of a portion of the M5
( line and an attached temporary flushing line (installation of the MS i
line was incomplete). During the incident, the installed portion of
, l the MS line was attached at one end at the containment penetration l
- for the line and attached at the other end to the temporary flushing ]
l line, which in turn was disconnected from a nozzle at the lower end of
the steam generator.
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3. During disconnection of the temporary flushing line from the steam
generator, the line vibrated about 14 inches with a loud echoing' 3
noise audible throughout containment.
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4. After disconnection of the temporary flushing line from the steam
generator, it was discovered that the installed portion of the MS
line had been placed incorrectly 6 inches vertically and 4 inches
horizontally from its design location.
5. During the incident, the MS line was forced into its proper design
location by the polar crane and 3- to 5-ton come-alongs. <
6. The force exerted by the polar crane was applied at "the expaasion
chamber." The force was monitored by "a big round genge [that]
look[ed] like [a] big clock."
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7. The magnitude of the applied force was estimated to be between 40
tons and 85 tons (The alleger stated: "I cannot remember the exact '
tonage").
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8. Four to six permanent pipe supports which had been installed prior to
the alleged incident were redesigned to maintain the MS line in its
forced position.
9. The incident was not supervised by engineering personnel.
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The alleger was concerned that " tension" induced in the MS line as a re-
sult of movement during the alleged incident was still present in the .
line.
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Based on TRT reviews of statements of the five persons interviewed by 01
between May 1982 and January 1984 who claimed to have knowledge of the
alleged incident, TRT interviews with B&R construction and QA/QC and TUEC i
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engineering and QA/QC personnel, and TRT reviews of the construction records
for the MS and Feedwater piping system, the TRT found the following
with respect to the alleger's claims:
1. The alleged incident was related to the Unit 1, Loop 1 MS line, Line
MS-1-RB-001-1303-2, and occurred in January 1982. The date of entry ,
for the Unit 1, Loop 1 reading of 31,250 (1bs) on M&TE Issue Record 1
for MTE Serial No. MTE-357 was January 16, 1982. The January 1982 {
date was also confirmed by two persons interviewed by 01. Additionally, f
component modification card (CMC) 61306 Rev. 4, March 23, 1982, which {
documented modifications to pipe support MS-1-001-007-C72K for the Unit 4
1, Loop 1 MS line contained the note: " Pipe was moved." RIV also i
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found that the alleged incident had occurred in January 1982 and was f
related to the Unit 1, Loop 1 MS line. p
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The configuration of the piping involved was as described by the al- t
2.
leger, except that the temporary flushing line had not been discon-
ected from a steam generator nozzle but was disconnected from a field
weld (FW) in a feedwater line to which it had been attached.
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TRT reviews determined that the Loop I temporary flushing line as '
shown on B&R Drawing FSM-00165, Rev.1 (undated) was intended to be -
attached at field weld (FW) locations FW-5 and FW-4 in Loop 1 MS Line -
32"-MS-1-01-1303-2 and Loop 1 Feedwater Line 18"-FW-1-19-1302-2,
respectively. Additional TRT reviews of documentation packages for the E
field welds verified that the temporary flushing line was installed as
intended by temporary field welds FW-T5 and FW-T4 to the Loop 1 MS and 6 -
Feedwater lines, respectively, on February 5,1980 and June 19, 1980, g
respectively.
The TRT was informed by a TUEC mechanical field engineer that, prior
to the incident, the temporary flushing line was cut at approximately
30 ft, below the 899-ft, 91-inch
s elevation and that the piping involved ;
in the alleged incident consisted of the portion of the MS line between j
the containment penetration and FW-5 and the attached portion of the '
temporary flushing line above the cut. !
Because of the expanded scope of its assessment, the TRT finds it
unlikely that the temporary flushing line was attached to the steam
generator, since the line was intended to be a steam generator bypass
line. If the temporary flushing line were attached to the steam
generator, it could not function as a bypass line.
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3. The alleger's claim that the temporary flushing line vibrated about
14 inches when the line was disconnected coulc not be substantiated.
Two persons interviewed by 01 and/or the TRT refuted the claim.
4. Prior to the incident, B&R discovered that the installed portion of
the MS line had deviated from its initial installation position prob-
ably due to settlements at temporary supports (for the installed pip-
ing) during construction and/or flushing. B&R could not provide
surveying documentation to substantiate the deviation in location;
however, four persons interviewed by OI and/or the TRT supported this
TRT determination. Furthermore, since B&R and TUEC personnel had
informed the TRT that temporary support of the installed piping had
been provided by wooden cribbing and " lashing" (retal cables) the TRT
cid not consicer settlements at the support, unreasonable. This TRT
finding was in agreement with the related RIV finding but was contrary I
to the alleger's claim that the MS line had been incorrectly installed
and that operations during the alleged incident were necessary to force
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the MS line into its correct location.
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The TRT could not determine exactly how much settlement had occurred
at the temporary supports.
Vertical movements of the MS line (at >
unspecified locations) during the alleged incident were reported by
persons interviewed by 01 and/or the TRT to be between 1-1/2 inches
to 6 inches and between 4 inches and 5 inches for horizontal
mvvement s ,
l i l 5. The alleger's claim that the polar crane and 3- to 5-ton come-along l
were used to move the MS line was confirmed in whole or in part by all j
but one of the persons interviewed by 01 and/or the TRT who claimed
to have knowledge of the incident. !
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A TUEC mechanical field engineer, who claimed to have been involved in
the incident, stated that initial attempts to use the main polar crane
were unsuccessful and the 20-ton auxiliary crane was used during the ,
incident "because it could be controlled more accurately" (TUEC Of- l
fice Memorandum, " Comanche Peak Steam Electric Station," February 4, 8
1983). However during an interview with a rigger who was involved
in the alleged incident, the TRT was informed that initial attempts to
use the "small" (auxiliary) crane were unsuccessful due to uncontrolled .
slippage and that subsequently the "large" (main polar) crane was }
used to move the MS line. $
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All but one person' interviewed by 01 and/or the TRT confirmed the 3
alleged use of both the polar crane and come-alongs during the alleged Y
incident. The one person discussed the use of the polar crane only. 5.
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6. The alleger's claim that the force exerted by the polar crane was Y
applied at "the expansion chamber" was partially confirmed by two ~t
persons (including a TUEC field mechanical engineer) interviewed by i
01 and/or the TRT who stated that the force was exerted at a thermal 4
expansion loop in the MS line). The alleger's claim that the force [;
was monitored by " clock"-like instrument (actually a dynamometer) was .
substantiated. M&TE Issue Record for MTE Serial No. MTE-357 docu-
mented a dynamometer reading related to the alleged incident. RIV -
inspections also substantiated the claim regarding the use of a dyna-
mometer during the alleged incident.
During an interview with a TUEC field mechanical engineer, the TRT
was informed that during the alleged incident a chain was installed
between the legs of the expansion loop to which the polar crane
force was applied. The chain was intended to distribute the polar
crane force in the section of the installed MS line consistent with
the manner in which the section of line was to be supported in its
design configuration. The use of the chain during the alleged inci-
dent was refuted by the alleger and a rigger involved in the
incident.
7. The magnitude of the force exerted by the polar crane could not be
determined exactly. As noted previously, M&TE Issue Record Ior MTE
Serial No. MTE-357 recorded a value of 31,250 (lbs.). However, a
rigger involved in the incident stated that the force applied was
47,500 lbs. Additionally, the magnitude and points of application
of the forces applied by the alleged 3-to-5-ton come-alongs could
not be determined.
8. The alleger's claim regarding the redesign of previously installed
pipe supports was substantiated. A TRT waldown inspection of the MS
line and review of pipe support designs dete,rmined that modification
_
of- permanent pipe supports'had-been made to accommodatema-movemenbof
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the MS line in an eastern direction. Visual inspections of spring
hanger supports MS-1-001-002-C725 and MS-1-001-001-C725 and vertical
seismic (snubber) support MS-1-001-007-C72K indicated that support
details had been modified to accommodate a lateral movement of the M5
line in the eastern direction. CMC 61306 Rev. 4, March 23, 1982 (which
stated that the reason for the change was " Pipe was Moved") indicated
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Comanche Peak SSER 10 N-106
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that modifications were made to support MS-1-001-007-C72K to accommo- '
date lateral movements of the MS line in the eastern and northern
directions.
ments of the MS Modifications to acocmmodate eastern and northern move-
line were also noted at seismic supports MS-1-001-
005-C72K and MS-1-001-006-C72K. !
9.
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The alleged incident was performed under the cognizance of the TUEC
field mechanical engineering group. Seven persons interviewed by 01
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and/or the TRT disagreed with the alleger's claim that the alleged
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incident was not supervised by engineering personnel. A TUEC field
mechanical engineer stated that he was present during movement of the
' MS line "to insure that the activity was within proper engineering
limitations for the equipment" involved. This TRI finding was in
agreement with the related RIV finding.
- - The TRT reviewed a TUEC analysis performed in February 1983, 1 year after
< the incident, to assess strasses induced in the MS line during the inci-
dent.
The TRT review concluded that the analysis was inadequate since
<
only operations during restoration of the MS line to its initial position
were addressed.
The conditions during flushing, including the piping
af fected, the weight of added water, and settlement of temporary supports,
were not addressed.
During its investigation of allegation AP-13 the TRT also found that the
Unit 1, Loop 4 Main Steam Line 32"-MS-1-004-1303-2 was involved in a
similar indident which has not been evaluated by TUEC. M&TE Issue Record
for MTE Serial No. MTE-357 also documented a dynamometer reading of
17,750 lbs. on January 16, 1982 related to the Loop 4 MS line.
g The TRT
noted that the dates recorded on the M&TE Issue Form for both the Loop 1
3' and Loop 4 dynamometer readings were identical (the recorded dates were
January 16, 1982).
The TRT noted, however, that it was unreasonable that
both the Loop 1 and Loop 4 lines could have been moved on the same day.
Furthermore, during an interview with a rigger who claimed to have been
;
involved with both incidents, the TRT was informed that the incidents had
.i
not occurred on the same day.
t-
further TRT reviews determined that the use of temporary supports during
the alleged incident was in noncompliance with the requirements of Sec-
tion 7.4, " Adjustments of Hangers," and 7.5, " Temporary Hangers," of G&H
Specification 2323-MS-100, Piping Erection Specification," Rev. 5,
4 Februa y
F 26, 1979 (Rev. 5 was in effect during the alleged incident).
Section 7.5 stated that all piping was to be erected in its permanent
hangers, but where not possible, the Owner's approval was to be obtained
for the use of temporary hangers. Section 7.4.2, " Adjustment Prior to
Testing and Flushing," only addressed requirements prior to hydrostatic
testing of erected piping systems. During an interview with the TRT a TUEC
chief engineer for piping informed the TRT that Owner (/ Engineer) approval
' for the use of temporary supports during piping erection was granted to
' B&R by TUEC approval of B&R Procedure PCP-1, " Process Pipe Installation,"
Rev. O, June 7, 1977 (TUEC approval in note TUF-3388, July 28, 1977 on B&R
' BRF-6565, July 11,1977). Paragraph 4.6.1 of Procedure PCP-1 stated that
pipe spools were to be rigged into permanent or temporary supports during
erection. The TRT questioned whether Note TUF-3388 satisfies the intent
.
Comanche Peak SSER 10 N-107
U
.
. .
, . .
of the approval requirement in Section 7.5 of G&H Specification
-
2323-MS-100.
In view of the above identified noncompliance with the rcquirements of G&H
Specification 2323-MS-100 and the TRT observation that the requirements of
Sections 7.5 and 7.4.2 of this G&H Specification were unchanged in Rev. O,
March 1,1976 through Rev. 5, February 26, 1979, the TRT interviewed TUEC
engineering and B&R construction personnel to determine the practice re-
garding temporary supports. The TRT determined that there was no formal
procedure to evaluate the adequacy and temporary supports during piping
installation. A B&R general superintendent stated that the previous undoc-
umented practice of the unrestricted use of lashing was discontinued due ,
to problems with uncontrolled movements of the piping. The TRT was in- .
formed by the general superintendent that the previous practice was modi-
fied to require that the piping was to be fixed against movement at every }j
fourth support during erection. The TRT found that the above practice e
was not in compliance with the requirements of G&H Specification 2323-MS-100 l[
for temporary supports, Rev. O, March 1, 1976 through 5, February 26, 1979
that all piping be erected in its permanent supports and that Owner ap- j%
proval be obtained for the use of temporary hangers. p
i
The TRT also noted that Sections 7.4.2 and 7.4 of G&H Specification l
2323-MS-100, Rev. 6 and 7, March 15, 1982 and December 2, 1983, permitted
the use of temporary hangers without Owner's approval but required that
during flushing and hydrotesting of steam lines with demineralized water
the deadweight supports were required to be in place with all spring hang-
ers in the locked position.
5. Conclusion and Staff Positions:
AP-4
Based on its investigation the TRT determined that the allegation regard-
ing springing in RCS piping could not be substantiated. The TRT believes
that the alleger had mistakenly identified jacking activities during in-
stallation of the hot-leg piping between the reactor vessel and the steam
generator as cold springing activities. Jacking was used to maintain verti-
cal positioning of the steam generator during axial motions of the hot leg
piping due to weld shrinkage during welding.
Other TRT investigations of cold springing and springing determined that
no piping systems were intended to be or had been cold sprung but that,
contrary to the requirements of TUEC Procedure CP-EP-4.0, unauthorized and
undocumented springing of piping systems had occurred. The TRT, however,
assessed that the safety significance of the unauthorized and undocumented
springing practice may be negligible; guidelines for the unauthorized and
undocumented springing stresses to only 1000 psi at nozzles.
The TRT found that the failure of all issues of B&R Procedure CP-CPM-6.9E
and B&R Instruction Ql-QAP-11.1.26 to adequately reflect the springing and
cold springing requirements of all issues of G&H Specification 2323-MS-100
were in noncompliance with the requirements of B&R Procedure CP-CPM-6.1 ana
B&R Instruction CP-QAP-6.1, respectively.
Comanche Peak SSER 10 N-108
.
J
-
-
1
'
. .-
l
i
Additionally tso matters relating to an inconsistency in responses of B&R
QA were
nel personnel
noted.and possible misleading statements by B&R construction person-
The TRT cont. acted the alleger on August 20, 1984 and October 29, 1984 to
8 arrange
tigations.
a meeting to discuss his allegation and the results of its inves-
The alleger declined to have a meeting with the TRT.
AP-9
The TRT investigation of alleged localized heating of a stainless steel
line to obtain proper alignment of the line could not substantiate the
allegation.
Because the alleger declined to meet with the TRT, the TRT could not con-
duct a followup interview to discuss its findings with the alleger.
AP-13
9' The TRT investigation of allegation AP-13 concluded that the allegation
g was substantiated in part. The TRT found that the alleger had mistakenly ,
'
4
identified the repositioning of a Unit 1, Loop 1 Main Steam line due to
during initialofinstallation.
settlements temporary supports as the correction of alignment errors
The TRT determined that the partially in- .
stalled main steam line had been used in flushing operations and sagged ;
due
tion.to settlement of temporary supports during flushing and/or construc-
TUEC's analysis to assess stres'ses in the main steam line due to
the repositioning operations was inadequate because stresses due to the ,
full sequence of events involved in the incident were not evaluated. .
The TRT also determined that similar repositioning operations had been
performed on the Unit 1, Loop 4 main steam line. No assessment of this
second incident has been performed by TUEC. An apparent anomaly in the
recorded dates of the Loop 1 and Loop 4 MS line was noted.
TRT reviews also found that the B&R construction practice of using tem-
porary supports during piping erection was not in compliance with the G&H
l Specification 2323-MS-100 requirements.
Based on the deficiencies identified in its assessment of allegation AP-13,
the TRT concluded that it has safety significance and generic implications.
Additional information required to resolve concerns related to the alle-
gation was requested in an NRC letter to TUEC dated November 29, 1984.
During a telephone call with the alleger on October 31, 1984, the TRT pre-
sented the results of the assessment of allegation AP-13 and the TRT con-
clusions to date. The TRT accompanied the alleger on a walkdown inspection
t,
on November 7, 1984. During the walkdown the alleger identified the loca-
tion on the expansion loop where the force exerted by the polar crane was
applied.
Subsequently, the alleger declined to have any further contact
with the TRT.
6. Actions Required:
Based on its evaluation of AP-13, the TRT will require
the following TUEC actions: .
. 4
1
Ccmanche Peak SSER 10 N-109
.
'
. .l
(1) Modify Gibbs & Hill. Specification 2323-MS-100, and institute proce-
dures to support the main steam line during flushing and provid'e I
temporary supports for piping and equipment in general to assure that
the quality of affected piping and equipment is not affected.
(2) Assess stresses in the portions of the Unit 1, Loop 1 main steam and
feedwater lines that were affected in the sequence of events involved
during their initial installation, flushing and final installation.
Conditions of concern are:
a. the condition when the lines were full of water and temporary .
supports had sagged or settled.
b. the condition when vibrations of the temporary line could have
occurred.
c. the condition when forces were applied by the polar crane and
come-alongs.
&
These assessments shall be based on appropriate piping configurations
involved.
(3) Perform a nondestructive examination of locations in the Unit 1, Loop 1
main steam and feedwater piping involved where stresses greater than
relevant stress allowables were exceeded during the conditions of
concern in a. through c. above.
(4) Review the existing baseline UT examinations for those portions of
the Unit 1, loop 1 main steam and feedwater piping involved in all
the conditions of concern in a. through c. above for unacceptable
indications.
(5) Review records of hydrostatic testing of the Unit 1, Loop 1 main steam
and feedwater piping to verify the quality of piping involved in the
incident.
(6) Provide similar assessments for circumstances involved in the lifting
incident identified during the TRT inspections of the Unit 1, loop 4
main steam line. -
(7) Provide assessments of effects on quality of safety-related piping
and equipment which were involved in similar incidents of sagging,
settlements and failures, if any, of temporary supports.
(8) Document the results of analysis, examinations and reviews and submit
them in a report for TRT review.
'
Reference Documents:
.
AP-4
1. NRC 01 Report 50-445/84-006. 1
2. G&H Specification 2323-MS-100 Rev. O through Rev. 8. R
l'
, I,
;
Comanche Peak SSER 10 N-110
,
J w
.
'
, ...-
3. Westinghouse Process Specification PSPAS-01, Rev. O.
4. Traveler HE-78-014-5505. *
5. Traveler HE-78-026-5505. ,
6. Traveler HE-79-035-5505. .
l
7. Traveler ME-79-036-5505. '
8. Traveler ME-79-037-5505.
9. B&R BRF-9181. i
10. B&R BRF-9211. !
11. Southwest Fabricating & Welding Drawing Q-6214-T8X.
12. B&R BRF-9348.
13. NC.R M-2305.
14. B&R QI-QAP-11.1-31,.Rev. 3.
15. TUSI CP-EP-4.0, Rev. 3. ,
16. B&R QI-QAP-11.1-26, Rev. O through Rev. 15
l
17. ASME B&PVC Section III.
18. ANSI B31.1
19. B&R CP-CPM-6.1, Rev. 6.
20. B&R CP-QAP-6.1, Rev. 4
21. Data Package WE-78-006-5505.
22. Data Package WE-78-028-5505. '
23. Data Package WE-78-025-5505.
24. Data Package WE-78-016-5505.
25. Data Package WE-78-024-5505.
26. Data Package WE-78-029-5505.
27. Data Package WE-78-030-5505.
28. Data Package WE-78-019-5505.
29. Data Package ME-78-009-5505.
30. Data Package ME-78-019-5505.
31. Data Package ME-78-010-5505.
32. Data Package ME-78-013-5505.
33. Data Package ME-78-021-5505.
34. Data Package ME-78-008-5505.
35. Data Package ME-78-020-5505.
36. Data Package ME-78-007-5505.
37. Data Package ME-78-015-5505.
38. Data Package ME-78-022-5505.
39. B&R Drawing BRP-RC-1-520-001.
40. Bechtel Specification 10466-M-204, Appendix D.
41. FSAR Figure. 5.4-12.
42. FSAR Figure. 5.4-12A.
43. FSAR Figure. 5.4-128.
4
44. FSAR Figure. 5.4-13.
4 45. FSAR Figure. 5.4-14.
46. FSAR Figure. 5.4-15.
47. FSAR Figure. 5.4-16.
48. FSAR Figure. 5.4-17. '
49. FSAR Figure. 5.4-18.
50. FSAR Figure. 5.4-19.
51. ASME B&PVC Section III.
52. ANSI B31.1
Comanche Peak SSER 10 N-111
* '
.
._. ._ _ ___.____. __ _ _..___. _ __. _ _ . _ _ _.___...-- .-.._. --__ _ k
F
-
, . ; ,-
.-
AP-9 -
1. NRC Reg. Guide 1.44.
2. G&H Specification 2323-MS-100, Rev. O, 8.
3. NRC IE Report 50-445/79-06; 50-446/79-06. 3
4. B&R CP-CPM-6.9E, Rev. I through Rev. 7. 1
5. B&R CP-CPM 6.90, Rev. 1 through Rev. 6. ;
*
6. FSAR Fig. 6.2.2-1.
7. B&R Heating Alignment Record Folder. }
8. FSAR Table 6.2.2-4.
9. J NRC OI Report 50-445/84-006.
t
10. Deposition of alleger A-21, July 12,1984, pp.1-35.
11.
12.
Deposition of alleger A-21, July 12,1984, pp. 49500-49514. l
Conversation Record, alleger A-21, 9/1/84 .
AP-13
i.
1. G&H Specification 2323-MS-100, Rev. O through Rev. 8.
2. ;
B&R letter BRF-6565. *
3. B&R PCP-1, Rev. O.
4 NRC IE Report-50-445/83-27.
5. CASE letter 2/3/83 with alleger's affidavit.
6. CASE Letter 11/28/83 with allegers affidavit.
7. NRC TRT Technical interview with alleger 9/19/85.
8.
9.
CPSES ASLB Memorandum and Order (Clarification of Open Issues).
NRC OI Report A4-83-005.
10. :
NRC OI Memorandum " Assistance to Inspection Report No. 14-83-005. ~
11. NRC 01 Supplemental Report Q4-84-007.
12. B&R M&TE Issue Record for MTE Serial MTE-357.
13. <
B&R CMC 61306 Rev. 4.
14. B&R Drawing FSM u0165, Rev. 1.
15. p
16.
TUEC Memorandum " Comanche Peak Steam Electric Station," Feb. 4, 1983. .;
B&R Mk. No. MS-s1-01-007-C12K Data Package.
17. B&R Mk. No. MS-1-01-001-C725 Data Package.
18. B&R Mk. No. MS-1-01-002-C725 Data Package.
19. B&R Mk. No. MS-1-01-005-C72K Data Package. .
I
20. B&R M. No. MS-1-01-006-C72K Data Package.
21. B&R Drawing BRP-MS-1-RB-002, Rev. 13. ,
22. B&R Drawing BRP-FW-1-R8-002, Rev. 18,
23. B&R MS-1-R8-001 FW-TS Data Package.
24. B&R FW-1-RB-001 FW-T4 Data Package. ..
25. ADLPIPE Output, 2/24/83.
26. AOLPIPE Output, DAMS 2, 8/24/84.
27. ADLPIPE Output, DAMS 5 8/24/84. 3
,
'.
.
1
Y
I
J
t
.
,
Comanche Peak SSER 10 N-112
: '
.
. *i
.. .
!
.
:
4. Actions Required of TUEC
TUEC shall submit additional information to the NRC, in writing, including a
program and schedule for completing a detailed and thorough assessment of the
issues identified in the following sections. This program plan and its imple-
-mentati.on will be evaluated by the staff before NRC considers the issuance of
an operating license for Comanche Peak, Unit 1. The program plan should address
the root cause of each problem identified and its generic implications on
. safety-related systems, programs, or areas. The collective significance of
.
these deficiencies should also be addressed. The program plan should also
include the proposed TUEC action to ensure that such problems will be precluded
from occurring.in the future. The specific actions required of TUEC are
*
described in the following sections.
'
4.1 Civil and Structural (C&S) Area
4.1.1 Rebar Improperly Installed or Omitted (See Attachment 2, C&S Category 6)
.
4
-
Provide an analysis of the as-built condition of the Unit I reactor cavity
l that verifies the adequacy of the reinforcing steel between the 812-foot
#
'
and 819-foot, -inch elevations. The analysis shall consider all required
load combinations.
4.1.2 Falsification of Concrete Compression Strength Test Results (See Attach-
ment 2, C&S Category 8)
.
-
Determine areas where safety-related concrete was placed between January
1976 and February 1977, and provide a program to assure acceptable con-
} crete strength. The program shall include tests such as the use of random -
Schmidt hammer tests on the concrete in areas where safety is critical.
The program shall include a comparison of the results with the'results
of tests performed on concrete of the same design strength in areas where
J the strength of the concrete is not questioned to determine if any signifi-
cant variance in strength occurs. TUEC shall submit the program for these i
tests to the NRC for review and approval prior to performing the tests. j
:
4.1.3 Maintenance of Air Gap Between Concrete Structures (See Attachment 2,
C&S Category 11)
f ,
?
- Perform an inspection of the as-built condition to confirm that adequate r
separation for all seismic. Category I structures has been provided. !
- Provide the results of analyses which demonstrate that the presence of i
rotofoam and other debris between all concrete structures (as determined )
by inspections of the as-built conditions) does not result in any signi- ,
ficant increase in seismic response or alter the dynamic response charac-
teristics of the Category I structures, components, and piping when cota-
{
1
pared with the results of the original analyses. y
)
4
3
.
h(h&-% ,j
Comanche Peak SSER 8 K-16 N. ,
6 )
,
-
-
*~
$$ . .
l! '
s
:
: e. (1) The reinforcing steel that was placed between the 812-foot and
b
819-foot, \-inch elevations in the reactor cavity wall of the
L
!
.
Unit 1 Reactor Building was completed and inspected to draw-
ing 2323-51-0572, Rev. 2. After the concrete was placed, Brown
& Root received Rev. 3 to the drawing showing a substantial in-
[ crease in reinforcing steel over that which was installed.
l G&H
engineering was informed of the omission by Brown & Root non-
conformance report C-669, which is referenced in the Brown &
Root internal deficiency report CP-77-6. G&H engineering replied
i
that the omission of this additional reinforcing steel did not
in any way impair the structural integrity of the structure.
i G&H stated that the additional rebar was added as a precaution
against cracking which might occur in the vicinity of the neutron
detector slots should a loss of coolant accident (LOCA) occur.
A portion of the omitted reinforcing steel was placed in the
next concrete lift above the 819-foot, 1/2-inch elevation. G&H
stated that this was done to partially compensate for the rein-
[
4
forcing steel omitted below and to minimize the overall area
,
subject to possible cracking.
The TRT requested documentation to indicate that an analysis was
performed supporting this conclusion. The TRT was subsequently
U
informed that an analysis hac* at been performed.
, (2) In response to Brown & Root construction's Request for Informa-
p tion or Clarification (RFIC) RBCR-37, Design Change / Design Devia-
@ tion Authorization (DC/DDA) No. 832 was issued stating that the
!- configuration of the 2x9-No. 9 reinforcing bars (two rows by
[ nine layers), as shown on drawing 2323-S1-0572, Rev. 4, could be
changed to a continuous circumferential arrangement. The TRT
,j reviewed this drawing and determined that these bars were among
g those omitted in the concrete placement between the 812-foot and
1 819-foot, 1/2-inch elevations and subsequently placed above the
d 819-foot, 1/2-inch elevation (See e(1) above.) Revision 4 shows
U each of the four sets of No. 9 bars used to form the configura-
:'
x tion required were to be bent in two places to form an approxi-
mate circular configuration when placed. The DC/DDA stated the
bars could be bent to a specified radius to form a true circular
,' arrangement. The change, therefore, only affected the way in .
which the bars were bent and did not reduce the load-carrying )
; capacity of the structure. 1
1
(3) During the placement of reinforcing steel within the triangular
columns surrounding the reactor cavity at the 826-foot, 11-inch
., elevation, interferences were encountered. The horizontal tails
i
,
of the No.11 vertical reinforcing bars were interfering with
14-inch-diameter sleeves already in place. The TRT reviewed
l DC/DDA No. 6918 and the attached sketches which showed that I
' six bars were cut and replaced with bars tailed up to achieve
total anchorage and three bars were bent down to clear the
sleeves.
Also, due to congestion problems, the design of the No. 4 e
stirrups surrounding the ten No.18 circular bars was modified j
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CPSES/FSAR
-e 10.3 MAIN STEAM SUPPLY SYSTEM ,
,
\
10.3.1 DESIGN BASES "
The main steam supply system is considered part of the Main Steam,
Reheat, and Steam Dump System. It conveys steam from the outlet of the
steam generators to the various system components throughout the
Turbine Building. The . steam is primarily used for driving the main
turbine and for heating service in the MSRs. In addition, it is used
for various auxiliary services such as the following:
1. Steam generator feed pump turbine
2. Auxiliary feed pump turbine
.
3. Steam dump' system
,
4. Turbine shaft seal system
t.
5. Plant process steam system
Steam from the outlet of the four steam generators (a total of
15,140,016 lb/hr, at 1000 psia and 0.25 percent moisture) is routed in
the main steam piping to the turbine. The piping is sized for a
pressure drop of less than 25 psi. The design pressure-temperature
rating of the main steam piping is 1200 psig at 6500F. The full-load
steam flow from the reheaters to the feedwater pump turbines is
approximately 242,472 lb/hr at 152.1 psia. The steam supply from the
main steam lines to the auxiliary feedwater pump turbine is in the
range of 9,575 to 51,524 lb/hr at 1200 to 125 psia, respectively.
The main steam supply system is used for plant cooldown. This is
achieved by progressively lowering the pressure of the steam
generators; the decay heat and the sensible heat are removed by the
n
10.3-1
: eta-26-3h
'
- .
Als ii
IT
*
.
-
. .~.
CPSES/FSAR
generation of steam. When steam generator pressure has been reduced to
125 psia, the Residual Heat Removal (RHR) System is placed in I
operation.
The main steam supply system is designed to meet all applicable
requirements of 10 CFR Part 50. This system is designed in accordance
with NRC Regulatory Guides 1.26 and 1.29.
The main steam safety valves are rated to pass 105 percent of the ,
engineered safeguard design (ESD) steam flow at a pressure not
exceeding 110 percent of the main steam system design pressure (ESD -
flow equals 104.5 percent of the guaranteed 3425 MWt). These safety
valves prevent overpressurization of the system as specified in the
ASME 8&PV Code.
The steam generator power-operated relief valves pass 10 percent of the
steam flow used for the plant design at no-load stean pressure
(1107 psia).
(
The main steam isolation valves (MSIVs) are designed to stop flow from
either direction within 10 sec (five sec af ter receipt of signal to
close) to prevent uncontrolled steam release from more than one steam !
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generator. <
,
The steam dump system design bases are given in Section 10.4.4. .
The main steam lines from the steam generator, out through the
Containment, and up to the first moinent restraint beyond the MSIV, are
)
designed in accordance with the ASME B&PV Code, Section III, Code
Class 2. Beyond this moment restraint, the main steam lines to the
main turbine and the lines to tne feedwater pump turbines, the
condensers (steam dump), and the reheaters are non-nuclear-safety- !
related and are designed in accordance with ANSI B31.1, Power Piping.
10.3-2 k
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CPSES/FSAR
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The lines to the auxiliary feedwater pump turbine are designed in
accordance with the ASME B&PV Code, Section III, Code Class 3. !
The MSIVs, integral bypass valves, and bypass piping are of Code Class
1 design. (The applicant has optionally upgraded this equipment to
Code Class 1.) The MSIV actuators, safety valves, and power-operated
relief valves are of Code Class 2 design.
The piping is insulated, and where exposed to the outdoor environment,
it is suitably protected from the weather.
The total plant releases associated with a LOCA, including steam side
leakage, are within the 10 CFR Part 100 limits. The steam generator
shell and lines which emanate from the steam generator shell side are
barriers against release of containment atmosphere.
Steam is conveyed from the steam generators to the main turbine by four
steam lines. Upstream from the MSIVs, each line is provided with five
spring-loaded safety valves and one power-operated relief valve. It is
essential that the heat load be evenly shared between the four loops of
the Nuclear Steam Supply System (NSSS). To achieve this, the four main
steam lines are interconnected by a pressure equalizing header
'
downstream of the steam generator isolation valves.
l '
The sizes and routing of the main steam piping are such that under
design flow, a pressure drop of approximately 22 psi is obtained 13
between the steam generator nozzle and the turbine stop valve inlet *
with a steam velocity of approximately 115 ft/sec. Each steam line
uses a drain system to remove accumulated condensate from the line.
l
To ensure steam supply to the auxiliary feedwater pump turbine (even in
! the steam generator isolation), two separate steam supply lines are
provided from two main steam lines.
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AMENDMENT 13
i0.3-3 DECEMBER 15, 1980 a
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CPSES/FSAR
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A direct connection is provided downstream of the isolation valves *for
the startup of the steam' generator feedwater pump turbines.
Table 10.3-1 shows the design bases of the main steam piping.
The environmental design bases are given in Section 3.11.
55lTheinserviceinspectionrequirementsaregiveninSection6.6.
The inservice inspection program for the turbine generator is described
in-Section 10.2.3.6.
10.3.2 DESCRIPTION
The main steam supply system and interconnected piping are shown
schematically on Figure 10.3-1. The nuclear safety class is applied to
the main steam lines from the steam generator, up to and including the
first moment restraint beyond the MSIVs located outside the (
Containment. The nuclear safety class is also applied'to the blowdown
and process sampling lines from the steam generator, up to and
including the pneumatically operated isolation valves. (See Figure
.10.3-1, sheet'l of 2.) The non-nuclear-safety class portion includes
the remaining portion downstream of the moment restraint. (See Figure
10.3-1, sheet 2 of 2.)
Each main steam line has a number of branch-off lines located
12 downstream of the main steam isolation valve. All these branch-off
'
Q040.136 -lines are provided with steam shut-off valves as listed in Table
10.3-11.
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10.3-4
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AMDEMENT 55
-JULY 19, 1985~ . .
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CPSES/FSAR
.
(.3 The graphs of the steam generator compartment pressures
resulting in the maximum differential pressure on the steam
generator are presented as Figure 6.2.1-29. These result
from the 763-in.2 hot-leg split.
The graphs of the pressurizer compartment pressures
resulting in the maximum differential pressure on the
pressurizer are presented as Figure 6.2.1-30.
For all differential pressures on structures or components,
the structure or component across which the differential
_ pressure is to be calculated must be located on the
l schematic drawings mentioned in Section 6.2.1.2.2,
Paragraph A. When the relevant volume nodes have been
! located, the differential pressure history can be found
from the differential in the pressure-time histories given 10
e ' for the volumes in Section 6.2.1.2.3, Paragraph D. QO22.3
(
E. Structural Peak Pressures
I. Peak pressures, used in compartment structural design, are
given in Table 6.2.1-23. Note that in the case of the
reactor cavity, the nozzle cavity pressure represents the
peak pressure which could be experienced by Volumes 1-8
(the design pressure for the hot-leg inspection compartment
being set equal to that of the cold-leg compartment, even
though evaluation shows a higher pressure would exist in a
cold-leg inspection compartment). The pressure between El .
823.125 ft and El. 834 ft.-0.5 in. is given by nodes 9 to
16, between 809.844 ft and 823.175 ft is given by nodes 17
to 24, between 798.667 ft and 809.844 ft is given by nodes
25 to 28, between 783 ft 7 in. and 798.6672 ft is given by
node 29. Volume 32, the free containment volume, is
evaluated for maximum pressure by procedures described in
<
6.2-14k
fotA-W%AMENDMENT 10
MARCH 31, 1980
- _ _ _ _ . _ _ . _ _ _ - - _ _ _ . _ _ _ _ _ _ . _ _ - _ _ _ _ - . _ - _ _ _ . _ _ _ . _ . _ _ _ -. ..J
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CPSES/FSAR
TABLE 6.2.1-23
SUBCOMPARTMENT PRESSURE SUEBARY
Design Pressure
Location (psia)
Mozzle cavity (el. 823 ft to
*
el. 829 ft 4 in.) - 475
Reactor cavity (el. 829 f t 1/2 in. to
el. 834 f t 1/2 in.) 67
Reactor cavity (el. 822 f t to
el. 829 f t 1/2 in.) 75
Reactor cavity (el. 810 ft to
el. 822 f t) 44
Reactor cavity (el. 798 f t 10-1/2 in,
to el. 810 ft) 7
10
( Reactor cavity (el. 783 f t 7 in. to
el. 798 f t 10-1/2 in.) 7
Steam generator compartment -
(below el. 858 f t 6 in.) 46
Steam generator compartment
(el. 858 ft 6 in. to 895 ft 9 in.) 17
Steam generator compartment
(above el. 895 f t 9 in.) 12
Pressurizer compartment
(above el. 853 ft 6 in.) .
21.0
Pressurizer compartment
(below el. 853 f t 6 in.) 28.0
AMENDMENT 10
MARCH 31,1980
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NUREG-0797
Supplement No. 7 .
.
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Safety Evaluation Report
related to the operation of
Comanche Peak Steam Electric Station,
Units 1 and 2
Docket Nos. 50445 and 50-446
Texas Utilities Generating Company, et al.
_
U.S. Nuclear Regulatory
l Commission
Office of Nuclear Reactor Regulation
January 1985
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ACRONYMS AND ABBREVIATIONS
AA -
independent assessment program allegation
AB -
American Bridge
AB ,- bolt allegation
AC ;-
concrete /rebar allegation
ACI -
American Concrete Institute
AD -
design of pipe / pipe support allegation
ADS -
audit discrepancy report
AE -
electrical allegation
AE00 - Office for Analysis and Evaluation of Operational Data (NRC)
AFW -
auxiliary feedwater system
AH -
hanger allegation
-
AI -
intimidation allegation
AISC - American Institute of Steel Construction
AM -
miscellaneous allegation
ANI -
authorized nuclear inspector
/J;S -
American Nuclear Society
ANSI - American National Standards Institute
A0 -
protective coating allegation
AP -
pipe and pipe support allegation
APC -
AMP Product Corporation
AQ
-
quality assurance / quality control allegation
AQB
-
QA/QC bolt allegation
AQC
-
QA/QC concrete /rebar allegation
AQE
-
QA/QC electrical allegation
AQH
-
QA/QC hanger allegation
AQO
-
QA/QC coating allegation -
AQP
-
QA/QC pipe and pipe support allegation
AQW - QA/QC welding allegation
ARMS - Automated Records Management System
ASLB - Atomic Safety and Licensing Board
ASME - American Society of Mechanical Engineers
ASTM - American Society for Testing and Materials
AT -
acceptance test
AT -
test program allegation
AV -
vendor / generic allegation
AW -
welding allegation
- B&PVC - Boiler & Pressure Vessel Code
B&R -
Brown & Root, Inc.
BRIR - Brown & Root Inspection Report
BRHL - Brown & Root Hanger Locations
BRP -
Brown & Root piping isometric drawing
CAR -
Corrective Action Request
CASE - Citizens Association for Sound Energy
Comanche Peak SSER 7 . vii
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~C&L:,- Corner and Lada (computer program) *
C&S - civil and structural
CAT . Construction Appraisal' Team (NRC)
C8&I.- Chicago Bridge & Iron Company
CCS -
Component Cooling System
CEL - Coating Exempt Log
Code of Federal Regulations
'
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CFR
~CHN - construction hold notice
CILRT_- containment integrated leak rate test
CMC -
component modification cards
COT - construction operation traveler
CP -
Comanche Peak
CP -
construction permit
CPPE - Comanche Peak Project Engineering
CPSES - Comanche Peak Steam Electric Station
CPSIG - Comanche Peak Seismic Interaction Group
CSTS - Construction and Startup/ Turnover Surveillance Group (TUEC)
CVCS - chemical and volume control system
CZ-11 - Carboline Carbo zinc 11
,
DBA - design basis accident
i DCA -
design change authorization
DCC -
Document Control Center (TUEC)
DCTG : Design Change Tracking Group
.DCVGi- design change verification group '
DE -
Division of Engineering (NRC)
j DFT -
dry film thickness
- DL -
Division of Licensing (NRC)
D-6 -
Ameron Dimetcote 6
, EDO -
Executive Director for Operations (NRC)
l E&I -
Electrical and Instrumentation
l ETG -
Electrical Test Group (TUEC)
i
FDSG - Field Damage Study Group (TUEC)
FJO -
field job orders
FP -
fire protection
FSAR - Final Safety Analysis Report
FW -
field weld
l. GAP - Government Accountability Project l GDC' - general design criteria
GE - General Electric Corporation
:GED
- General Equivalency Diploma
G&H --
Gibbs & Hill
GHH - Gibbs & Hill hanger (isometric drawing)
Comanche Peak SSER 7 viii .
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HFT -
hot functional test
HIR - hanger inspection report .
HP -
hanger package
HVAC - heating, ventilation and air conditioning system
HX -
heat exchangers
IAP -
Independent Assessment Program
IE -
Office of Inspection and Enforcement (NRC)
IEEE - Institute of Electrical and Electronics Engineers
IM -
interoffice memorandum (TUEC)
INP0 - Institute for Nuclear Power Operations
IR -
inspection report (NRC)
IRN -
item removal notice
ITT-G - ITT Grinnell
JTG -
Joint Test Group (TUEC)
JUMA - Joint Utility Management Assessment Group
LOCA - loss of coolant . accident
LP -
liquid penetrant
MAR -
maintenance action request
M&P -
mechanical and piping
MCC -
motor control center (GE)
MDB - master data base
MIFI - mechanical fabrication inspector
MIL -
material identification list (or log)
MIME - Mechanical Equipment Inspector
MQE
-
Mechanical Quality Engineering
MRS -
manufacturer's record sheet
MWDC - multiple weld data card
N/A - not applicable
NCR - nonconformance report (TUEC)
NDE - nondestructive examination
HDT - nondestructive testing
NI -
never inccrporated
l NONSAT - nonsatisfactory l NOV - Notice of Violation (NRC) l
NPSI - Nuclear Power Service Incorporated
NRC -
U.S. Nuclear Regulatory Commission
NRR -
Office of Nuclear Reactor Regulation (NRC)
NSSS - nuclear steam supply system
l O&M -
Operations and Maintenance (TUEC)
! OBE -
operating basis earthquake
l
OI -
Office of Investigations
OJT -
on-the-job training
l l
Comanche Peak SSER 7 ix
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~ OL -
operating license -
ORNL - Oak Ridge National Laboratory'
PC -
protective coating
PET -
permanent equipment transfer
PFG -
paper flow group
-PFS -
pipe fabrication shop
PS4R - Preliminary Safety Analysis Report
PSE -
, Pipe Suoport Engineering (TUEC)--
<
PT - -
preoperational test-
PWR -
pipe whip restraints ,
P-305
~
Carboline Phenoline 305
QE -
quality engineer
,
QA -
quality assurance
QAI;
_
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quality assurance investigation (TUEC)
QC
-
quality control ,
l
RCB '
Reactor Containment Building
RES - Office of Nuclear Regulatory Research (NRC)
RFIC - request for information or clarification (B&R)
RG -
Regulatory Guide (NRC)
RI -
NRC Region I Office
. RIR -
receipt inspection report (TUEC)
RIV -
NRC Region IV Office
RHRS - residual heat removal system
RPI - rod position indication
RPS - report process sheet (TUGCO)
RPV - reactor pressure vessel
RPVI - reactor pressure vessel reflective insulation
RRI - Resident Reactor Inspector (NRC)
RV -
reactor vessel
RWN -
room work notifications
SAP -
startup administration procedure
SALP - -Systematic Assessment of Licensee Performance (NRC)
SAT -
satisfactory
SAVC - structural assembly verification card
SER -
Safety Evaluation Report-(NRC)
SIS -
Special Inspection Services
SMAW - shielded metal arc welding
SNM -
special nuclear material
50RC - Station Operations Review Committee
SRIC - Senior Resident Inspector for Construction (NRC)
SRT' - Special Review Team (NRC)
SSE -
safe shutdown earthquake
SSER - Safety Evaluation Report Supplement
SSPC - Steel Structures Painting Council
SSWP - station service water pumps
SSI -
safe shutdown impoundsent
i
Comanche Peak SSER 7 x .
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STE~ - system test engineer
SWA -
s.tartup work authorization *
TDCR - test deficiency change request
TDR -
test deficiency report
10 CFR 50 - Title 10 rode of Federal Regulations Part 50
TIDC --- Division of Technical Information and Document Control (NRC)
THE -
TUEC Nuclear Engineering
TP -
test program
TPD -
test procedure deviation
Tr -
transcript
TRT - Technical Review Team (NRC)
TSI -
thermolag
TSMO - Technical Services Mechanical Drafting
TSP -
tri-sodium phosphate
TUEC - Texas Utilities Electric Company
TUGC0 - Texas Utilities Generating Company
TUSI - Texas Utilities Service, Inc.
'
UCC -
University Computing Company
1 UT -
ultrasonic test
VCD - vendor-certified drawing
VT -
visual weld (inspector)
WDC - weld data card
WFML - weld filler metal log
WPS - welding procedure specification
Comanche Peak SSER 7 ,
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Ecluate the adequacy of craft personnel training in the use of instaJla- y
tion manuals to establish root causes and appropriate corrective actions. f
,
This action shall be. integrated with other actions concerning craft i {
?=
personnel training addressed under QA/QC Category 8, "As Built." 0
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4.1.3 Electrical Equipment Separation (See Attachment 2, E&I Category 3) h
-
Reinspect all panels at CPSES, in addition to those in the main control e
room for Units 1 and 2, that contain redundant safety-related cables within /
conduits or safety and nonsafety-related cables within conduits, and : ,
', either correct each violation of the separation criteria, or demonstrate t
by analysis the acceptability of the conduits as a barrier for each case
where the minimum separation is not met.
4
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Reinspect all panels at CPSES, in addition to those in the main control !
, . room identified in Table 1 of SSER for E&I Category 3, and either correct i
each violation of the separation criteria concerning separate cables and ;r
cables within flexible conduits, or demonstrate by analysis the adequacy - i.
of the flexible conduit as a barrier. -
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Correct two instances of violation of the separation criteria inside
- panels CPI-EC-PRCB-09 and CPI-EC-PRCB-03 concerning a barrier that had
.been removed and redundant field wiring not meeting minimum separation.
.
. Submit the analysis that substantiates the acceptability of the criteria
stated in the electrical erection specifications governing the separation ,
between independent conduits and cable trays. ;
;
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Evaluate the adequacy of the QC inspection program as related to the u
deficiencies identified above to establish root causes and appropriate i .'.
corrective actions. These actions shall be integrated with other actions
l' ,
addressed under E&I Category 6, " Electrical QC Inspector Training and -
Qualifications," and QA/QC Category 8, "As Built."
4.1.4 Control Room Ceiling Fixture Supports (See Attachment 2, E&I Category 4) i.
i -
Substantiate (1) the adequacy of the overall seismic support system i <.
4 installation for all the items located above the ceiling in the control y
room, including nonsafety-related conduit, suspended ceiling and lighting !
and (2) the adequacy of the seismic support system installation for -
+
nonsafety-related conduit in Seismic Category I areas of the plant other :
.'
.
< than the control room. This action shall be integrated as appropriate
with other actions addressed under Civil / Structural Category 14, " Seismic "
Design of Control Room Ceiling Elements." i
y
4.1.5 Electrical QC Inspector Training / Qualifications (See Attachment 2, N
E&I Category 6)
.
{
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Evaluate the testing program for QC electrical inspector qualifications . di.
and develop a testing program which optimizes administrative guidelines, F
procedural requirements and test flexibility to assure that suitable
proficiency is achieved and maintained.
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m- atwr w rvy y ag we rg- w-w vWWp-w-M-tr-^ q wwwmee em + WC'"'"~P- ". "8'"""N-WTW"""'"9"*"""*N~***'*"MW***"Y""* " --" *"~ '
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1. Allegation Category: Electrical and Instrumentation 4, Control , Room
Ceiling Fixture Supports
2. Allegation Number: AE-17
3.
Characterization: It is alleged that the field run conduit, drywall,
and lighting installed in the area above the equipment panels in the
control room were classified as nonseismic, and as such were only
supported by wires.
4. Assessment of Significance: The implied safety significance is that the
seismic qualification of certain equipment located above the ceiling in
the control room could be indeterminate and consequently its behavior
during a seismic event could not be predicted.
The central concern of this allegation is whether Texas Utilities Electric
Company (TUEC) considered the positions of Regulatory Guide (RG) 1.29,
"
" Seismic Design Classification," as augmented by Final Safety Analysis
Report (FSAR) Section 3.2.1.2, " Seismic Category II," during the design of
the support systems in the control room for the nonsafety related field
run conduit, for the suspended drywall ceiling, and for the lighting
fixtures.
,
Regulatory Guide 1.29 states that nonsafety-related structures, systems,
or components whose failures could reduce the functioning of any plant
;
l l ,
feature to an unacceptable safety level or could result in incapacitating
j
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injury to occupants of the control room should be designed and constructed
so that the safe shutdown earthquake (SSE) would not cause such failure.
1
FSAR Section 3.2.1.2 provides TUEC's commitments to these positions, and
d designates as seismic Category II the nonsafety related equipment that
,
will be encompassed by the positions of RG 1.29.
t
Field Run Conduit. The NRC Technical Review Team (TRT) examined conduit
seismic installation notes and detail drawings, design change author-
) iji
izations (DCAs), work packages, physical configuration drawings and other
documents pertinent to this issue. The TRT also inspected conduit
lh i
l installation in the area above the control room ceiling and determined
that the safety-related' conduit was fastened by seismic Category I
,
.;
supports typical of those used in other areas of the facility. The
L nonsafety-related conduit was secured by supports which were of a dif-
ferent design than those for safety-related conduit. None of the non-
li safety-related conduits examined by the TRT were greater than 2 inches in
; diameter. In addition, they were not supported by seismic C:ategory I
J
supports and did not have seismic Category II cable restraints. The TRT
i dete~rmined that engineering drawing 2323-5-0910, " Conduit and Junction
!
Box Supports," did not require seismic Category II cable restraints for
j
nonsafety-related conduits less than or equal to 2 inches in diameter,
but required them for conduits greater than 2 inches in diameter.
<
The TRT also examined similar nonsafety-related conduit installations
t
in other seismic Category I areas of Unit I and found that seismic
Category II stainless steel cable restraints were used as bcckup to the
,
nonseismic dead weight supports for the conduits greater than 2 inches
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in diameter. Tl.e TRT staff also found that the installation of nonsafety-
related conduit less than or equal to 2 inches in diameter in the' control
room was consistent with that used throughout the plant.
,
Suspended Drywall Ceiling. The TRT found that the suspended ceiling
above the central part of the control room was made of drywall sheets
arranged to form a sloping wall around that area. These drywall sheets i
were fastened to a metal framework (metal batten) supported by thin- :
'
walled channels (1-1/2-inch by 1/2-inch) attached to the primary building
concrete. The metal framework was also attached to the concrete by a i
system of 1/8-inch stainless steel cables such that if the thin-walled
channel supports failed during a seismic event, the weight of the framing
and drywall would be assumed by the cabling.
Lighting Fixtures. The TRT reviewed the installation of the lighting
fixtures over the control panels and central part of the control rcom
'
and found that they were supported from an intermediate substructure
of "unistrut" by light-weight conduit. The substructure was likewise
supported by light-weight conduit from the primary building ceiling.
,
The conduit used is typical of that supporting the light fixtures in
suspended ceiling applications. Parallel with each lighting support
conduit are two 1/8-inch stainless steel cables which would assume the
load if the support conduit or its attachment were to fail. Other
individual light and reflector assembly fixtures, separate from those
supported by the intermediate "unistrut" substructure, were secured by
a similar type of conduit and backup cable design arrangement with the
cable attached to the edge of the light reflector assembly.
Based on the review of engineering drawings and direct inspection of
the installation, the TRT determined that the positions of RG 1.29, as
augmented by FSAR Section 3.2.1.2, were not met by the installation of
the fixtures located in the area above the panels and central part of the
control room.
As discussed above, the nonsafety-related conduit in the area above the
control room suspended ceiling was not fastened by seismic Category I
supports and/or seismic Category Il cable restraints. With regard to the
suspended drywall ceiling, it appeared that the installation met TUEC
commitments to the positions of RG 1.29. However, the final resolu-
tion of this technical issue, including the nonsafety-related conduit
support system, will depend on the review and approval by the TRT of an
t
analysis to be provided by TUEC concerning the adequacy of the seismic
support system installation in the control room.
e
The TRT inspected selected seismic Category I areas of the plant, reviewed
( associated engineering drawings, and determined that only nonsafety-related
conduits of less than or equal to 2 inches in diameter were not fastened
s
by seismic Category II cable restraints.
h
i' 5. Conclusions and Staff Positions: The TRT concludes that the installation
b of the nonsafety-related conduit in the control room appears to be
4 inconsistent with the positions of RG 1.29. Accordingly, this part of
the allegation is of concern. With regard to the suspended ceiling and
J-46
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< t lighting supports, the acceptability of the installation will depend on ,
the approval by the TRT of the analysis to be provided by TUEC concerning i
j
the adequacy of the seismic Category II restraints in the control room. I
" This technical issue, including the nonsafety related conduit support
!
system, will be resolved after the review of TUEC's seismic analysis
1 substantiating the adequacy of the overall seismic support system installa-
tion in the control room. The results of the TRT review of TUEC's
J
.
analysis will be reported in a supplement to this SSER.
f i
s
Based on the review of other seismic Category I areas of the plant,
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the TRT concludes that the acceptability of the installation will
depend on TRT approval of TUEC's analysis of the adequacy of the
seismic support installation for nonsafety-related conduits in areas
4
of the plant other than the control room.
c
il The TRT further concludes that the lack of analysis to substantiate the
adequacy of the seismic design installations inspected may be an indica-
,
tion of weakness in the QA/QC program concerning design control. This
area is addressed under the QA/QC Category 1, " Design Process."*
5. Action Required: TUEC shall perform the following actions prior to fuel
*
load:
.I
(a) Provide the TRT with analyses that substantiate (1) the adequacy of
the overall seismic support system installation for all the items
located above the ceiling in the control room, including nonsafety-
Q
t
related conduit, suspended ceiling, and lighting fixtures and (2) the
adequacy of the seismic support system installation for nonsafety-
l related conduit in seismic Category I areas of the plant other than
the control room. This action shall be integrated as appropriate
with other actions addressed under Civil and Structural Category 14,
" Seismic Design of Control Room Ceiling Elements."
i
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(b) Evaluate the adequacy of the QA/QC program related to the deficiencies I
!
identified above to establish root causes and appropriate actions. l
These actions should be integrated with other actions addressed under
.
the QA/QC Category 1, " Design Process."
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*The TRT evaluation of QA/QC allegations is in progress and will be published
j in a subsequent supplement to this SSER. i
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C COMANCHE PEAK RESPONSE TEAM
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ACTION PLAN
Item Number: II.a
.
Title:
Reinforcing Steel in the Reactor Cavity
Revision No. 0 1
2
Description Original Issue
Revised to Reflect incorporates
NRC Comments SSER
fed 4L /
10l6 } 8$
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' Prepared and
Recommended by:
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Page 1 of 7
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ITEM NUMBER II.a
Reinforcing Steel in the Reactor Cavity
.
1.0 DESCRIPTION OF ISSUE IDENTIFIED BY NRC
.
-
The TRT investigated a documented occurrence in which reinforcing
steel was omitted from a Unit I reactor cavity concrete placement
between the 812-foot and 819-foot 1/2 inch elevations. This
reinforcement was installed and inspected according to drawing
.2323-S1-0572, Revision 2. However, after the concrete was placed.
Revision 3 to the drawing was issued showing a substantial increase
in reinforcing steel over that which was installed. Gibbs & Hill
Engineering was informed of the omission by Brown & Root Non-
conformance Report CP-77-6. Gibbs & Hill Engineering replied that
the omission in no way impaired the structural integrity of the
structure. Nevertheless, the additional reinforcing steel was
added as a precaution against cracking which might occur in the
vicinity of the neutron detector slots should a loss of coolant
accident (LOCA) occur. A portion of the omitted reinforcing steel
was also placed in the next
level. This was done to' par
te lift above the 819-foot -inch
compensate for the reauforcing
steel omitted in the previous .te lift and to minimize the
overall area potentially subject . cracking.
' The TRT requested documentation indicating that an analysis was
performed supporting the Gibbs & Hill conclusion. The TRT was
subsequently informed that an analysis had not been performed.
Therefore, the TRT cannot determine the safety significance of this
issue until an analysis is performed verifying the adequacy of the
reinforcing steel as installed.
2.0 ACTION IDENTIFIED BY NRC
Accordingly, TUEC shall provide an analysis of the as-built
condition of the Unit I reactor cavity that verifies the adequacy
of the reinforcing steel between the 812-foot and 819-foot 1/2-inch
elevations. The analysis shall consider all required load
combinations.
3.0 BACKGROUND
3.1 Information Supplementing NRC Description of Issue
The concrete placement of the reactor cavity wall between
elevations 812'-0" and 819'O " was made according to
Revision "2" of Gibbs & Hill drawing 2323-S1-0572. Subsequent
Revision "3" of the same drawing added reinforcing steel in
(-
f part of the wall that was already constructed in accordance
with the prior revision of the drawing. Upon receipt of this
drawing in the field, SDAR CP-77-6 and NCR-C-669 vere issued
to document that the concrete placement had been performed to
.
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Revision: 2
Page 2 of 7
ITEM NUMBER II.a
(Cont'd)
3.0 BACKGROUND (Cont'd)
Revision "2" of the drawing and the reinforcing steel added
A
per Revision "3" was not installed. Gibbs & Hill reviewed the
design with respect to this as-built condition and determined
the as-installed reinforcement to be acceptabic. The Gibbs &
Hill Structural Job Engineer had these rebars added on
" Revision "3" based on engineering' judgement to minimise the
possibility of cracking, although calculations for the cavity
had not required these bars. The Gibbs & Hill reply forwarded
by GTN-19823 confirmed the as-built condition did not impair
the structural integrity of the reactor cavity wall. Specific
calculations justifying the adequacy of the reactor cavity
were not generated by Gibbs & Hill, because the initial design
per Revision "2"
of the drawing along with its design
calculations did not require the reinforcement.
3.2 Preliminary Determination of Root Cause and Generic
Implications
The preliminary reviews conducted to date have not identified
.
a root cause for this item. Implementation of the action plan
is necessary before a root cause determination can be made.
The consideration of generic implications will include a
review of all reinforcement omissions as documented on project
NCRs. Generic implications will be evaluated based on the
results of the implementation of the action plan and the
results of the root cause determination.
4.0 CPRT ACTION PLAN
4.1 Scope and Methodology
This action plan Ie designed to assess the design adequacy of
the existing as-built condition of reactor cavity wall and
other areas within Units 1 & 2 where rebar were omitted
including an evaluation of the engineering / field design change
interface. An analysis of the as-built reactor cavity wall
vill be performed to demonstrate adequacy of installed
reinforcing steel considering all applicable loading
combinations. Engineering calculations with applicable
assumptions stated therein will be performed to evaluate the
subject wall with "as installed" reinforcing steel between
elevations 812'-0" and 819'-01". s A third-party reviewer will
verify the adequacy of the calculations.
The circumstances and engineering evaluation that led to the
.
'
provision for, and subsequent deletion of, the subject
reinforcement in reactor cavity wall vill be analyzed in light
of design intent and structural integrity of the wall.
.
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Revision: 2
-_ Pag,e 3 of 7
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ITEM NUMBER II.a
(Cont'd)
4.0 CPRT ACTION Pl.AN (Cont'd)
.
,
,
In order to consider possible generic implications, all
- instances of reinforcement omissions for Units 1 & 2 as
'
documented in project NCRs will be researched. This effort
will cover all the safety related Class I building structures.
A review of every case will be made to ascertain proper
engineering evaluation and documentation exists in support of
the disposition of each item. Additional documentation will
be developed as required to insure appropriate disposition.
As a further test, a random sample of 60 concrete pour cards
will be selected and reviewed to verify that the current
design documents pertaining to reinforcement were'used in
construction. This activity will also provide an increased ~
level of assurance that all instances of reinforcement
omissions were identified. For this purpose, the revision
numbers of the reinforcement drawings referenced on the pour
cards will be comparsd to the current design documents and all
reinforcement differences noted and evaluated.
,
,
An overall review of procedures governing design changes will
be performed to verify the adequacy of methods for controlling
-
Laplemention of design changes into construction. Emphasis
will be placed on a review of project procedures used to
convey impending design changes immediately affecting ongoing
+ construction activity. The effectiveness of procedures
controlling such construction activities will be verified by
reviewing implementation of all rebar omission cases and major
embedments. The engineering field interface review of major
l l embedments will serve as another " test" of the adequacy of
controls implemented by the civil discipline. The flow of
activities initicted by engineering to convey impending design
changes to the field as well as action leading towards
appropriate construction hold will be identified and
evaluated. ;
t
1
i L Upon conclusion of the procedural implementation review of ! rebar omissions and embedments, a determination will be made
i
of the potential applicability of any findings to other areas
within the civil discipline. The engineering field design '
change interface review will be broadened to include further
evaluation outside the civil discipline if findings suggest f'
that other disciplines could have similar problems. !
The overall process for closure of this issue will be reviewed
by the third-party reviewer. I
i(/ !
_,
The attached logic diagram identifies major tasks and the
l interrelationships of taska necessary for meeting the E L
preceding objectives.
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Revision: 2
Page '.4 of 7
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l ITEM NUMBER II.a
(Cont'd)
4.0 CPRT ACTION PLAN (Cont'd)
4.2 Responsibilities
I.
~ The organizations and personnel that will participate in this
effort are described below with their respective scopes of
work.
4.2.1 Comanche Peak Project Civil Engineering
4.2.1.1 Scope
-
Reviewing project NCR's for rebar
omission cases
- Developing additional documentation if
required to insure appropriate
disposition of rebar omission NCR's.
'
.
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Reviewing pour card for rebars
-
Reviewing pour card for embedments
-
Aaatstance in procedural review
governing design changes and control of
construction activities
-
Assistance in overall engineering
evaluation and development of Action
Plan Results Report
4.2.1.2 Personnel
Mr. C. R. Hooton Project Civil Engineer
Mr. D. C. Patankar Civil / Structural Lead
Engineer
4.2.2 Gibbs & Hill, Inc., New York, N.Y.
4.2.2.1 Scope
-
Cibbs & Hill, Inc., Structural
Department - will perform the analysis /
design calculations as required under
this action plan including the required
design review of these calculations.
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Revision: 2
Pag,e 5 of 7
.
ITEM' NUMBER II.a
,
(Cont'd)
4.0 CPRT ACTION PLAN (Cont'd)
4.2.2.2 Personnel
I-
Mr. E. L. Beskor Structural Job Engineer
Mr. A. M. Kenkre Structural Squad Leader
Mr. S. Sengupta Senior Engineer
Mr. C. Zion
-
Senior Engineer
Mr. M. N. Shah Senior Engineer
4.2.3 Third-Party Activities
4.2.3.1 Scope '
.
-
perform design review of calculations
prepared by Gibbs & Hill, Inc.
-
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review of engineering / field design
change interface.
-
overview of NCR and pour. card reviews
-
evaluation of overall conclusions
-
preparation of Results Report
4.2.3.2 Personnel
Mr. H. A. Levin TERA - CPRT Civil /
Structural Review Team
Leader
Dr. C. Mortgat TERA - Senior Structural
Engineer
Dr. J. Arros TERA - Structural
'
Engineer
4.3 Personnel Qualification Requirements
Participants in the implementation of this action plan meet 1
the personnel qualification requirements of the Program Plan
or the CPSES Quality Assurance Program as applicable.
.
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Revision: 2
Page 6 of 7
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ITEM NUMBER II.a
(Cont'd)
s
4.0. CPRT ACTION Pl.AN (Cont'd)
4.4 Procedures
Calculations and evaluations performed by CPPE Civil
-
*
Engineering and Gibbs & Hill will be performed in accordance
with the procedures normally applicable to those activities
for CPSES. Third-party activities will be conducted in
accordance with applicable CPRT guidelines.
4.5 Standards / Acceptance Criteria
Building Code Requirements for Reinforced Concrete -
ACI-318-71 and stipulations of section 3.8 of FSAR form the
basic standard / acceptance criteria of calculations performed
under this action plan.
4.6 Decision Criteria
' -
If the documentation supporting "use as-is" rebar omission
approvals is found insufficient, the documentation will be
supplemented as required to provide complete justification for
_
prior engineering judgements.
The results and conclusions of analysis / calculations performed
and the evaluation of procedures governing the engineering /
field d' sign change interface and the evaluation of
implementation of these procedures in two areas of the civil
discipline will provide the basis for determination of the
potential applicability to other areas and the need to broaden
the review further.
5.0 SCHEDULE
Cavity Wall Analysis: Complete
Documentation Review: 04/26/85
..
Procedure Review: 05/10/85
Procedure Implementation Review: 05/10/85
Conclusions: 06/01/85
,
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COMANCHE PEAK RESPONSE TEAM
ACTION PLAN
,
Item Number: II.e
,- Title: Rebar in the Fuel Handling Building
Revision No. 0 1 2
Revised to Reflect Incorpora tes
Description. Original Issue NRC Comments SSER
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Pagg 1 of 7
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ITEM NUMBER II.e
.
Rebar in the Fuel Handling Building
1.0 DESCRIPTION OF ISSUE IDENTIFIED BY NRC
*
.
The TRT investigated an alleged instance of unauthorized cutting of
.
reber associated with the installation of the trolley process aisle
rails in the Fuel Handling Building. The claim is that during
installation of 22 metal plates in January 1983, a core drill was
used to drill about 10 holes approximately 9 inches deep. The TRT
reviewed the reinforcement drawings for the Fuel Handling Building
and determined that there were three layers of reinforcing steel in
the top reinforcement layer of the slab. This reinforcement layer
consisted of a No.18 bar running in the east-west direction in the
first and third layers, and a No. 11 bar running in the north-south
direction on the second layer. The review also revealed that the
layout of the reinforcement and the trolley rails was such that the
east-west reinforcement would interfere with the drilling of holes
along only one rail location. However, if 9-inch holes were
drilled, both the first and third layers of No. 18 reinforcement
would be cut. Design Change Authorization No. 7041 was written for
authorization to cut the uppermost No. 18 bar at only one rail
location, but did not reference authorization to cut the lower No.
18 bar. DCA-7041 also stated that the expansion bolts and base
plates may be moved in the east-west direction to avoid
interference with reinforcement running in the north-south
direction. The information, described in DCA-7041, was
substantiated by Gibbs & Hill calculations. If the ten holes were
actually drilled 9 inches deep, then the allegation that the
reinforcement was cut without authorization would he valid.
2.0 ACTION IDENTIFIED BY NRC
[
Accordingly TUEC shall provide: i
-
Information to demonstrate that only the No. 18 reinforcing
steel in the first layer was cut, or
-
Design calculations to demonstrate that structural integrity
is maintained if the No. 18 reinforcing steel on both the
first and third layers was cut.
3.0 BACKGROUND
-
3.1 Information Supplementing NRC Description of Issue
It is alleged that during the drilling of holes for
{* --
installation of the Hilti Kwik bolts for process aisle trolley
rails at El. 810'-6", 10 holes were drilled approximately 9
inches deep. This depth would cut through top and bottom
J
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Revisien: 2
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Page, 2 of 7
ITEM NUMBER II.e
(Cont'd)
3.0 BACKCROUND (Cont'd)
(1st and 3rd layers) of the east-west No. 18 reinforcing steel
.
located in the top reinforcement of the concrete mat. No. 18
:
reinforcing steel also runa longitudinally along the aisle
rails in the east-west direction but without a potential for
interference since the Hiltis and rail do not line up. The
governing design document (DCA #7041) authorized cutting of
only top (1st layer) No. 18 bar at only one rail location. It
is evident from a review of drawings that reinforcement would
be encountered at only one rail location in east-west
direction due to the spacing of rebars running east-west
compared to the spacing of rails and Hilti bolts. The design
required the rail base plates to be so located in east-west
direction to avoid cutting of 2nd layer of No. 11 rebar which
runs in the north-south direction. A field inspection has
verified that at one location in the east-west direction, the *
.
1st and 3rd layers of rebar could have been cut, based upon
-
the size of Hilti bolt installed.
.
3.2 Preliminary Determination of Root Cause and Generic
Implications
-
The preliminary reviews conducted to date have not 1!entified
a root cause for this ites. Implementation of the action plan
is necessary before a root cause determination can be made.
Generic implications will be evaluated based on the results of
the implementation of the action plan and the results of the
root cause determination.
4.0 CPRT ACTION PLAN
4.1 Scope and Methodology
This action plan is designed to assess the as-built condition
of the concrete mat at elevation 810'-6" of the fuel handling
building, to assess the work of the construction crew that
could have cut an additional rebar without proper t
,.
authorization, as well as a review of controls governing rebar
cutting. [
{
Design calculations will be generated to demonstrate that the l-
-
structural design requirements of the concrete mat at
i
elevation 810'-6" will be met assuning a No. 18 bar in the 3rd f
layer is cut.
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Revision: 2
Pag,e ,3 of 7
_
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ITEM NUM3ER II.e
.
(Cont'd)
4.0 CPRT ACTION PLAN (Cont'd)
The adequacy of procedural controls govmdag rebar cutting
-l either at Hilti or core bore installt*. lens will be reviewed.
- These activities will include control ci~ cebar cutting
machines, craft procedures, inspections and proper engineering
authorization.
All the cases from Unit 1 & 2 where rebar cutting was
-
requested for installation of Hilti bolts will be studied. A
determination of the possibility of additional (i.e.
unauthorized) rebar cut will be made based on reinforcement
pattern in slab or wall vis-a-vis embedment depth of the Hilti
bolt installed. Field inspection will be conducted in such
cases to verify actual installed embedded length of Hilti
bolts.
The Hilti bolt installation work performed by the construction
crew that installed the subject 'Hilti' bolts, will be
reviewed over a period of eight months (i.e. timeframe crew
worked together) surrounding the subject installation. This
.
review will include installations where rebar cut
authorization was requested by this crew and a potential
violation of such authorization by this crew while installing
Hilti bolts.
The attached logic diagram identifies tasks and the inter-
relationship of taska for resolution of this action plan.
- 4.2 Participants Roles and Responsibilities
! , i The organizations and personnel that will participate in this
effort
work.
are described below with their respective scopes of
4.2.1 Comanche Peak Project Civil Engineering
i
4.2.1.1 Scope
.
l
-
will perform design calculations
l documenting the adequacy of the I
elevation 810'-6" mat
l
-
will review controls for cutting
rebar i'
i l
-
will evaluate work of subject I
!
( construction crew and selected
_
review of other rebar cut situations
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-
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ITD( NUMBER II.~e
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4.0 CPRT ACTION Pl.AN (Cont'd) .,
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will; assist' tai overall? engineering
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3. - Action Plan Results Raporti
4.2.1.2 Personnel
Mr. C. R. Hooton Project Civil Engineer
Mr. D. G. Patankar Civil / Structural Lead
Engineer
.
. -
,
Mr. S. A. Raz Structural Engineer
4.2.2 Cibbs & Hill - Site Design Review Team
.
4.2.2.1 Scope
-
site design Review Team will design
review calculations performed by
4-
CPPE Civil Engineering
i
.
'
6 4.2.2.2 f
} Personnel .-
v
Mr. B. Wilcoxson Design Review Group
Supervisor
Mr. B. K. Bhujang Structural Group Lead
Mr. k. P. Shah Principal Engineer
4.2.3 Third-Party Activities ',
4.2.3.1 Scope ,
t
. .
-
overview inspection to determine. j
actual lengths of Hilti boltsfused. I
where rebar cutting was required for {
Hilti bolt installation *
'
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-
review of analysis / calculations to
verify adequacy of elevation 810'-6"
mat
-
review of procedural controls for
rebar cutting
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(Cont'd)
, , , ?n 4.0 CPRT ACTION PLAN (Cont'd) *
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verification 'of,isplementation of"
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(; installation by. subject crew and
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other reber cut ofcuations-
-
l overall evaluation of resolution of
action plan
-
preparation of Results Report.
'
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4.2.3.2 Personnel
l .
Mr. H. A. Levin TERA, CPRT Civil /
. Structural Review Team ,,
Leader
Dr. C. Mortgat TERA, Senior Structural
,
Engineer
Dr. J. Arros TERA.' Structural Engineer
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'
'Mr. G. Lagleder^ ^
Soutdures""t Researc' h i
Institute
4.3 Personnel Qualification Requirements
Where inspections require the use of certified inspectors,
qualification will be to the requirements of ANSI N45.2.6 at
the appropriate level. CPSES personnel will be qualified in
accordance with applicable project requirements. Third-party
inspectors will be certified to the requirements of the third-
party employer's quality assurance program and specifically
trained to the requirements of the CPSES quality procedures.
. ,
Other participants will be qualified to the requirements'of
,
'
the CPSES Quality Assurance Program or to the specific
requirements of the Program Plan. '
4;
4.4 Procedures '
Calculations and evaluations performed by CPPE Civil
Engineering and Gibbs & Hill will be performed in accordance
with the procedures normally applicable to those activities
for CPSES. Third-party verification activities will be
, conducted in accordance with applicable CPRT guidelines. A
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procedure will be developed:for inspection lof Hilti bolt ..
installations where* embedded len
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(Cont'd)
4.0 CPRT ACTION PLANI (Cont'd),
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p[~.c 4'.5 " Standards
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/Adc,otence * Criteria,
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* ACI-318-71 Building Code Requirements For Reinforced. Concrete
and stipulations of FSAR Section 3.8 formed the basic
standards and acceptance criteria for original design of
concrete sat at El. 810'-6" in Puel Handling Building which
are documented in Calculation Book No. SFB 102C Section 1.
The design calculations generated with this action plan will
be consistent with the original design criteria.
'
4.6 Decision Criteria
-
The results and conclusions of analysis / calculations performed
as well as conclusions drawn from:
-
review of procedural controls on rebar cutting
activity,
t
-
review of situations where rebar cutting waa' necessary
for Hilti bo2t installation, and
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review of work performed by subject construction crew
.
.
will determine the acceptability of the as-built condition of
the subject
other areas. case as well as the potential applicability to
Any other unauthorized rebar cuts will be
evaluated.
Procedural controls will be revised to strengthen
rebar cutting operations if determined appropriate.
5.0 SCHEDULE
Mat Analysis: i
Complete i
j
(
Procedure Review Complete
l
Procedure Revision: *
!
05/06/85 i
Documentation Review:
.
.
04/26/85
Inspections:
05/06/85
Engineering Evaluation (as required): 05/06/85
Results Report: I
06/01/85 ,
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COMANCHE PEAK RESPONSE TEAM
ACTION PLAN
Iten Number: II.d
,
.
Title: Seismic Design of Control Room Ceiling Elements
.
Revision No. 0 1 2
Description-
Revised to Refleci. Incorporates
Original Issue NRC Consnents SSER
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Prepared and
Recommended by: 7
Review Team Leader
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Revision: 2
Page i of 18
( .
ITEM NUMBER II.d
Seismic Design of Control Room Ceiling Elements
1.0 DESCRIPTION OF ISSUE IDENTIFIED BY NRC
The TRT investigated the seismic design of the ceiling elements
l installed in the control roca. The following matrix designates
those ceiling elements present in the control room and their
seismic designation:
1. Heating, Ventilating and Air '-
Conditioning - Seismic Category 1
2. Safety-Ralated Conduits ! - Seismic Category I
3. Nonsafety-Related Conduits - Seismic Category II
4. Lighting Fixtures '
- Seismic Category II
5. Sloping Suspended Drywall Ceiling - Non-Seismic
6. Acoustical Suspended Ceiling - Non-Seismic
7. Louvered Suspended Ceiling - Non-Seismic
According to Regulatory Guide 1.29 and FSAR Section 3.7B.2.8, the
seismic Category II and non-seise.ic items should be designed in
such a way that their failure would not adversely affect the
functions for safety-related components or cause injury to
operators. *
*
For the non-seismic items (other than the sloping suspended drywall
ceiling), and for nonsafety-related conduits whose diameter is 2
inches or less, the TRT could find no evidence that the possible
effects of a failure of these items had been considered. In
addition, the TRT determined that calculations for seismic Category
I
II components (e.g., lighting fixtures) and the calculations for
the sloping suspended drywall ceiling did not adequately reflect
the rotational interaction with the non-seismic items, nor were the
fundamental frequencies of the supported masses determined to
assess the influence of the seismic response spectrum at the
control room ceiling elevation would have on the seismic response
of the ceiling elements.
l 2.0 ACTION IDENTIFIED BY NRC i I Accordingly. TUEC shall provide
O The results of seismic analysis which demonstrates that the
(, ] l
-
yw non-seismic items in the control room (other than the sloping
suspended drywall ceiling) satisfy the provisions of
Regulatory Guide 1.29 and FSAR Section 3.75.2.8.
(-
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.
-
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Revision: 2
Page 2 of 18
.
ITEM NUMBER II.d
(Cont'd) *
2.0 ACTION IDEhTIFIED BY NRC (Cont'd)
-
d An evaluation of seismic design adequacy of support systems
for the lighting fixtures (seismic Category II) and the
\, CO"{. *
,
suspended drywall ceiling (non-seismic item with modification)
_
.
which accounts for pertinent floor response characteristics of
the systems.
-
Verification that those items in the control room ceiling not
installed in accordance with the requirements of Regulatory
Guide 1.29 satisfy applicable design requirements.
-
The results of an analysis that justify the adequacy of the
non-safety related conduit support system in the control room
for conduit whose diameter is 2 inches or less.
# ' .
-
The results of an anaJysis which demonstrate that the
foregoing' problems are not applicable to other Category II and
~ non-seismic structures, systems and components elsewhere in
'
-
the plant.
.
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3.0 BACKCROUND
_
3.1 Discussion of Specific Technical Issue
Regulatory Guide 1.29 states: "Ihose portions of structures,
systems, or components whose continued function is not
required but whose failure could reduce the functioning of any
plant feature included in Items 1.a through 1.q above to an
unacceptable safety level should be designed and constructed
so that the SSE would not cause such failure." Specifically
Item 1.n states: "The control room, including its associated
vital equipment, cooling systems for vital equipment, and life
support systems, and any structures or equipment inside or
outside of the control room whose failure could result in
incapacitating injury to the occupants of the control room."
The specific issue involves the ability of the control room -
ceiling and other non-safety related ceiling elements, i.e.
conduite and lighting fixtures to remain in place during a
seismic event thus avoiding the potential of disabling
operators due to its failure. Portions of the ceiling are
non-seismic, non-safety related and do not have provisions of
seismic Category II installations which are seismically
supported or restrained as described below.
In reviewing the design of the control room ceiling the TRT i
,
has requested that anslyses be provided which demonstrate that
, the provisions of Regulatory Guide 1.29 and FSAR section
3.7B.2.8 have been satisfied for all non-safety related items. ;
i
.
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Revision: 2
Page 3 of 18
ITEM NUMBER II.d
. , (Cont'd)
3.0 BACKGROUND (Cont'd)
This request encompasses the architectural ceiling system and
non-safety related conduit whoce diameter is two (2) inches or
, less.
.
The TRT has requested that seismic calculations for the
support systems of the lighting fixtures and the suspended
gypsum ceiling in the control room reflect all loading
conditions that would be experienced due to a seismic event.
In addition, an analysis has been requested by the TRT to show
' that the present design of the attachment of gypsum to its
frame will ensure separation will not occur during a seismic
event.
3.2 Description of Existing Ceiling and Other Ceiling Elements
1
The control room ceiling at the location of the control board
area proper is comprised of three (3) ceiling systems. Refer
i to isometric sketch II.d-1. '
( The ceiling system directly adjacent to the control boards at
,
elevation 839'-6" has suspended louvered panels with exposed
grid utilizing interlocking main and cross tees for panel
.
support. The louvered ceiling is directly below the lighting
fixtures above the control boards and is supported by 12
gauge, minimum, cold drawn wire attached to the seismically
restrained unistrut lighting support grid l'-0" above. The
2'-0" x 4'-0" louvered ceiling panels are supported by the
main and cross tee grid and by closure strips at edge
intersections. Above the louvered ceiling, at the location of
the lights is mineral acoustical tile which rest on the light
fixture flanges with tee sections installed perpendicular to
the light fixtures for end of tile support.
The second ceiling system, a sloping gypsua vall, is near the
center of the control board area and extends from elevation
839'-6" to the underside of the above floor. This sloping
gypsum wall was originally constructed as non-seismic and
non-safety related. The construction used 1 " supporting
channels attached to the underside of elevation 854'-4" floor
slab with a 16 gauge C channel secured with 2-3/8" 6 Hilti
Kwik Bolts. The vertical 1 " channel supports for attaching
the horizontal furring channels is attached to the supporting
channels at the bottom of the wall. The upper attachment for
the 1 " vertical channel is by a bolted connection to a 16
gauge C channel secured to the underside of elevation 854'-4"
r
floor slab with 2-3/8" 6 Hilti Kwik Bolts. The bolted -
(, connections used k" 9 bolts. The 1 " channels are by U. S.
Cypsum, constructed of cold rolled 16 gauge steel with 19/32"
.
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._ . __ _ _ .__ _ __ .
. _ _ . _ _ _ _ . __ _ _ _ . _ _
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Revision: 2
Page 4 of 18
ITEN NUMBER II.d'
(Cont'd)
3.0 BACKGROUND (Cont'd)
flange and 1 " depth. Horizontal " hat shaped" furring
channels are placed on l'-0" C.to C for attaching the 3/8"
.
'
-
gypsum panels. The horizontal furring channels are secured to
{ the support channels with galvanized furring channel clips.
i
A review of the mass involved in the sloping gypsum well l
determined that seismic restraint was necessary to assure !
integrity of the ceiling system during a seismic event. To
provide assurance that the sloping gypsum well framework would
remain in place during an SSE, restraints were added by
attaching stainless steel cable through alternating horisontal
furring channels next to each vertical 1 " channel. The six
1/8" stainless steel cables are suspended from two (three
each) angles which are anchored to the underside of elevation
854'-4" floor slab. In addition the furring channel
,
attachment to the vertical channels was reinforced by adding
i
2-k" self tapping sheet metal screws at each intersection.
,
The 3/8" gypsum board is attached to the sloping wall furring i
;
\
channels with 1" type S bugle head screws on 12" vertical and
,
7" horizontal spacing. Screw attachments, of this type, in -
gypsum panels meeting ASTM C-36 requirements typically exhibit
60 pound pull out each.
The sloping gypsum wall was evaluated for acceptability and
compliance with Regulatory Guide 1.29. Based on the original
design and addition of seismic restraint cables, and
considering the quantity and pull out strength of she screw
<
'
attachments securing the gypsum panels, the sloping gypsum
> ! wall was considered acceptable as restrained.
The third ceiling system is at elevation 847'-2" in the center
of the control board area. This ceiling is again the louvered
ceiling configuration which is supported by suspending the
main tees from the seismically restrained unistrue lighting
support grid above. Suspension of the main tees is by 12
i gauge, minimum, cold drawn wire. The 2'-0"x4'-0" louvered i_ '
ceiling panels are supported by the main and cross tees grid
and closure strips at sloping wall intersection. ,
All lighting fixtures in the control room complex are l
seismically restrained in accordance with the restraint
i i
details shown on Gibbs & Hill drawing 2323-El-1704-01. The
lighting fixtures located in the control board ares proper are
also attached to seismically restrained unistrut grid
framework. The seismic restraint of these fixtures and
;
O-
members was deemed necassary to ensure compliance with the
requirements of Regulatory: Guide 1.29.
'
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Revision: 2
Page 5 of 18
.
( ITEM NUMBER II.d
(Cont'd)
-
3.0 BACKGROUND (Cont'd)
3.3 Approach Selected
*
.
TUEC has reviewed the issue raised by the TRT, and a decision
has been made to modify the control room ceiling in the
interest of an expeditious resolution. A recent assessment of '
the ceiling design indicates that the structural integrity of
the three systems could be demonstrated, but the process would
be time consuming. Details of such evaluations involve
modeling assumptions or test configurations for which it is
'
anticipated that technical consensus would be required. Since
' such details are not likely to be discussed in published '
literature or regulatory consunications the resolution could
delay closure of the issue. The approach taken to the ceiling
design modification is to establish a design which can readily
be qualified seismically and to subject the design to a third-
party review.
Action Plan Ites Number I.c addresses the other specific
technical issue, 2 inch and smaller non-seismic conduit.
N
In addition to the specific technical issues there were
questions raised by the NRC regarding the general methods
utilized for addressing potential damage to safety-related
items caused by non-Category I items. This aspect was
discussed during the October 23,'1984, public meeting
regarding the plan for responding to TRT issues. Based on
that meeting CPRT has elected to conduct a third-party design
verification of the damage study program to address generic
implications of the two specific technical issues previously
identified. The verification effort will include an
assessment of the control room lighting fixtures and all other
4
plant architectural features, for example non-seismic
partitions, other architectural ceilings, doors, etc.
4.0 CPRT ACTION PLAN
4.1 Scope and Methodology
The resolution of the control room ceiling seismic design
issue will be addressed as three related but distinct tasks:
-
Seismic interaction design of control room ceiling
elements
-
Verification of seismic damage assumptions for 2 inch
h,a and smaller conduit
.
-
- , . - - , - - - , - , - - . - - . , ,- .. , ,- n. - ,, , , - . - .
. . . - _ ___ _ _ _ _ _ ._ ._ __ _ -
_ __ _- _
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Revision: 2
Page 6 of 18
<
ITEM NUMBER II.d
(Cont'd)
4.0 CPRT' ACTION PLAN (Cont'd) t
i '
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Verification of seismic damage study methods and
implementation, including non-seismic architectural
,
.
features review
l ,
The first task is restricted to a technical resolution of the
specific issue regarding seismic interaction from control room
ceiling as identified by the TRT. The second task is
incorporated in its entirety within Action Plan Item I.c. The
third task is being undertaken to evaluate generic
' implications of the issues raised in the first two tasks, and :
to independently verify the adequacy of the resolutions. '
' Design changes, based on necessity or expediency, might result , ,
to address results of the verificatior effort. !
4.1.1 Ceiling Element Design
1 -
The present design of the ceiling systems was based on 1
the premise that failure of architectural features with .
small masses would not be adverse to the occupants of
4
#
l
\ the control room. Although a preliminary design
'
-
assessment supports the position that the design
complies with Regulatory Guide 1.29, that assessment
i
relies on that same judgement. In lieu of efforts
directed at developing confirmatory analysis to further
support the position and to finalize the structural
evaluation, TUEC has elected to revise the design of-
~
the ceiling systems to preclude unacceptable seismic
interaction. Figure II.d-1 is a flow diagram-for
activities associated with the design effort. The-
,
major steps are discussed in the following-
subparagraphs.
4.1.1.1 Any unacceptable interactions will either be
eliminated (e.g. by moving the safety related
, item) or prevented by restraining the motion ,
of the ceiling structures through the use of
horizontal seismic restraints on the ceiling
support system. These horizontal restraints
will act together with the e'xisting vertical
cable seismic restraints. The design
criteria will incorporate displacement
i
limitations based on an interaction
assessment of the ceiling and safety related
items.
' f The existing ceiling structures, consisting
( of an interlocking grid cf unistruts, dead
.
'
weight supporta and vertical seismic
. ,
9
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,
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,
Revision: 2
Page 7 of 18
ITEN NUMBER II.d
(Cont d)
.
4.0 CPRT ACTION PLAN (Cont'd)
restraints, together with the new horizontal
seismic restraints will be evaluated and
*
.
augmented where necessary to confirm their
adequacy to carry loads from any contributing
structural and architectural elements.
4.1.1.2 TUEC has elected to remove the gypsum panel
and replace it with lightweight metal panels
included in the evaluation below.
4.1.1.3 An evaluation of the acoustical and louvered
ceiling panels will be performed to
demonstrate that the physical arrangement of
these modular lightweight architectural
features meet the requirements of Regulatory
Guide 1.29 and FSAR Section 3.78.2.8. This
evaluation will either confirm that
individual components will not fail due to
seismic interactions or the design will be
modified to incorporate seismic restraints
which prevent the component from falling.
,
4.1.1.4 Items which are designed and constructed to
seismic Category I and seismic Category II
criteria receive Quality Control inspection
in accordance with applic.able criteria
established per 10CFR50 Appendix B. Items
which had been non-seismic did not require
inspection by the Quality Control
Organization. To comply with the TRT request
for verification that control room elements
comply with the design additional QC will be
performed as follows:
-
Existing Category II QC records will
be reviewed to verify required
inspections were performed
-
All design modifications to
Category II ceiling elements will
receive a QC inspection with
Engineering inspection and overview
for other items (e.g., if- restraints
of lowered panels are required.)
(,,
-
The third-party will review
,
inspection documentation.
a e
_ _ _ _ . _ _ _ _ _ . _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ _
*
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Revision: 2
Page 8 of 18
.
( -
ITEM NUMBER II.d
(Cont'd)
4.0 CPRT ACTION plt.N (Cont'd)
4.1.1.5 All design and analysis activities associated
with the evaluation of seismic interaction of
-
.
architectural features above the control room
and with the ceiling design modifications
will be subjected to a third-party review.
This will include an independent verification
of the methods used to evaluate displacement
interactions, failure of architectural
features, interaction consequences, and a
design verification of all ceiling
modifications to confirm the seismic
capability,
i.1.2 Conduit of Diameter Less than or Equal to 2 Inches
See 1.c Action Plan.
4.1.3 Damage Study Verification
The third task of the II d Action Plan to resolve the
- TRT issues is directed at a determination of the extent
to which winilar aspects of the Unit I damage study
program could require technical resolution.
Additionally if technical issues develop, the. plan
includes a course of action to determine the
applicability to Unit 2 and resolve them for both
units.
,
( The seismic /non-seismic interaction study, which was l
performed, in 1983, involved the walkdown of 287 rooms.
All potential interactions were evaluated to the
acceptance criteria developed for the study. Methods
for resolution of potential interactions of a falling
source impacting a nuclear safety class target
consisted of analysis, evaluation, use of barriers.
l ,
administrative controls, addition of seismic supports
or restraints. Each of these activities includes
l pertinent requirements of the CPSES QA Program. !
The design of the ceilings in the control room was
predicated on the position that failure of
i
architectural features with small masses would not
adversely affect the occupants of the control room and,
j consequently, the safety of the plant. On this basis,
Engineering advised the Damage Study Group that the
, !. control room architectural features in question should !
,- not be evaluated as part of the Damage Study Program.
!
Therefore, these features were not evaluated by the
,
i
*
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Revision: 2
Page 9 of 18
ITEM NUMBER II.d
(Cont'd)
4.0 CPRT ACTION PLAN (Cont'd)
Damage Study Group. Exclusion of other areas of
evaluation due to similar inputs and assumptions is a
,
.
focal point for the Damage Study verification.
The activities in this phase include a project review
of all plant architectural features, and design
modification, if required, to resolve resulting *
concerna. The preliminary TUEC assessment of the
entire damage study area indicates that if omissions
exist they would most likely be in that aspect of the
program. Also included in the action plan is an
extensive third-party evaluation of all aspects of the
Unit 1 Damage Study Program. The major steps are
discussed in the following subparsgraphs. Refer to
Figure II.d-2 which is a flow chart of activities.
4.1.3.1 Architectural features e.g., doors, ceramic
tile, gypsum board, etc., were not viewed as
structural in nature therefore considerations
(', such as seismic qualifications may
inadvertently have.been omitted.
A TUEC evaluation will be performed on
architectural specifications and drawings tc
identify non-seismic sources to be evaluated
in accordance with Regulatory Guide 1.29 and
FSAR Section 3.7B.2.8. Any architectural
features in Units 1 and 2 which are
determined to have a potential for seismic
interaction will be either modified or
subjected to a damage study assessment.
4.1.3.2 A third-party review of the architectural
featu.res evaluation will be performed
including verification of:
.
-
The entire structural assessment
which is performed by CPPE-TUCCO to
determine which architectural
features have a potential for
seismic interaction.
-
The damage study of all
architectural features which
,
potentially cause seismic
interactions.
'
..
.
^
.-r-,_
. - . . - - . . . . -
'
'
.
,
Revision: 2
Page 10 of 18
ITEM NUMBER II.d
(Cont'd)
4.0 CPRT ACTION Pl.AN (Cont'd)
4 -
A design review of all resolutions
to unacceptable interactions
, including review of all drawing
-
revisions which result.
4.1.3.3 Procedures and methods for conducting the
seismic damage study will be evaluated by a
third-party including a ~ review of interfaces
between the damage study group and other
disciplines which either provide input for
, ~
the evaluation or act upon the damage study
group's recommendations. This evaluation
will examine whether there are other generic
areas in addition to the architectural
features which may have been omitted.
- '
4.1.3.4 To identify the set of all non-Category I
. systems, structures and components which are
,
'
considered for seismic interaction with
- ~'
safety-related items, selection criteria have
'a
been established. The exclusion of Category
-
II items, is one of the selection criteria
. because they are supported or restrained to
preclude seismic interaction. Design
,
criteria for Category II items and all other
selection criteria will be subjected to
third-party review. The detailed Category II
criteria will be compared with Category I
criteria and methodology as well as the more
general Category II FSAR coc:mitments. Where
- the Category II criteria is not the same as
Category I or is less detailed the
acceptability will be assessed based on
application of engineering principals
j applicable to the specific methodology.
4.1.3.5 The basis for determining specific physical
i
interactions for displacements of items
identified in Section 4.1.3.1 will be
subjected to third-party review.
4.1.3.6 Criteria for evaluation of the consequences
of interactions will be subjected to third-
party review.
i
o
6
.
4 4
w v. + --.. .-- , - - , +.. e-. . . % ,m.,--., , . , , , . -.,---.m,e. #w.-.,-,s,w- ,,n,wy,-.-.y .. , , .m,y-r,,- ---x-, -,-,- .- , - - - . . , - - . - , - . . - - -
-
..
Revision: 2
Page 11 of 18
ITEM NUMBER II.d
(Cont'd)
4.0 CPRT ACTION PLAN (Cont'd)
4.1.3.7 . Implementation of-the methods for performing
the damage study will be design reviewed in a
-
,
third-party engineering audit of the
documentation, including as necessary .
physical verificatica. The engineering audit, !
will assess adequacy of documentation and
validity of engineering conclusions.
Included within this task is a review of the
way in which classified and declassified
rooms were distinguished and addressed.
The basis for selection of implementing
documents reviewed will be those which are
most substantially supported by engineering
judgements as distinct from simple
calculation or direct application of rules.
The selection process will be thoroughly
documented. Based on a complete review of
the various methods for categorizing
interactions as acceptable, cases will be
t
( selected to check the implementation of each
-
method. This checking of implementation
will be weighted towards interaction
resolutions which are, of necessity, less
L
subject to procedural definition and more
dependent on individual judgements.
li 4.1.3.8 A third-party comparative damage assessment
will be performed in the selected rooms to
assess the consistency with which
interactions were identified. The rooms will
be chosen based on a selection criteria which
.
consider the following aspects:
-
Functional and physical aspects of
the systems and components within
the space
l -
Importance of components and systems
to safety
-
Representativeness of the room from
the standpoint of physical proximity
of potentially interactive items
-
Balance of types of non-seismic
items, i.e. architectural, conduit
! - 'y - piping, equipment
_ -. , . . _ . - . . - - . . - . - - . .- -
._...-,-..-.---..--.-,._-_.O
_-_ - - _ _ _ _ _ _ - _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ . _ _ _ _ . -- -_ _ _--_ .
'
.-
Revision: 2
Page 1.2 of 18
ITEM NUMBER II.d
(Cont'd)
4.0 CPRT ACTION PLAN (Cont'd)
4.2 Procedures To Be Used
-
,
In addition to the applicable analysis, design, construction
and QC procedures which will be used to implement the control
room ceiling modifications, precedures directly applicable to
the damage study will be used. 71ey are CP-EI-4.0-53,
" Maintenance of Damage Study Analysis" and CP-EI-4.0-63,
" Review of Architectural Specifications and Drawings to
Identify Non-Seismic Sources".
4.3 Participants Roles and Responsibilities
.
The following organizations and personnel will participate in
this effort:
4.3.1 Comanche Peak Project Engineering and Gibbs & Hill
4.3.1.1 Scope
-
perform architectural features
(L' evaluation (4.1.3.1)
l
-
recommend and implement design
modifications resulting from
architectural review or third-party
verification of damage study program
-
design (as required) structural
horizontal restraint system for
ceiling elements (4.1.1.1)
-
redesign sloped ceiling panel system
(4.1.1.2)
-
evaluate and modify, as required,
acoustical and louvered ceiling
panel restraints (4.1.1.3)
-
assist in preparation of Results
Report
4.3.1.2 Personnel
Mr. C. R. Hooton Project Civil Engineer
Mr. M. Wells
. -
Engineering Specialist
.
Mr. D. A. West Damage Study Engineer
.
. _.________m______
. _ - - - - _
,. ..
.
.
Revision: 2
Page 13 of 18
= l
s ITEM NUMBER II.d l
(Cont'd) i
4.0 CPRT ACTION PLAN (Cont'd)
Mr. J. Eichler Gibbs & Hill, Manager of
Civil / Structural
,- Departnent
Mr. E. Berkor Gibbs & Hill, Structural
Job Engineer
Mr. M. Pope Gibbs & Hill, Structural
Engineer
4.3.2 Brown & Root. Inc.
4.3.2.1 Scope
-
install ceiling system supports and
restraints
-
provide as-built data on
architectural features
d
4.3.3 TUCCO Quality Assurance
(
- 4.3.3.1 Scope
-
inspect control room ceiling
modifications
-
inspect other modifications
resulting from damage study
interaction resolution
4.3.3.2 Personnel
Mr. P. Halstead Quality Engineering
Supervisor (Acting)
4.3.4 Third-Party Activities
4.3.4.1 Scope
- review of control room ceiling
analysis and design (4.1.1.5)
-
review of architectural features
evaluation (4.1.3.2)
(
,
, , -~ _ , , , , - - - - - - - , - - ,,,.,r--, -e , , - - -
, . r,
___ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _
'
a: -
Revision: 2
Page 14 of 18
ITDi NUMBER II.d
(Cont'd)
4.0 CPRT ACTION PLAN (Cont'd)
.
-
review seismic damage study
procedures and criteria (4.1.3.3,
.
,
4.1.3.4. 4.1.3.5 and 4.1.3.6)
-
Design review implementation of
damage study procedures (4.1.3.7)
-
perform independent damage study
, walkdown (4.1.3.8)
-
prepare Results Report
4.3.4.2 Personnel
Mr. H. A. Levin TERA CPRT Civil /
Structural Review Team
Leader
Mr. D. Witt TERA Senior Mechanical
Engineer
.
Mr. P. Streeter TERA Senior Mechanical
>
Engineer
4.4 Standards / Acceptance Criteria
.
(
[ The FSAR will be used as the base document for establishing
- acceptance criteria. Where the criteria in the FSAR requires
i increased detail to define specific requirements not covered
( by the standards and regulatory guidance referenced in the
FSAR, the basis will be established engineering principles
--
applicable to the same or similar situations.
'
4.5 Decision Criteria
The control room ceiling design will be modified if it is
determined that its behavior during a seismic event may lead
to unacceptable interactions with safety related items or have
l-
a potential of interfering with operators in the control room.
The Damage Study / Architectural Features evaluation will
-
l l
establish whether or not the plant design conforms with
requirements of Regulatory Guide 1.29 and FSAR section
3.7B.2.8. Any identified deficiencies will be evaluated and
corrected as necessary to meet these commitments.
(
.
.
_ - - - - _ _ - . _ _ - - - _ _ -
l
.
,. ..
f
Revision: 2
Page 15 of 18
.
ITEM NUMBER II.d
(Cont'd)
5.0 SCHEDULE
Ceiling Analysis and Redesign: 06/08/85
. Verification of Design: 06/08/85
Architectural Features Review: 06/01/85 -
Third-Party Review of Architectural Features: 07/01/85
'
Damage Study Criteria and Procedures Verification: 05/15/85
Design Review of Damage Study Implementation: 05/30/85
Independent Damage Study Walkdown 05/30/85
Resolve Damage Study Findings: 06/26/85
Review Resolution of Findings: 07/12/85
Results Report: 08/12/85
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*I .
,
*
,
_ COMANCHE PEAK RESPONSE TE.'M
ACTION PLAN
>- Item Number: VII.b.2
,
.
Title: Valve Disassembly
.
Revision No. 0 ..
,
Description Original Issue
-
5 Prepared and
Recommended by:
g
Review Team Leader
Date 6'hf'EC .
-
- -
-
...
,
Approved by:
Senior Review Team h 4). 1
Date (o f1,1 fy(
;
: FotA-W 36 A\\\
. .. . . . ..
~
.
...
, ,
,
.
,
Revision: O
Page 1 of 9
,
'
r~sg
,
k' ITDI NUMBER VII.b.2
.
Valve Disassembly.
>
1.0 DESCRIPTION OF ISSUE IDENTIFIED BY NRC
"The TRT found that installation of certain butt-welded valves in
. three systems required removal of the valve bonnets and internals
-
prior to welding to protect temperature-sensitive parts. The three
systems involved were the spent fuel cooling and cleaning system,
-
the boron recycle system, and the chemical and volume control.
system. This installation process was poorly controlled _in that
disassembled parts were piled in uncuntrolled areas, resulting in
lost, damaged, or interchanged parts. This practice created the
potential for interchanging valve bonnets.and internal parts having :
different pressure and temperature ratings."
_
2.0 ACT' ION IDENTIFIED BY NRC
Evaluate the TRT findings and consider the implications of these
findings on construction quality. "... examination of the potential
safety implications should include, but not be limited to the areas
3
or activities selected by the TRT."
' '
" Address the root cause of each finding and its generic
implications..."
" Address the collective significsnee of these deficiencies..."
" Propose an Action Plan...that will ensure that such problems do.
,
not occur in the future."
'
3.0 BACKGROUND
3.1 Information Supplementing NRC Description of Issue
Other possible reasons for valve disassembly include
hydrotest, flushing, purging and repair, and therefore many
different valve types are potentially affected.
. Additional background information such as valve manufacturers,
types,' sizes, ratings, installation dates, etc. will be
4
obtained as a part of the implementation of this Issue-
Specific Action Plan.
3.2 Preliminary Determination of Root Cause and Generic
Implications
/ The preliminary reviews conducted to date have not identified
, ! '
( a root cause for this item. Implementation of the action plan
is necessary before a root cause determination can be made.
Generic implications will be evaluated based on the results of
the root cause determination.
-
c
4 4
, . - . - - -..
.
.
.. ,
Revision: O
Page 2 ,of 9
.
ITEN NUMBER VII.b.2
(Cont'd)
4.0 CPRT ACTION PLAN
4.1 Scope and Nbthodology
.
The scope of this action plan is to evaluate if valves that
required disassembly were properly reassembled; and, if not,
whether an improperly reassembled valve could result in a code
violation or have a safety consequence. Valves installed in
Units 1, 2 and Common will be considered in this plan. Any
valve which has been disassembled will be included in this
plan, regardless of reason for disassembly or plant system.
The plan consists of two phases. Phase I consists of
reviewing construction and QC procedures and verification of
QC documentation for adequacy of material traceability and
control; and performing analysis to determina the safety
consequences and/or code violations of any valves that were
potentially reassembled incorrectly. Further evaluations will.
1
be made in Phase II to determine root causes of the problem,.
and whether the problem is generic or specific in nature.
'*
The specific methodology is described below (see Flow Chart,
j Attachment 1):
4.1.1 Identify the valves that required disassembly. The
generic valve types will be determined by reviewing the
valve specifications, purchase orders, operation and
installation instruction manuals, construction, test
l and other piping procedures and by vendor contact, as
required. The specific valves in question can then be
j identified from the valve list.
In addition to generic valve types that require
- disassembly, other specific valves disassembled for
!
test and/or repair will be identified by reviewing
! operations travelers. l
A reference data base will be established for the
specific valves identified. The disassembly / reassembly
history will be reviewed. Applicable manufacturer
drawings will also be obtained.
,'
4.1.2 Review applicable procedures, for both construction and
QC, to determine if they provided adequate controls of
materials during valve disassembly and reassembly. In
addition to proper matching of components, the
,_ procedures will be reviewed for their adequacy to
identify and replace parts damaged during the
,/ disassembly, storage and reassembly process.
3
.
*
. .
.
. _ . - _ _ . . . _ . . _ . _ . _ . _ _ _ . _ _ _ . _ . _ _ _ . . _ . . . _ _ _ . . _ _ .
- __ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ - _ _ _ - - - - - - _ - - -
. _ . .
.
.
.
!
Revision: O
Page 3,of 9
,
.
I ITEM NUMBER VII.b.2
..
(Cont'd)
4.0 CPRT ACTION PLAN (Cont'd)
If procedures have changed during the course of
construction the historical file of procedures will be
'
-
reviewed to determine if taproper reassembly were more
likely to occur during a particular time frame. If the
procedures for Units 1, 2 and Common are different,
they will each be evaluated.
4.1.3 In parallel with the procedure review, an analysis will
be made to determine the safety consequences of
improperly assembled valves. The analysis will include
,
' potential failure modes resulting from improper
reassembly of the valves in question. For example,
reassembly using a valve bonnet with a lower service
rating than called for in the design drawings could
result in external leakage. The generic effect on the
system, e.g. loss of system pressure would then be
identified. A sample format for this analysis is shown I
in Attachment 2. I
'
In addition, an evaluation will be made to define
potential code violations which could result from
improperly assembled valves.
s
4.1.4 The conclusion reached based on the procedure and QC
documentation review performed to date is that improper
reassembly was possible due to insufficiently rigorous
procedures or inconsistencies in documentation. A
random sample will be selected, which will be
statistically based at the 95/95 confidence level. The
random sample size will be based on the population size
to be determined in Phase I.
A second set of valves will be selected to provide
greater assurance that the valves were correctly
reassembled. The sample would be randomly selected
from a population of those valves originally identified
in the specific TRT issue and will be statistically
based on the 95/95 confidence level.
4.1.5 Manufacturers drawings and disassembly procedures will
be reviewed and documentation packages will be
assembled for those valves selected in the random
sample. Inspection procedure will be predicated on the
results of this review. If review of the documentation
-
for a specific valve indicates probable improper
reassembly, reinspection will include a verification of
internal parts.
.
4
_ _ _ . _ _ . , _ _ - _ _ _ _ _ . _ . _ - _ - _ . _ _ _ _ _ - - _ - _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -
_- _- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _
.
.
.
.
,
Revision: O
Page 4,of 9
.
ITEM NUMBER VII.b.2
(Cont'd)
4.0 CPRT ACTION PLAN (Cont'd)
4.1.6 Deviations that are found will be evaluated and if
safety significant.or a code violation, will require
'
-
the sample size to be expanded. All safety significant
discrepancies will be evaluated for root causes. In
addition, all valid discrepancies will be reviewed, as
a group, to identify any adverse trends and their root
causes.
TUGCO will be advised of specific hardware
discrepancies, if any, to be entered into the CPSES NCR
L Program for resolution and disposition.
Programmatic implications, if any, will be reported to
i
.the Manager QA/QC Programmatic Issues, and those with
potential impact on other action plans will be
identified and addressed by the affected Issue
l Coordinators.
4.2 Procedures
L
Operations, construction and/or QC procedures now in effect
will be reviewed and if found satisfactory, will be used for
>
disassembly, inspection, reassembly and test as required.
4.3 Responsibilities
f
The organizations and personnel that will participate in this
effort are described below with their respective scopes of
work.
4.3.1 TUGC0 Comanche Peak Project Engineering CPPE
4.3.1.1 Scope
-
Assist the QA/QC Review Team in the
identification and provision of all
necessary specifications, drawings,
procedures and other documentation
necessary for the execution of this
action plan.
-
Assist in determining the physical
location of the valves selected for
inspection.
-
Process NCRs that may be generated
due to this action plan.
*
. .
_ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . - _ . _ _ . - --
-
t
.,,
Revision: O
Page 5,of 9
.
f
ITDI NUMBER VII.b.2
(Cont'd)
.
'
4.0 CPRT ACTION PLAN (Cont'd)
4.3.1.2 Personnel
y Mr. C. Moehlman TUGC0 Coordinator
4.3.2 CPRT-Civil / Structural Review Team
4.3.2.1 Participate in the evaluation for safety
significance.
' 4.3.2.2 Personnel
Mr. H. A. Levin Review Team Leader
4.3.3 Brown & Root Millwright Shop
4.3.3.1 Scope
Disassemble and reassemble valves, as
required, for inspection.
4.3.3.2
'
Personnel
Mr. C. Moehlman TUCCO Coordinator
l
4.3.4 CPRT-QA/QC Review Team
4.3.4.1 Personnel
All activities not identified in 4 3 1. . , 4.3.2
and 4.3.3 above will be the responsibility of
the QA/QC Review Team.
4.3.4.2 Personnel
Mr. M. Obert Issue Coordinator
Mr. J. L. Hansel QA/QC Review Team Leader
4.4 Personnel Qualification Requirements *
Personnel qualification and training will be in accordance
with the CPRT Program Plan. '
4.5 Sampling Plan !
-
The samples will be selected to provide a 95/95 confidence
'
level. '
*
. .
g -
, e - --. _m, .w_. m , ~ .m.., .. , ,.,_e.-_. . . _,.,,,,_y..,__,.._.,,-,-.,.~__-.,,s ~ , . _ . , , , - . ,.~._m, ,ys ,_
.
.
., .
Revision: O
Page 6,of 9
,
.
- ITEM NUMBER VII,b.2
(Cont'd)
4.0 CPRT ACTION PLAN (Cont'd)
-For large populations this means that a minimum random sample
size of 60 is required, with*zero safety significant
'
-
discrepancies in order to accept a population. If one or more
safety significant discrepancies are discovered, the sample
size will be increased as required. If no additional
significant discrepancies are discovered the population will
be accepted.
If more than one safety significant discrepancy is found in
the initial sample, the sample size will be increased up to
100 percent for reinspection.
NOTE - In all cases, reinspections performed for expanded
samples vill look at only the attribute associated with
the safety significant discrepancy.
4.6 Acceptance Critoria
Acceptance criteria will be based upon a review of the
following:
..
4.6.1 Site Q.C. inspection acceptance criteria for valve
$ installations.
4.6.2 A detailed review of specifications, drawings,
referenced codes and standards in order to identify and
verify safety significant attributes and minimum
acceptance criteria necessary to meet design
requirements. Inspection checklists and instructions
will be developed bared on the results of this review.
4.7 Decision Criteria
4.7.1 The action plan will be closed if all the design
requirements are met, otherwise necessary corrective
action will be recommended to meet the design
requirements.
4.7.2 Programmatic and generic implications, if any, will be
identified and additional activities recommended as
required.
.
,,g..,..e, --m.--s - . . - - - - - - - - - ----~----,-'rsv 'T'
-
-
-
-l
Revision: O
Page 7,of 9
.
ITEM NUMBER VII.b.2
(Cont'd)
l
5.0 SCHEDULE ,
i
,
The schedule for implementing the Plan is as follows:
l
5.1 Phase I i
Start: April 8. 1985
Completion: July 5, 1985
Assumptions
(1) SRT approval of the plan by July 1,1985.
5.2 Phase II
Start: July 5, 1985
Completion: October 15, 1985
Assumptions
j (1) Phase I results indicate that a random sampling program
is required.
(2) Valve disassembly is required for inspection.
(3) Satisfactory results are obtained from the first
sample, and therefore an expansion of the sampling
program is not required.
!
r"
:
. .
.
.
4
Revision: O
Page 8 of 9
.,
.
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. . . -
0 ,. . .
Revision! O
Page 9 of 9
.
ITDt NUMBER VII,b.2
ATTACHMENT 2
.
.
.
.
GEntRtc 5ArtTY Cons (OutmCES ANALYSIS
. . .: DESCRIPflon SAFETY !PR($5uRE POTinTIAL .
I
POTENTIAt.
class RATIrt. RfA55teLY Ena0R FAILURE & (FrtCT5
!TT.Grinnell {*
Diaphro p Valve 3 300pst 1. Bonnet asseely from C.S. 1.
. nanu al No fatters. A11 tonnets are 5t. St. =tth internals
valve of same meterials.
3/4 inch
stainless Steel 2. Dennet assemely from 150 pst 2. No failure. The bonnet and disparagn thicknesses are the
va've same for 150 pst and 300 pst valves.
'
! 3. Ioanet assently from non-A5mt 3. a. Potental fatture during a setssic event. Loss of
j
g valve function. Isakage.
i b. Code violation.
ITT.Grinnell
Diasheep valve 2 150 pst 1. Sonnet assently from non.A9E 1. a. Potenital fallare during a seism * * event. Less of
manusi velse function. leekage.
3/4 inca
Carmen Steel 2. Bennet assemely from A5ft III. b. Come violation.
'
,
Class 3 valve
, 2. Code violetton.
TT.Grinnell *
Otapeep Valve ! 2 1. Sonaet assesely from C.S.
nanual
150
valve
!1.
;
No failure. All bonnets are 5t. St. =tth internals of
same materials.
. incn
stainless 5 teel 2. Sonnet assemely from non-A5PE a. Potential failure during a setsetc event. Loss of
' valve ' 2.
function, leatage.
3. Bonnet assesely Asnt Ill. Class 3 3. Code violetton.
valve
.
t
/
.
.
.. . ._
<
~
"
*
..
.
'
VA LVE RE INSPECTION CHECKLIST
ISSUE 3Z[I.b.2
T/.G 110.
'
ITEIA IJO. GErf. SAMPLE 7
j. 76 7;2C . 4/4 SUB. sat-IPl.E D
RIR NO.
/34-5
VALVE MARKillGS:
YES NO
DODY C BONNET ACCESS. @ _-
DODY NO. 6 349 -3/ ~6
BONNETT NO. E34B-26-5
TRA JE AEILITY *
'
ACCEPT] REJECT] DATE: #I/IPS'
/
REtAAf NS: 5/W 79 /So??-8- 30 M[oD6 MM W-
,
4
<,, ./;
INS BY- y Fo/A-E-36 A se
ajPELT * e c ,~Se Se - 'S e< Eo .
,
. ',
'
'.
VA LVE RE INSFECTION CHECKLIST
ISSUE EI.b.2 .
-
TAG NO. ITEM NO. GE'4. SAMPLE 7
MON"Od// 32I SUE. CAMPLE D
RIR NO.
6 834
,
VALVE MARKINGS:
.
YES NO
GODY C BONNET ACCE 55. [] {
BODY NO.
EONNET' NO.
T; A fi./ lL!TY '
i ACCEF ,
u
i REJECT [ ATE:
g -
.
'
REMAGN5:
0edf Of th74108 /E'Sub47d0s MEW Y'
Vey mow
i
'
(ppd
L
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'
IN PECTE,C BY:
l ,, , ~ -- , . .
'
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.
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-
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; -
ISSUE COORD. . .,
,
- ,,
,
'
.
, . , - .
VA LVE RE INSFECTION CHECKLIST
ISSUE E.b.2 .
TAG NO. ITEM NO. GEN. SAMPLE 7
2o3-( 27+ SUE. LAMPLE [
RIR NO.
visb
VALVE MARKINGS *
.
YES NO
BODY 8 BONNET ACCE55. ]]
BODY NO. Y og/ -/3 9/
EONtJET- NO. E M 4P/ C - 9
TL A li.t ?iL!TY :
ACCE F-{: REJEOT] [ ATE: 8I5/85
.
"? '
I
; REMAr.kS : OfN ?S~SS5?/Y 0# c
l
!
i
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7
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AP /
5:
l lN,5F.r~.f' EC
(&Eg/6,BY*x f/Erds- ,,
Inte Supy
/h/'
'&n/sifs'
lSSUE COQRD.
I
.
.
,
.
. .
VA LVE RE INSFECTION CHECKLIST
ISSUE BI.b.2 -
TAG NO. ITEM NO. GE'4. SAMPLE O
/-70/4D . 542 sus. EAMPLE [
.
RIR NO.
/ 3 %~
VALVE MARKINGS: -
.
-
YES NO
BODY $ BONNET ACCESS. 5I 7
a _J .
BODY NO. E 34Q -7-4
EONNETT NO. f34 7 o -/
1
-
T; A .'i A EILITY :
!
AC CE '-{i REJECT] [4TE: 5 8f .
!
REMAGNS: hops On m RYA M MK -VM 73~~M27 -2-72 'S
#iss, .
!
/h
,
j,
-
APP S
INSPECTEC BY"
O,d/dm. E4/4Y
g, [[g'
QMgr
INSR StJPk ISSUE CO'ORD.
-
- .
- , .
VA LVE RE ItsSFECTION CHECKLIST
ISSUE H.6.2 -
TAG NO. ITEM NO. GE'J. SAMPLE D
2PCV-6//T tr7 SUE. EAMPLE V
RIR NO.
2 VV3
VALVE MARKINGS: .
e
YE S NO
BODY 8 BONNET /,CCE 55. [ ]
BODY NO. S 3 22 - 2D
EONNci no. rics,- 9 7
T; A fi. A clL! Y .*
ACCE4-@ REJET,T] [4TE. 8b!#J
l neuAnus: ys ps-/sp7-5 -/ co cove oors -sq
.
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AP -
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9/b_
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.
*N INS'R SUP(. IS S U E C O'O R D. l
' . ,- _
.. , ,
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VA LVE RE INSFECTION CHECKLIST
[SSUE MI.b.2 . *
TAG NO. ITEN NO. GE's. SAN:FLE V
/[M-8//9 .
/28k SUE. EA? IPLE D
RIR NO.
/0//0
VALVE MARKiliGS:
YES NO
-
'
L
ECDY $EONr1ET 200E55.{O :
EOGY NO. _D2A -57 BMB7&t/
ECNt'ET' ho. DVA C 7 ?/0
i
l
:
TPA 21; EIL! Y : '
ACCEP-@ REJECT] EATE- # I '
'
REMARKS: 0 d6 L}]7M 7Arj AQ.5pi4 fg m lsz yi45 .
t
!
'
,
,
.
9bkr/
APPROVALS:
INS E'TEQ BY:
Mfg 7p3hf INSR SUPF ISSUE COO _PD.
ci- oi e - o i REv. 6 - *
,
..
3- .,
.
VA LVE RE INSFECTION CHECKLIST
ISSUE III.b.2 .
TAG NC. ITEM NO. GEN. SAMPLE 7
2' E//d . 887 SUE.5A-IPLE V
RIR NO.
f
//#df
VALVE MARKINGS:
. YES NO
ECDY tBONNET 'CCE55. 'O
, - LJ
'
BODY NO. - 784
, l
ECNtJE HO. _ 38S7
7;A E ' EILi Y :
i
ACOEF-
REJECT] EATE: @b85'
REMARKS:
fgj pgp y Qg ggpyf
.
i. -
.
APPROVALS:
1 J /7 8 ef ~
INSR SUPP: ISSUE COORD.
CI- Ol e - O f RF V. O - - - - - - . . - _ _ - - - - - -
..,- ?. . -
'
-
. . .-
!_
l,, -
Page 1 of 2
.
.
t
A TION RESEARCH CORPORATION
EVYUdNTROLLE D COPY
CONTROL No ?L Q-
COMANCHE PEAK RESPONSE TEAM
<
.'
QUALITY INSTRUCTION FOR ISS11E SPECIFIC ACTION PLAN VII.b.2
CHANGE NOTICE 001
INSTRUCTION NO: QI-018
REVISION: 0
ISSUE DATE: 09/19/85
i
(.
,
REINSPECTION OF PREVIOUSLY DISASSEMBLED VALVES
The QI identified above is hereby changed as
shown on the attached pages of this notice.
.. -
Prepared by: -/[/[CJ Date: '7 / P *
_ / /
t
Approved by:
Date: 9 f [
Issue Coordinator '
I
Approved by: .
Date: 7-/9- 6
On-Site QA 1epresentative
) Approved by:
Qy..vI..,..... der
Date:
f. f. 8 7
_
.
;
\
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Date :- 8//b/25 Pa a e / of /
v Re ;n e,oee t;o a
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-
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*E$ Data. Base r&specdion Repor6 F 0 Z A /lz 3,/gg,
:;n: uu. CPRT (ERC) Checknst/Wlafvig
??? %u- Accessibillig ikIkdowa eh e eklis t
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---________ ______ ________-_______ _______-__ _______________________________ ________-________ ___________________
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of the hearings. However, any proposed set.edule should be keyed to the
tining and adequacy of responses to discovery requests; only in this way
n*
will it be possible for the Board to fulfill its promise to CASE that
all parties (as memorialized in the Board's 6/29/84 Memorandum and
Order (Written-Filing Decisions, #1: Some AWS/ASME Issues)).
-
1. The Board should invite the Applicants to respond to CASE's .
A - rs to Applicants' seventeen various Motions for Summary
Disposition by a date certain, which CASE suggests should be
September 4, 1985 g /.
3. The Board should also itivite the NRC Staff to respond to CASE's
} responses to Applicants' Motions for Summsary Disposi, tion. Since
, the Staff has not yet responded (with one exception, to our
~-
, recollection) to any of Applicants' Motions oe-CASE's responsesf ~
it might be appropriate for the Board to allow the Stinff/
additional time, which CASE suggests should be by September 16,
1985.
/2f2 CASE considers this amount of time to be very generous, in;1ight of the
many months which have already passed without Applicants responding.
Further, additional time should not be necessary since Applicants
apparently have already been working on this; they have sta,ted that
they are " currently committed to provide the Board with corrections to
all affidavits." See Applicants' 7/5/85 First Partial Response to Ripe
Discovery Raquests, page 5, Footnote 2.
2
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4 Since Applicants are now apparently attempting to phase out
Cygna's independent protocol-controlled effort with the CPRT Plan
?I
which is governed by no such restrictions (and therefore qualifies
only as yet another Applicant-dependent effort), CASE suggests
that the Board request that Cygna Energy Services also respond to
the Applicants' Motions for Summary Disposition and CASE's
responses to them. CASE suggests that the same time be allowed
Cygna as is allowed the NRC Staff, until September 16, 1985.
5. The Board should invite Applicants to respond to CASE's First,
Third, and Fourth Motions for Summary Disposition (filed 10/6/84,
.
11/2/84, and 1/14/85, respectively) /3,/. We suggest that the date
for the filing of such response should be September 4, 1985.
6. The Board should invite the NRC Staf f to respond to CASE's First,
Third, and Fourth Motions for Summary Disposition. We suggest
,
that the date for the filing of such response should be September ~
~
16, 1985.
7. The Fos?d should invite Cygna Energy Services to respond to CASE's
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First , Third, and Fourth Motions for Summary Disposition f4,/. We
suggest that the date for the filing of such response should be
Septembe r 16, 1985.
/3/ CASE's 10/22/84 Second Motion for Summary Disposition was covered by
the Board's 10/26/84 Memorandum (Intent to Retain Academic Expert).
See also the Board's 11/9/84 Memorandum (Testiomony from Dr. Arthur P.
Boresi).
-
/4/ We note that Cygna " responded" to CASE's Third Motion for Summary
Disposition by basically stating that they didn't want to answer it,
they wanted to have hearings instead. This is clearly contrary to the
intent of NRC regulations governing Motions for Summary Disposition and
their responses, since one of the primary reasons for them is to narrow
the issues for trial so that hearing time is not unnecessarily wasted
by the Board and parties. .
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8. The Board should order Applicants to file by a date certain
exactly what their intentions are regarding Cygna's , role in the
*
design review.of Comanche Peak. This is necessary s%nce it
appears that Applicants are now apparently attempti'g n to phase out
Cygna's independent protocol-controlled effort with the CPRT Plan
.
...ich is governed by no such restrictions. CA'E suggests that the
date for filing be September 4,1985.
9. If any of the parties does not respond at all or does not respond
by the date designated by the Board (with any extensions being
granted only upon a showing of extraordinary necessity), the Board
should rule upon what is in the record regarding each Hotion for
Summary Disposition and responses.
Applicants' previous and latest litigation strategies have already
severely damaged CASE's due process rights in these proceedings. The Board
, should require Applicants, NRC Staff, and Cygna (should it want to
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participate) to comply with NRC regulations regarding Motions for Summary
'
Disposition, as modified with the approval of the Board and all parties.
.
__ CASE urger, that the hoard adopt these preliminary steps which will
allow the Board and parties to better determine the status of the record
regarding design / design QA issues. After that time, it will be appropriate
to consider the next step to resolve these issues or to deny the operating
,
license.
,
Respectfully submitted,
22 53 N
s.) Juanita Ellis, President
p'E
CAS (Citizens Association for Sound
Energy)
, 1426 S. Folk
Dallas, Texas 75224
214/946-9446
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- In reviewing some of the welding information in our files, additional -
support of CASE's position regarding the willful and deliberst -
misrepresentations by Applicants' counsel was found. '5
We call the Board's attention to the Board's Order at page 1 (which
CASE asked the Board to reconsider in our 1/7/85 Motion for Reconsideration
at pages 20 and 21). The Board stated:
". . . we find that Henry Stiner had a long-standing absentee problem
at work and that he was discharged from the plant because of his
absenteeism, not because he gave information to a OC inspector about a
~ gouge ina pipe preceding
"
the three day a*asence that precipitated his
termination. . .
.
And in Applicants' 1/22/85 Reply to CASE's Motion for Reconsideration
' "
of Licensing Board's Memorandum (Concerning Welding Issues), Applicants
counsel argued against CASE's Motion, stating:
" CASE alleges that the Licensing Board erred in finding that Mr. Stiner
was discharged because of a longstanding absentee problem and not
because he gave information to a QC inspector regarding a gouge in a
pipe. Accordingly, CASE moves that this finding be stricken and the
Board exclude its findings regarding Mr. Stiner's credibility from . .
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consideration concerning the issue of his termination. CASE's Motion
. at 20-21.
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" Applicants maintain that the unrefuted evidence currently in the _
. record and briefed by Applicants and CASE provides substantial evidence
to support the Licensing Board's finding regarding Mr. Stiner's
termination. See, e.g., Applicants' Proposed Findings at 3 and CASE's
Proposed Findings on Welding Issues at 6-7 (September 9,1984). For
this reason the motion for reconsideration should be denied."
(Emphases added.)
CASE calls the Board's attention to Applicants' 8/30/82 Answer to
CASE's Motion for Protective order at page 5, second paragraph (copy of
applicable portions are attached for the Board's convenience)~, khere
Applicants' counsel made the represent'ation to the Board:
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8/15/85
UNITED STATES OF AMERICA
NUCLEAR REGULATORY C0lWISS10N
BEFORE T14E ATOMIC SAFETY AND LICENSING BOARD 3'
s-
In the Matter of ' Docket Nos. 5.0-445
and 50-446
TEXAS UTILITIES ELECTRIC
COMPANY, '~ -
et al.
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l
(Application for an
(Comanche Peak Steam Electric l Operating License)
Station, Units 1 and 2) l
CASE'S PROPOSAL REGARDING DESIGN / DESIGN OA ISSUES
IN RESPONSE TO APPLICANTS' 6/28/85 CURRENT MANAGEME M VIEWS
AND MANAGEMEM PLAN FOR RESOLUTION OF ALL ISSUES
on 7/29/85, CASE filed its Initial Response to Applicants' 6/28/85
' '
Current Management Views and Management Plan for Resolution of All 1s, sues,
4.1.J. e hwiporate herein by reference. As stated at that time, CASE is
very conscious of the Board's previous admonitions that parties have a
$ responsioility to present an orderly case to the Board Lif. CASE is still
in the process of preparing a coherent, comprehensive proposal of how we , .
.
,
believe the rest of the case should be handled. We are still unable (as we
.
were then), primarily because of the uncertainties associated with receipt
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oi discovery responses, to siet forth a proposed scheduled for the remainder
~ /1/ It should be noted that Applicants have not followed the Board's orders
,
.in this regard. They have instead presented a Plan to the> Board in
bits and pieces, which is still incomplete and lecks sufficient
specificity to enable a reviewer to thoroughly critique it. This is
not helpful to the Board. Further, it is grossly unfair te CASE and
puts CASE at a tremendous disadvantage and severely damages our due
process rights in these proceedirige. In addition, new information
which CASE received just yesterday and is still reviewing clearly
indicates that Applicants have been planning this for many months, far
in advance of what has been indicated before. We will be addressing
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this in more detail in a later pleading.
?
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UNITED STATES OF AMERICA **
NUCLEAR REGULATORY COMMISSION 2. .
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BEFORI THE ATOMIC SAFETY AND LICENSING BOARD
In the Matter of }(
)(
TEKAS UTILITIES ELECTRIC )( Docket Nos. 50-445-1
COMPANY, et,jal. }{ and 50-446-1
(Comanche Peak Steam Electric )(
Station, Units 1 and 2) }(
CERTIFICATE OF SERVICE
By my signature below, I hereby certify that true and correct copies of
.
CASE's Proposal Regarding Design / Design QA Issues in Response to Applicants'
6/26/05 Current Management Views and Management Plan for Resolution of All
Issues
have been sent to the names listed below this 15th day of August ,19 BJ_,
byr EX,EEEXEXXXII where indicated by * and First Class Mail elsewhere.
.
Federal Ex,=ress -
-
* Administrative Judge Feter E. Elech. * Nicholas S. Reynolds, Esc.
'
U. S. Nuclear Regulatory Coe=ission Eishop, Liberman, Cook, Purcell
4350 East / West Eignvay, 4th Floor & Reynolds
.ethesda, Maryland 20E14 1200 - 17th St., N. V.
Washington, D.C. 20C36
* Judge Elizabeth E. Johnson
Oak Ridge National Laboratory * Ceary S. Mizuno, Esq.
P. O. Box X, Building 3500 Office of Executive Legal
Oak Ridge, Tennessee 37830 Director
U. S. Nuclear Regulatory
* Dr. Kenneth A. McCollom Commission
c/o Neal McCollem . Maryland National Sank Bldg.
4851 Winesanker Way - Room 10105
Fort Worth, Texas 7E123 7735 Old Georgetown Road
Bethesda, Maryland, ,20814
* Dr. Walter H. Jordan . Chairman, /. comic Safety and Licensing
881 W. Outer Drive Board Panel
Oak Ridge, Tennessee 37830 U. S. Nuclear Regulatbry Commission
Washington, D. C. 20555
* Administrative Judge Herbert Grossman
U. 5. Nuclear Regulatory Commission
4350 East / West Highway, 4th Floor
Bethesda, Maryland 20814 .
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9
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Chairman _
Renea Hicks, Esq.
Atomic Safety and Licensing Appeal Assistant Attorney General
Board Panel Environmental Protect $on' Division
U. S. Nuclear Regulatory Commission Supreme Court Building -
Washington, D. C. 20555 Austin, Texas 78711,.
Mr. Robert Ma**.in * Anthony Z. Roissan Esq. '
Regional Adminictrator, Region IV Trial Lawyers for.Public Justice
U. S. Nuclear Regulatory Commission 2000 P Street, N. W., Suite 611
611 Ryan Plaza Dr., Suite 1000 Washington, D. C. 20036
Arlington, Texas 76011
Mr. Owen S. Merrill
Lanny A. Sinkin Staff Engineer
3022 Porter St., M. W...e304 Advisory Committee for Reactor
Washington, D. C. 20008 Safeguards (MS H-1016)
U. S. Nuclear Regulatory Commission
Dr. David H. Boltz Washington, D. C. 20555
2012 S. Polk
Dallas. Texas 75224 Robert A. Wooldridge, Esq.
Worsham, Forsythe, Sampels
. William Counsil, Vice President & Wooldridge
Texas Utilities Generating Company 2001 Bryan Tower, Suite 2500-.
Skyway Tower Dallas, Texas 75201
400 North Olive St., L.B. 81
Dallas, Texas 75201 Thomas G. Dignan, Jr., Esc.
Ropes & Gray
Docketing and Service Section 225 Franklin Street
(3 copies) Boston, Massachusetts 02110
, Office of the Secretary
U. S. Nuclear Regulatory Commission Ms. Nancy H. Williams ..
-
Washington, D. C. 20555 Project ManaFer
- Cygna Energy Services
Ms. Billie P. Carde 101 California Street, suite 1000
-
Governcent Accountability Project San Francisco, California
--. 1901 One St ree t . N. W. 94111-5894
Vashington, D. C. 20009
Mark D. Nozette, Counselor at Law
Heron, Burchette, Rockert & P,othwell
1025 Thomas Jefferson Street, N. W.,
Suite 700
Washington, D. C. 20007
02D D ', , * ;
g ,i's.),Juanita Ellis, President
VCASE (Citizens Association for Sound Energy)
1426 S. Polk -
Dallas, Texas 75224
214/946-9446
2
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Chairman Ranen Hicks, Esq.
Atomic Safety and Licensing Appeal Assistant Attorney General
Board Panel Environmental Protection. Division
U. S. Nuclear Regulatory Commission Supreme Court Building
Washington, D. C. 20555 Austin, Texas 78711 {~
. ;.
5
Mr. Robert Martin * Anthony Z. Roissan. Esq.
Regional Administrator, Region IV Trial Lawyers for Public Justice
U. S. Nuclear Regulatory Commission 2000 P Street, N. W., Suite 611
611 Ryan Plaza Dr., Suite 1000 Washington, D. C. 20036
Arlington, Texas 76011
Mr. Owen S. Herrill
Lanny A. Sinkin Staff Engineer
3022 Porter St., N. W., f304 Advisory Committee for Reactor
Washington, D. C. 20008 Safeguards (MS E-1016)
U. S. Nuclear Regulatory Commission
Dr. David H. Boltz Washington, D. C. 20555
2012 S. Polk
'
Dallas, Texas 75224 Robert A. Wooldridge, Esq.
Worsham, Torsythe, Sampels
Villiam Counsil, \* ice President & Wooldridge
-
Texas Utilities Generating Company 2001 Bryan Tower, Suite 2500
Skyway Tower Dallas, Texas 75201
A nn a - * b Olive St., L.B. 61
Dallas, Texas 75201 Thomas G. Dignan, Jr. , Esq.
Ropes & Cray
. Docketing and Service Section 225 Franklin Street
(3 ccpies) Boston, Massachusetts 02110
Office of the Secretary
U. S. Nuclear Regulatory Commission Ms. Nancy H. Williams
,
Washington, D. C. 20555 Project Manager - -
*
Cygna Energy Services
Ms. Eillie P. Carde 101 California Street, Suite 1000
, Ccvern=ent Accountability Project San Frarcisco, California
1901 Oue Street, N. U. 94111-559'
~~
Washington, D. C. 20009
Mark D. Nozette, Counselor at Law
Heron, Burchette, Ruckert & T.othwell
1025 Thomas Jefferson Street, N. W.,
Suite 700
Washington, D. C. 20007
DA . fA/6
s.) Juanita Ellis7 President
-
SE (Citizens Association for bound Energy)
1426 S. Polk -
Dallas, Texas 75224
214/946-9446
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UNITED STATES OF AMERICA
NUCLEAR REGULATORY COMMISSION 8~
! *:
'
! BEFORE THE ATOMIC SAFETY AND LICENSING BOARD . 1
In the Matter of )(
}{
. TEXAS UTILITIES ELECTRIC }{ Docket Nos. 50-445-1
COMPANY, et al. }( and 50-446-1
-(Comanche Peak Steam Electric }{
station, Units 1 and 2) }{
CERTIFICATE OF SERVICE
.
By my signature below, I hereby certify that true and correct copies of
-
CASE's 8/15/85 Supplement to CASE's 1/7/85 Motion for Reconsideration of' Board's
uno'oa %c-=nd.sn (Concerning Weldina Issues) (LBP-84-54)
:. . . Li. . . 4 c r. : to the names listed below this 15th day of August ,198 5 ,
by: EKp%ENEXXXII where indicated by * and First Class Mail elsewhere.
Federal Ex, tress
..
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* Adcinistrative Judge Peter E. Bloch * Nicholas S. Keynolds, Esq.
~
l'. S. Nuclear Regulaccry Com=ission Eishop, Liber =an, Cook, Perce11
. 4353 Eas:/ West F.ighway, l.th Floor & Feynolds
~
lethesda, Maryla :d 20 S i t. 1200 - 17th St., N. W.
Washington, D.C. 20036
* Judge Elizabeth E. Johnson
Oak Ridge National Laboratory *
Ceary S. Mizurso Esq.
P. C. Box X, Building 3500 office of Ixecutive Legal
cas. raoge, 'lennessee 37830 Director
U. S. Nuclear Regulatory
* Dr. Fenneth A. McCollan Com:sission
c/o Neal McCollers . Maryland National Bank Bldg.
4851 Winesanker Way - Room 10105
Fore Worth, Texas "16123 7735 Old Georgetown Road -
Bethesda, Maryland 20814
- :
* Dr. Walter H. Jordan Chairman, f.tomic Safety and Licensing
. 881 W. Outer Drive -
Board Panel
Oak Ridge. Tennessee 37830 U. S. Nuclear Regulatory Cosmaission
Washington, D. C. 70555
'
* Administrative Judge Herbert Crossaan
U. S. Nuclear Regulatory Commission
4350 East / West Highway, 4th Floor
Bethesia, Maryland 20814
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,ptats.us= by an NRC licensee lies with the Department of Labor,
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jyursuant to 42 U.S.C. 55851, and not with the NRC. .This Board,
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;being vested with only such authority,.as the Commission may .
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delerate, accordingly, is not empowered to pr'oVTWe CASE with-
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the relief it seeks. Thus, the Board should deny CASE's motion
.
as being not within.the Board's autho rity.
B. In Any Event, The Motion Is Without Merit
'
. Assuming arguendo that the Board had jurisdiction over a
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claim under Section 210 of the Energy Reorganization Act.
-
42 U.S.C. 85851, nevertheless the Board should deny CASE's
*
motion as being without merit or substance. As to Mr. Stiner,
even CASE concedes that Mr. Stiner's efforts to reverse his
.
terminations in 1980 and 1981 started "long before CASE had ,
any Id. a Mr. and Mrs. S r I nse r would h t- t c re r I f yi n g In these ,
} proceedings" (CASE Hotion,, ut 1). Mr. .Stiner's situation is ,
,
purely a personnel matter between him and Brown & Root. His
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efforts to obtain his personnel records pre-date these hearings
,.
and therefore have no rational connection to them. Further,
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as a_ forcer Brown & Root employee who was fired for unsatisfactory
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job performance (and not matters r' e lated to these hearings), ,,
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Mr..Stiner's election to testify in these hearings does not ,
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. bring him within the sqope of Section 210 of the Energy ,
,
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Reorganization Act. .
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As to Hra. Stiner, the allegations raised by CAS,E are
false. Applisants have taken ne astion "in retaliation for her
testifying in the operating license hearings for Comanche Peak"
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of Mrs. Stiner. Accordingly, CASE's motion should be denied.",
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e. III. ' CONCLUSION -
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For the foregoing reasons. Applicants urge the Board
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deny CASE's motion for lack of jurisdiction over the subject ,,- ,
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matter or, in the alternative, for lack of merit. ~
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Respect u y Subnitted,
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Nicholl Reynolds
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William A. Horin - -
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Debevoise & Liberman
1200 17th Street, N. W.
--
Washington., D. C. 20036
(202) 857-9817 .
*
Counsel for Applicants
* '
l August 30,.1982 - l . .
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. , UNITED STATES OF AMERICA *
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NUCLEAR REGULATORY COMMISSION
l *
U BEFORE TEE ATOMIC SAFETT AND LICENSINC' BOARD
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In the Matter of ) . -
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, or.EEEAS UTILITIES CENERATING ' ) Docket Nos. 50%A)5 and
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; COMPANY, et,al. ) 50-446 ,
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1(Comanche Peak Steam Electric ) (Application for '
' 0 ::':.:, Units 1 and 2) ) Operating Licenses) ,
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APPLICANTS' ANSWER TO CASE'S
- MOTION FOR PROTECTIVE ORDER ,
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Pursuant to 10 C.F.R. 82.730(c). Texas Utilities Cenerating
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Company, et,al. (" Applicants")..hereby submit their answer to
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CASE's Motion for Protective 0,r d e r , served August 12, 1982, as
- si;;1: :.::d by letter from CASE dated August 19, 1982. For
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the reasons set forth below, Applicants urge the Atomic Safety
and Lluensing Board (" Board") to deny CASE's motion as beyond
the jurisdiction of the Board or, in the alternative, as being -
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without merit or substance. ,
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I. BACKCROUND
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During the evidentiary hearings conducted July 26-30, 1982, .
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CASE identified and' sought a subpoena for the attendence of - .
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Mrs. Darlene Stiner, a QC inspector at the,Comanch's Peak site.
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CA8E stated that Mrs. Stiner possesses information relevant to ..
matters at issue in Contention 5. The board granted the requested .
" *
' subpoena, Tr. 2964. Mrs. S t in e,r is scheduled to testify upon
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s
the resumption of the evidentiary hearings on September 13, 1982.
. On August 11, 1982 Applicants' Counsel became awar( that
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, d;,Mr s . Stiner had been engaged in efforts during working hours to
,
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"His [Mr. Stiner's] efforts to obtain his personnel records pre-date
these hearings and therefore have no rational connection to them.
Further, as a former Brown & Root employee who was fired fbr
unsatisfactory job performance (and not matters related to'these
hearings), Mr. Stiner's election to testify in these heatings does not
bring him within the scope of Section 210 of the Energy Reorganization
Act." (First emphasis in the original; second emphases added.)
CASE does not accept the representations by either Applicants'
witnesses or of Applicants' counsel regarding the reasons alleged by
Applicants for Mr. Stiner's firing. Whether or not the Board ultimately
accepts Applicants' reasons, however, the above representation by
Applicants' counsel is inconsistent with the later sworn testimony of
Applicants' own witnesses. It is a gross misrepresentation to the Licensing
*
Board, and the Board should so rule.
0/. asks that the Board consider this in connection with CASE's 1/7/85
Motion for Reconsideration. Further, we ask that the Board take this and
other misrepresentations of Applicants' counsel into consideration in the
overall context of these proceedings and in any future Board decisions as, to
whether or not to accept representations of Applicants' counsel.
Respectfully submitted,
_
J t .e~ b6~ fi>
s.) Juanita Ellis, President
ASE (Citizens Association for Sound
Energy)
1426 S. Polk
Dallas, Texas 75224
214/946-9446
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CASE submits that this new and significant information, especially when ,
combined with the information contained in our 1/7/85 Motion fq'r.
Reconsideration and its 3/16/85 and 3/19/85 Supplements, fully kindicates
and supports Mr. Stiner and his testimony in this regard.
turther, since the "one overriding factor regarding the Board's
decision involves Mr. Stiner's incredible statement" regarding the manner in
which, and the time in which, he completed a " plug weld" (Board. Memorandum
at page 59, second paragraph), this clarification and substantiation of Mr.
'
Stiner's testimony should add credibility to any other testimony of his
.tich the Board questioned and go far towards establishing his overall
. .
credibility -- and correspondingly call into question the testimony of
appiteaut. witnesses.
However, CASE does not ask that the Board close the record and rule
based on what is currently before it regarding the welding issues. There is
much additional information already in the TRT's SSER's and expected to he ,
in future SSER supplements which the Board should consider prior to closing
the record on welding. We therefore again ask that the Board continue to j
hold the record open awaiting receipt of this additional information '
j
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(although we must admit that we, and undoubtedly Mr. Stiner, would not be
adverse to a ruling by the Board at this point reversing its previous ruling
on this particular point).
Representations by Applicants' Counsel
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In our 1/7/85 Motion for Reconsid'eration (pages 7 through 19), CASE
submitted information regarding what we consider to be deliberate violations
by Applicants' counsel of the Board's specific orders.
-4 .
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It is important 'to remember the time frame during which the various
_
events discussed herein occurred. CASEfileditsProposedFindfngsofFact
^
on Welding Issues on 9/9/84. TheBoard'sMemorandum(ConcernindWelding
Issues) was issued December 18, 1984.
CASE now calls the Board's attention to SSER No. 10 /4/, page N-57,
regarding the TRT's investigation of " plug welding," as alleged by Henry
Stiner. In the next-to-last paragraph, last three sentences, it is stated:
"In a telephone interview with the alleger on September 10, 1984, the
alleger clarified this allegation. The alleger stated that the number
of electrodes used was only an estimate, and that such holes were
' capped' with a weld on either face and had slag and an air pocket in
the ciddle. A ' plug weld' made in this manner woulld obviously require
. fewer electrodes."
Thus, on September 10, 1984 (the day after CASE filed its Proposed
Findings), months before the Board's 12/18/84 Memorandum was issued and
before Henry Stiner or CASE realized that the Board had misinterpretted the
testimony in the record regarding this matter, Mr. Stiner explained to the
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TRT his method of " plug welding." His explanation at that time was
.
consistent with the interpretation giv,en de; CASE in our 1/7/85 Motten f er
_ Reconsideration. The TRT's SSER No. 10 car.e out in April 19F5.
CASE has not completed reviewing the several SSER's issued by the TRT
in recent months; in addition, we have been involved with various other
pleadings and matters relating to these proceedings. It was not until Mr.
Stiner recently called the statements by the TRT to CASE's attention that we
realized their full significance.
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/4/ NUREG-0797, Supplement No. 10, Safety Evaluation Report related to the
operation of Comanche Peak Steam Electric Station, Units 1 and 2,
Docket Nos. 50-445 and 50-446, April 1985
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The Board stated, on page 10, last sentence of second paragraph:
~.
"Mr.Stinersubsequentlytestifiedthathehadperformed20lor30 plug
welds in a single day (Tr. 10699-70)."
**
The Board stated, on pages 59 and 60:
"The one overriding factor regarding the Board's decision involves Mr.
Stiner's incredible statement that a 1 1/4 inch hole in two inch thick
material (on which he allegedly welded many times (Tr. 10683-84)) could
be easily welded in about two minutes (excluding the blending of the
weld with surface material (Tr. 10698-9)), and it would only require
two veld rods to complete (Tr. 11158)." (First emphasis added; second
emphasis in the original.)
"Mr. Stiner's sworn testimony on this point is not accurate and
reliable. The board believes that any welder who had ever veld-
repaired a misdrilled hole of this large size or smaller would have
, been able to at least provide a response that was in the ballpark. In
that Mr. Stiner was not able to do so, the Board questions whether Mr.
Stiner has ever performed a weld repair on a misdrilled hole. . ."
(copnasis added.)
In CASE's 1/7/85 Notion for Reconsideration, CASE explained (pages 31
through 39) the correct interpretation of Mr. Stiner's testimony as CASE
~
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understood it and as it was explained by Mr. Stiner. As discussed therein,
it appears to CASE that the Board has misinterpretted the testhmony in the
_ record regarding this matter.
However, CASE can understand that the Board might be reluctant to
accept some counsel's representations of the correct interpretation /3,/.
Fortunately, it is not necessary for the Board to do so, because there is
now new and significant information which fully supports and corroborates
CASE's interpretation of the record, and which vindicates Henry Stiner's
testimony.and credibility. - *
/3/ The Board has good cause to doubt representations made by Applicants'
ettorneys, as discussed later in this pleading; however, CASE does not
believe that it has given the Board reason to doubt its
representations.
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8/15/85
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UNITED STATES OF AMERICA
NUCLEAR REGULATORY COMMISSION
~
BEFORE THE ATOMIC SAFETY ff;D LICENSING BOARD -
8 .*
In the Matter of } '
Docket Nos. 50r445
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l and 21446
TEXAS UTILITIES ELECTRIC i
COMPANY, et al. l
-
Q (Application for an
(Comanche Peak Steam Electric i Operating License)
Station, Units 1 and 2) i
CASE'S 8/15/85 SUPPLEMENT TO CASE'S 1/7/85 MOTION FOR RECONSIDERATION
OF BOARD'S 12/18/84 MEMORANDUM (CONCERNING WELDING ISSUES) (LBP-84-54)
-
In CASE's 3/19/85 Supplement to CASE's 1/7/85 Motion for
.
Reconsideration of Board's 12/18/84 Memorandum (Concerning Welding Issues) ,
(LBP-84-54). we stated that we would be filing additional information (some
l
of which we believed would be new and signficant). There is still l
l
sdeftfonal information which we expect to file following further contact j
with the NRC Technical Review Team (TRT) to clarify a few matters, but there l
mie 6-v opeciaAc mattets wnacu we pea 2 eve are especially important to get
.
into the hands of the Board right away.
Board's Order at Page 10, last sentence of second paragraph; and page 59 and l
first full paragraph of page 60; re: Welding of Misdrilled Holest
l
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In CASE's 1/7/85 Motion for Reconsideration /1/ (pages 31-39) and '
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briefly in our 3/16/85 Supplement /2/ (bottom of page 32 and page 33),
CASE discussed statements of the Board regarding its strong doubts both of
the credibility of CASE Witness Henry Stiner and of his testimon,y.
/1/ CASE's 1/7/85 Motion for Reconsideration of Board's 12/18/84 Memorandum
(Concerning Welding Issues) (LBP-84-54) -
/2/ CASE's 3/16/85 Supplement to CASE's 1/7/85 Motion for Reconsideration
"""
of Board's 12/18/84 Memorandum (Concerning Welding Issues) (LBP-84-54)
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0. sed. holes i.a the same. cross sec.tionaL plane. at LL. used_
one., -Eh e. unuscol holes, ne.ed. nof- be_ Plug toelded."
:
:
:
x
.
't
#
s
.
I
I
l
!
l
i
,
t
: -
FotA -n-34
A/42.
'
sk n U ) lU. wnJ4-t 8-16-8.5"
#
d iG M ATURE ONTE
INPUT log s HEt- t NoT'E O .
t
_ _ . _ . __. ._. _ _ . . _ . . __ __-- _ _ ____ __o.A_r_ E __.._ . __
-
v -.,
.
-i
% x
*_ 06S6RMATIONS LOG - IMPUT
)
N( TSAP d o. -E 0-
<
N/ER F. PKG, (ser AeeticaeLE)
~
; .trEin FSE- 159 - 10 431 hv. _1
No. (o r APPL s cAet E)
.<
< UNIT No. 1 '
<
< LocAvios Roos 82. ; Safeauava huitA:ag .
') o G.sFRVATiord [DFF'ic 6NCT[CommFNT
e
'
, b.; n s.oe c.t.ia n o{ calla. tva.4 .tuPoovt 40v
Plug taa.Lds. No Plug totids found,
e
Otha< esewtons:
,
Cable. trav .s,upport 21- 60lL6 insfauation is not
- conzi.cLent taith design d.;winstons .shown o
,
,
.
1
-
,
e
.
:
l
Fo/A-%-3L
.
A '93
m
_
itdshi(D.
I d%nuks 9 -/4 - BS~
stGt4ATURE [ DAT E
.
Fottow- de saeET CometerE.: -
0-M-86
__ __ _ _ _ ___ _ __. __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.
1 -
T
N. _
osse RVATiohl S log
N 1 FoLLosd-OP -
.<
x INPUT
< I 06N TI FICATioN : 2eviewed kanac< Jocumme.1rou.
< Found that certain daarturei fvom h ha_ sic hawau d<aiutoo a,e pe,
M'ttnf aL 113ted Ln 40tt.1 d. And.1.e4 f.ke dv3Luink.
; C1kLt tray hanau was Fahr:c hi by CB tex. ~
s
s -
'
.t _ ACTIoM l RGSOLUTtohl .
' 9
'
Mo f urthu act:on vtouived.
,
' _Alote 3. of the. dyawing .tlates, that "attachmewis arc shown l
) ,ef e,tnte Daly. The datovm.*nsLion of attachutnt detail
'
to be. used_ to;tt be made in h /idd"
, \
..' Nobe t,in2- _toofsuit.
the drawing'
-
ALso:
states:#<Ynaximum LengfA
<
2Lso ? "brawing aPPUes to OPPosit e hand
/
.
4
.
b
to
.
.
.
e
y
f .
F01 k-%-M
5
bl%
udsAt b.t n+k
' 0-/6-85
1
s 8
/s iG ogivR E. [/ OSTE
IN90T log SHEET NorEO -
l
__
.w
. - _ _ _ _ _ _ . .. . g. . . ,
x
T 6856RVATio N S log -
<
I AlPUT -
\
)( <
.IS A P Alo. U . d.
N/ER F. Pc6, (e s= n eeci ca et c.) FSE - 00159 - 2&so
MLEth 140. (I f A PPLICAGc 6)
l<' uraiT No. d. -
- ( Lo cA'T~t o t4
-
Sale.Quards bu*L-dina
-
~
,
.
.
], _o G sFRVAT/ord / defic s FN CT ! CommFNY
1fIniPect.ed. fov Plug wet.d.s . de plug us4Ld2 found.
lt o t.hev obsc<vations : t'
- , E 2.sz. elate. Il sec.twe.d. -f.o n e c.c'Liag tuith on e_
.
Richmond. i.wse t ana. ono. 64itti kwi*-h olt .
j' l'
s >'
- ,
-
( G es. 2.tta ched. Ike.1 cl .
.
,
r a
4
.
&
L
0 4 D . -
fem -%.34
a '; AlW
ndsk4).Snuk5# g- i y . g5- l
' SIGt4ATURE D AT E
l
Fettow- 09 SHEET CometerE :
.
9 -/ 7 -8.5-
I -
. - _ _ _ - - - - . - - - - - _ _ . . . . - - . . - - _ . . _ _ . _ . . . . - - . -.
DATE . . - . _ . . . _ .
.0
'
.
m- r,
T .
"
T. oBsFRVATioN 5 log Follow /-UP
x T
<( .
x\ INPUT .l D6NTI FICATioN hoietued too,k Paaint foy erocer doco-
x '
rnentati on anA + tut e so r d d,awimas J Procedu es to,1eraonsit:tv
< et feuem
nnh.
T
T
-
i
't -
c
T. ACTI'oM / RESoLUTiohl ,
.
t
t Llo furthev action requived
s
s
FSE. - Oo isa - us o Rev. 3 .ref ers to FS E. - 001'43 for
i.w3EaLL24: ion details. FSE.- Ool73 permits latert.haage of
<
Richmond. in.tevis and llitti. bolts., and al. lotus to l' u se. HILii
< bolts, t.uhtn RI ckmond. [.nied.t arc. wel avaltable'.'(Ref. F.5E -00173,
l Skeel 2. , re. vision I2.).
) Ths. installed lickwsoud. incevf. taas insoccit4 on .IR - OISS
<(g ow 1- 13-8 3 3 UT ihickness tenort Mo. T- o(oS3, Per Procedure.
'
O.I- Q.P - t t. t4 - 8 R ev. o .
t
: ,
.
, e .
*
- FWA-%-st
A/'k
(b01 -
nfks 9. j y-g5
.i
# s iGNNTvRE / ogg
lN90T log s HEt: t NoT EO , 9 - / 7 -B.5-
0ArE
.-
-
,. ,-
1
' 2 Ad.Rohne.nba.vg
.
.
s/s/as
OOt-Of-icoPa. iiem founcL d.uving vein _tPaction of Ca6Lt. fray
ouAport F.i E - oo 1s3 - 2.'6 30, u nit. d i,usith ERI_ inseacto,
.
MaA rapant. h:41ee et:en uas in acco,aamc with DI- co7,
hug walds. .
Observation: Bua.eLat.c is secured. to the cet =9 u:(k oaa
hithrnond. Inic.et Snd. one Nitt' Kwick.. bott. N
Qachmo K
/ n:.it ,
'
/
/
O @/
<
1a s.otution - F,S E - o o ise - 2 6 3 o
A. 'A ne/e to FDE - Oo / ?9.
FsE-ool% Pn :h h.h,-clup'~p y h't/~ , ~4 kh c .d
H.x han , c~d nJL~ to " ou N.st: bat, tJu R, etna
&k sk<. mo4 w AaAAn. { ALl- FSE - 0of 99, 3haf 2. , stsnh.% /2
TdA d_ Lwt spects .Ta _o( 99 (z/is/ss)
U T ti,icA # eu.pt d. r-06,93
RI - &P - u . i 4 - 8 h. o .
.
i
, 4
-- . . - - + - - , . . , . ---,m.. - . - - - - . . ,.,,.-e- . - , , - - - - - - .
-
., - - - - . . , - - - - - , - .
. ._ - _ _. ._ .. ._ .
! 'y.. '.. '. : -
T
T
o856RVATlord S LOG - l AIPUT
*._. )
k. T5AP slo. Y.d
<
N/Eg:F. PKG. (,e neeci caete) FSE - 60159 -- 4o2.3/ _.4039
; 'CTEin No. (s f APPL e cAec 2)
'
<
< Ut4 tT N o. U h d 3. '
t Lo CAT: ora CaMr. Spreact roog
<
) o 6.sFRVATrord /oEFici Fw cv [ commFNT
i <
-
,
FSE. - 00(59 - 40n . Reinspection of cabu. t.<av haage< (ov etug werdt .
'{ No plug weld.s (cond.. t'
#
FSE. - Oo 153 - 4o3 9 . Located. adiacent to FIE -00159 - 402.3
'
,, b.ucPlatw of f.Mi SUPoovt .Ikwi i.Md.icalidHS
of potential tepaiv by plugtuelding af. One. Location.
'
'
'a
.l t
'a
,
,;
{
-
8.l2 1 C d. dkgYh.
,
4
a
4
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= ;
i
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i !
,
Fo/A-8 '-3'
.
Al97
72nas.tdla. S n a , 9 -iv.ss
- ,
6sier4ATvRE f oATE
'
Fou.ow. Ue saeET Comet.srE :
.
-
-
. - - - - _ _
. .,
-
_ $.
< .
N _
ORsFRVATiohl 5
< T
log - ! Follow /~OP .
'
'
-
.
-
T
%
INPOT .lO6NTIFICATioN -
<
T
T
'
. -
_
% : -
t
r, AcTIoni lREsoLUTiokl +
*
,
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.
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.
.
-
N
S IGMATURE OgTE
.IMPNY LoS : SHEET NoYED : :< ^ .
.i
- .,
- - , - - , - - - - - - - - - - - - - - - - - - - , - - - - -
'. .
~~
.E. WAan berg
.
9-6-85'
.
~C
Iwet.Pand.ent i.ns eecuon ,_ On:t # 1. C2Al, spren,4 voom.
cahte. tvas sueeovts
. ,
, . .
...
,
FS t _ o o i sa '- 402.1 @ no plug wetEs fou,id
"
Fit - ooisoa - 4039 _Incucauen ow bas.eptate
Of fotawkialitPalt bN Plug ;
*
Wrbi.ing of a. M 1dylLLed. hola. )
(Oht Localich).
em-um moro
@ d@
@ @
l
l
l
.. Fo/A46-34
A ,'*8
.
,
. r
-
, .
c. . - ; -
<
<
T_ T
o856RVATio N S log - I AJPUT -
ISAP Alo. Nc
<
\lER n F. PKG. (o f n eet s en eL 6
) .rTEtn No. (o r Aepts cneL2) b'8 SIOb l TDb6 Laeldi fM13f'l$ ,
Uta iT No. .d.
< Lo cxr noa hietaL Generaton bulld;ae. kooms 99 b lc. Sev. 884'.
<
) o GsFPWATioM /OFrici e's cT / commFNT
,
; Fictd obso v2 tron : mies,t s,ca,<2to 1 Fuel Oil bay tanks,
,
S m a LL- bove. PiFih3 Syilde.
-
d
A s- butts vertricat. ion hold tag.s arPLied .several
ih dHO11 ago.
) Chukfor problem s .
Roo m s, SS b/c vequire_ ha<d. key entry
,
*
- -
Potential conura vePorted b1 MRt_ Ichn Gib. son [Eu
ci -i3 85
0
II
.
FotA -86 -36 l
A'+1 l
Q
_ l\ 11 NOA/ hl. (
C ~~3 ,
(Omk. l,.r; [] - / } - g5 \
SIGt4ATURE [ D AT E
i
Fottow- or Saeer CometertE : .
-
ATE _
. - . .-- --
*
., .
R., '
.
-c
N oBsFRVATiohl S log : Follow /-OP
.
< T
-
<< -
'(t -
INPUT .t D6NTIFICATioN x < T T 1 . - T . . t
, () ACTioAl / REsoLUTiohl ,
. ( t
!* .\
s 4 s < i . < \ #
a
( (
- '
.
. 9 > >
.
1
5 IGN6TURE DATE
.I N PUT log st4EET NoTED . '
_ _____ - - - - -------- 3E - - - - ~ ~
"
I
- 1
M .R.), tson nan c e. rg
.0 -13 -8 5
F PRT - I s.c uc. s pe_cific action PLa ns,
.
EN ,7o com0Left anticipatect ant *cipated final tecort
veinspecflon com pletion c.omotetim by veview(fovec24t.)
ortevictso( b's Gird.
doturwentat;s. P26c.5
r4R.C flag E
a3roup.
Xa Skawed.wehL$_ S3 9ll8l85 10l2lBs s zl.2il8s
H.d hg wetd.s 99 9//6/Bs S/23/ss /2/21/gs
.e %ain Sicam ?ioing(NDE) 1.00 comolttrM. 9/20/85 12/2.i/B.S
.b iolly Crane. . Shims W /O/23/BS ///fo/B5 12/2//B5
qcc<ateaa)
S,E.irt. d Ate 1 :
. ra'iL m oticw itudy,wk c[ 9/10
. Jhim inspettions . - Sln
* p ent<JL in sfes.t*cus . .. n 10[ 4
Yolh- BV%
~
_i
. -__ . .- _ _ - ------_
_.
- - -
_. _ _
.t- ,
,
-<, '
'
K:~ 0656RMATIO M S
T. , . LOG ~ IMPUT~
k XSAP rslo. \l .6 '
<
\lER u f. PKG. (s f A ret s cA eL E) CV- I - 00 8 -Tl(a - A ss c.
?< TTEtn No. (if AeetecnetE)
.<
'
< UN IT No, i -
L o cATiet4
}
-
32 '.:.c.: 2. d._ b d'Er. ; AuxiLiarv 6utist a.
; _o6.sFRVATrord /oFricireacT / comment
-
<
l'
,
hiwceerte d. fo< Plug ws.Lat . tdo otug wetd4 found .
'4
Othtv obse< vat;on: 1"4.htc.k. has,c.0 Late t ake.s
f- t er, 4 Hitti botts,
o
,
r L ir) Leagui.
~Ihr e.g.
is botti ave. (4;Lti ktvik. b cLli , on 4 t ne6. *
Ie
'Tha. un c.onvasitoma t. Stud .tlands out S M8' from
the fac.c. of' use baie etate..
, ;
., ( Emb4d.mewt d.eeth, hott inst,ti.n o< Lougl i- w2u
s
\
head <d tod. fasten:n3 m,thod m, ed.c to b. 42eaw2h,/.
4
~
.
.5 ee. a.H.a c.k a.d. .Lk4c.h -
I ! 1 i ,
.
~
.
,
F8/A -E-3 /-
Alst
udoks LO.Suk 9 -ic. g s-
! N beeruse V
we
.
Y*LLoUS.Uf.SneET Comet. erg : 9-/4-85
:
,
-
_
"
DATE
?'i
- _ _ . - . -. - . - , _ - . - _- ._ - - _ _ _ - - ...
- _ _ _ _ _ _ _ < -r -
,,g' ; ',' .
$ OBsERNATsoH S LoS - ~< FoLLov/~ O P <
,g( lNPOT .s OGNTlFICATloN Checked waak pakaae dero >a an ia t to., .
<
< 9.svision L af d,2wina spec:nes 3 14itti bolts and 1 -Lhread ,od 1"e
x 0 ' .9 " t o.,3
<
r
1 '
.< '
t. Action! / REsoLUTiohl
c .-
.
(
' A' threaded nod. soec:Cied as 4"Lon9
base.0L2ts thikutts is
'. 1". Stud sla-dout from the basentale face is. S 'l8
i
It as ocars that em bedmeaf o/ tlie Gread ed_ tod il 2.'/B '.
*
.'. 1" _a PLste_
a 3 thitkneic i .itud itandout of S'/8' s fo'/8 stando
- n *lo"
tickmond intert used wit.k '$" t,kended Rod., A-3(o munterial .
Inita,LLatic a
; I n stvucbio n of baseolate was acomoGJhed. in accovdance. with T
LO.~.t- O.P- ll. l(o - 1, Rev. 2.1, and infoected. to tili procedure .
>
, \-CLA:s
( hotu asd R.'thmond. intest faitenu.L Can los. infeschanged'.
i.nto A Euse of &if(. TOC =CO 2nstruch
.supoorts lov Cla.s1 l.aspecticu ef NNS .teismic cate)go
\/ Piping are. prov'dal in EsowasasRoot
-
Procedurc O.1-0.AP- ll.1- 2.8, Fabrication a.d. issf au.a
ndaH lb.RTnweet:ca Class Cosus nnent dupAcits.
n ~/s.cof Jafelst
a
\ 8 # 9-IC- M
sisarruRE. OATE
Fo/A-st,-u
.IN90T log : SHEET NoT-EO --
.
...
D AT E .
- _ _ _ _ _ _ . . .. _
.. - ._ __
-
.:
*
t
.
. t
_
#. s\\
/ .- .
.
;
'
.
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l
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.__ ,, ,.
i
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..
-
C.V - I - 0 0 8 - T l (o - A 4 5"C. (. Pipe. support)
c-. O.k
.._ ~ ,
1" Sk. 'Eaie plate +_2kes- 4. ' 14Mi bot +.s.. / _.I."8# J.__ leo /Ag si ,
-
_ . . . . _. . ._ .
.bAA.Am ,RLV$.% M 3
'
,
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i 3 lltlti B e t h % . A .f. 1 L ean{d };
rod I "E x O'-4').ag. )
'
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.
_-_ s '
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'
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* ljr1],DDyjgf.599&_._, i
_
h $ Y hl { r kih $- :h % ' x ..: b Y
. .'
. .
9:*g .
_
_
: .
m . s, . .79,m .m
-
, ,
. ...w-
E.
y: ,y.. ..";. ,, ' >w :. :. .
.
- -.
, ,
, .s. .
-
:q -
'
g,,ygie
- .. ~ ,
gg ,I
.
g -, ;;r .....,g.. + ~ .=,. * g {( :o ? 7:,y M
-
: \\
_[l2,,
. ,. . .,y . J g--
,
, ,
,- .
.
_ ,, - a,- . , . , .8
g l
,.- . -
H. . . -l
's . ' g u puse q., *I p 2 M .) . - g g * " . ' - ^'
._._
:
_._2 . 4 .u...__e W<, g _. )_. A 4__* i'/s
. . . . . . _ -
<
l
%. 1
**
:
.
s a
k
:. s
. e. -
@..bs.thowns.ab4<g
x.
'
9-10-85
% OBS6RMATION S LOG - IMPU 1
T< T .
< ISAP N O. -
.d
T
\/ERI F. PKG, (IF A ppli co 6L E) s5 F- X - O SE - Oo3 - F4s R
.ZYEtn
No. (If APPcscAeLE)
<
< .Ut4 T No. lommon. '
< l.ocatiora EueL huD).*no -
<
j _o a.sERvkTiord /DEFici Fra CT ! Co mmFi4T
<
.
< Plug wet.d1 obse<ued. hs ~fom. Loesta<mm aad Ett DM* c,
,
' and. cketkaoC hv R. hanubevg.
' on 'INPt .su Apod SF -x - Os 5 -col- F45 R , -{_too Plug uald.
.
'Tcuits wert Pevformed
and. Do IMO,4 Plug t.oeld NEeuir'on h Jouth west hampv (b
W tc c fo DM d
d .t
fitA A Od k - Wt.ib
k it k h ? l h t e. O f b k e .C a m e. f uMoylb PPm i . 7khLku
-421k t. PL'2te.
E,Xhibit.i One unv4Pa'. red hIld,*lled ble.
-
ms<. et2tes Le ana' krcke< oestes (2.i take. fout
'
14:tu kwik betts
e.2 ck , 'It .er x s '/t L3 (T).
'
, . . -
l
\
!
L
o
,
WO rk Pac.kaa e d.ocu mentaion ~evle.wed at at bct_- BOP vautt.
f oth %-36 [
k
.ItdoM N nO WL.s~.A% 0- 20 - 85~
5, Isist4ATuRE 7 OAT E
Fou. v .uF SHEET CometerE : : 4-2o-85 .[
.=. ._ . . . . .
s, ,
t - 9..u.a. fl o w a. a c g
4
c 4-2.0-85 .
.K OBsERQATsad S 4 log . l 't T 'FoLLov/-UP .'
'
~'-
<
.<(
'
INPUT .t O6NTIFicAT1oN
,T Checked Wavk AAckage docu,w, datta .
T _ofMisdeiLLc4 helsi we<c me eat.<cd in a c.coM aac 4. win <4.oul
Procedurer and ~caui<tmanff Liited b,kw. '
c '
,
T <
3 - *
T -
p ACTioW / R65cLUTiord . i
,
M.o furthey ac_ tion .re.cguT,.e4 .
, i
'
Ba c.l<_opouncl: '
* C_ M C.
l klo. 4 l 30, vestilow 1, w a.1 issued on '2./s/7.9 I
'
L rev;st Locat iou of ancho< bou holn .a sd to renal,-
four mitd.eitted helet h, fintd. welding [ pt.g , pe, Wp_q No,
ll o'5 2. .
( *
Raon; r wtLLs wa<e. m nd..e 64 totLd4<s h\lO and ATN. *
*
Weld acd h/B'&) was withd4 awn o n w ald ovlo ter ra t <<_ -
Otuis.itiow * A-S4030, S rodi, for uie on IF-k. - oss-co3,-F4se.,
dior use. tuith WP.S S ll032. '
. W M9_ sourced % weld. f;ttemate< tat.-lo Meaf O 4 (2.E.7 0 H.
. Q.C. i osec.c. tion of field wotdt, wato:st, weld.e< e. Procedure _
%3t M ; c at ; c., wa1 A cc u e t a te c) ~
JM ' ca in10t d:co at poit Mc -
2 R - Ma s - 8588.
.
This Suppo<t (Cl. ass 5) tuas, originalty f abricated and in -
Stalled linsretted -la ua vo.qu:<e.ments of AMS.t 'B S I. I;
O.I- LQ.P
t t.l(o - i , hv. B ; WP.S - 11o17_)and. CP- cPH [. 9. ;
* %'t supo<t was dedast;fted to non /la(etv/non-Sei.r nie. Later '
,'
by 'b c A N.o. sou , hu.~4.
' n dolt4
lu.knale<c c).zo.gs-
sicourvae // oste.
F0(h -8b~% hl6f
.I M 90T , log SHEET NoT-EO -
9-2o-85 q
~ '
-
*
-
Dake.:- 7//4/PC Pa" a e . I of . l Re in s.oeeCon .
... .
_15 A PNo
. . 1 -
* .. .':u 1. ... . .LjMC
. ~.
..J
, .
!
h .:.~4 : -i-
PE+EE7A
,
< .
~ i ..] 9e. a..
dis. ....
Pka. -
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1. Allegation Category: Civil and Structural 15, Rebar Improperly Drilled
'
2. All ation Number: AC-13,~AC-14, AC-15, AC-18 and AC-40
3. Characterization: It is alleged that undocumented and unauthorized
holes were drilled through reinforcing steel (rebar). The issue includes
allegations relating to:
a.
~ the loan of rebar drills without proper documentation (AC-13),
'
b. l the unauthorized Jcutt:ing of rebar in non-specific locations
(AC-14,AC-18,AC-40),and
!
'
c. the unauthorized cutting of rebar used in the in tallation of the
trolley process aisle rails'in the Fuel Handling Building (AC-15).
4. Assessment of Safety Significance:
The NRC Technical Review Team (TRT)
contacted
his concerns. the individual who made allegations AC-13 and AC-14 to clarify
The TRT did not initially attempt to contact the individuals
who made allegations AC-18, AC-40, and AC-15.
t
a.
AC-13 concerns the loan of rebar drills allegedly used for the
unauthorized cutting of rebar. During the NRC investigation of this
matter, the NRC Office of Investigation (OI) interviewed nine
individuals alleged to have knowledge of unauthorized cutting of
rebar. These individuals provided sworn statements denying any
knowledge of this activity. These statements are a part of OI Report
A4-83-005 (May 20, 1983), which concludes that "there was no
testimony received indicating that holes were drilled or rebar was
cut without proper documentation, and no evidence was found to
contradict the testimony of these individuals." One instance of i
possible unauthorized cutting of rebar is discussed in a supplement
to the OI report (September 7, 1983). i
below in relation to allegation AC-15. This instance is discussed ,
j
Because the alleger did not specifically identify who made
i
unauthorized cuts of rebar, or where this cutting took place, the TRT <
'
attempted to quantify the amount of rebar that allegedly was cut
without authorization. In discussions with the TRT, the alleger ;
estimated that approximately five percent of the diamond core drill r
bits ordered by him were used in an unauthorized manner. He further !
;
estimated that one drill could be used to cut up to five rebars, !
depending upon the extent of cutting required. Although he could not
be specific as to how many drills he ordered, the alleger thought
that the number would be in the thousands. The NRC Region IV
Investigation.of this issue indicated that 415 diamond core drill
{.
bits were purchased during the period in question (IE Report 03-27).
Using the actual number of drill bits purchased, together with.the
information provided by the alleger, the TRT estimated that there
could be approximately 100 alleged unauthorized rebar cuts.
Considerint the large amount of reinforcing steel used in the plant,
and the fact that the structures consist primarily of heavily
reinforced concrete walls and slabs, the TRT determined that, if such
"
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unauthorized rebar cutting occurred, the amount involved would have
an inconsequential effect on the safety of the structures.
b. Allegations AC-14 AC-18 and AC-40 also raise questions regarding the
unauthorized cutting of rebar, but do not identify specific
locations. During the course of the NRC Region IV investigation of
this matter, the alleger provided a log book which, it was reported,
would identify the unauthorized and undocumented rebar cutting. I
However, the Region IV inspector could not identify one rebar cut '
listed in the log that was not authorized. The TRT also reviewed '
the log and came to the same conclusion. ,
~
In discussing this matter with the TRT, the alleger confirmed that
there was documentation supporting "ninety nine and three quarter
percent" of the rebar cuts identified in the log. As part of Report
83-27, the NRC Region IV investigator traced 32 authorizations,
approximately half of the documents noted in the log for the rebar
cutting. He found that in all cases rebar cuts were properly
identified on a design change authorization (DCA) or on a component '
modification card (CMC). In addition, the rebar cuts were traced to
and identified on specific building structural drawings, with the
corresponding authorizing document number. The TRT reviewed 10 CMCs
and confirmed the findings of the Region IV investigation.
In reviewing authorizations in the log, the TRT noted that certain
CMCs involved a number of rebar cuts in one area, and selected these '
for review. In one case, 7 different CMCs (3307, 3664, 3665, 3666, '
.
3667, 3668 and 3669) seemed to pertain te one area and accounted for
68 rebar cuts. Upon reviewing the documentation, the TRT found that g
these cuts were made in a tunnel area in the Fuel Building. (The g
g
alleger identified this as a location where a large number of ecbar 4
was cut.) However, the 68 cuts were ' arranged such that only 9 bars
actually had been cut. In another case, the log indicated 25 rebar f
cuts pertaining to CMC 00979. In this case, the TRT determined a
t
that all the cuts were made on one reinforcing bar in a support beam, t
Finally, the log indicated eight rebar cuts pertaining to CMC 3022. 6 '
Once again, these eight cuts were to one bar in a support beam. All .L
cuts were made in accordance with the rebar cutting criteria provided
by Gibbs & Hill. These examples also illustrate the point that a r.
large number of rebar cuts recorded are not necessarily synonymous
.
"
i with an identical number of rebar actually being cut. In all cases,
i
l ! i one bar was cut a numt.ar of times, but adjacent bars were not. Thus, [ !
the cuts were arranged 'o minimize the overall effect on the strength
of the structure. !p
The TRT estimates that approximately 335 rebar cuts are indicated in
the alleger's log. Discussions with the alleger revealed that he i
', l
believes he cut approximately five percent more rebar than was 'e
authorized, a number that corresponds to approximately 17 !
unauthorized rebar cuts. As noted earlier, such a number would have 5[
i,
little effect on the safety of the structures.
, ,
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c. Allegation AC-15 identifies a specific instance of the possible
unauthorized cutting of rebar. In this case, a former Brown & Root
'
,
employee stated he possibly drilled holes through rebar in a concrete
floor without a component modification card (CMC) or a design change
4
authorization (DCA). He explained that in January 1983 he drilled
i approximately 10 holes about 9 inches deep while installing 22 metal
;
? plates with a core drill. He said the metal plates were used to
secure the trolley process aisle rails located on the 810-foot,
6-inch floor level in Room 252 of the Fuel Handling Building.
.
'
The TRT inspected the trolley process aisle rails and its anchoring
( system and observed no violations of pro. ject drawings or
". specifications. The TRT reviewed the reinforcement drawings
:
(2323-S-0800 and 2323-5-0820) for the Fuel Handling Building to
i t
determine the location of rebar. The drawing showed three layers of
j
: i
i reinforcement in the upper part of the mat, which consisted of a No.
18 bar running in the east-west direction, in the first and third
f layers, and a No. 11 bar running in the north-south direction, in the
j second layer.
:
2
* The review of the reinforcement drawings (2323-5-0800 and 2323-5-0820) i
revealed that the layout of the east-west reinforcement and the I
&
trolley process aisle rails was such that only one bar of the east-
west reinforcement could be cut by drilling holes for rail anchors.
.
1
j
( However, if 9-inch holes were drilled, both layers of the No. 18
reinforcing bar would be cut. De' sign Change Authorization (DCA) No.
j 7401 was written for authorization to cut the uppermost No.18 bar at
only one rail, but it did not reference the authorization to cut the
lowermost No. 18 bar.
.
The DCA (No. 7041) also stated that the expan- l
sion bolts and baseplates could be moved in the east west direction
to avoid interference with the No. 11 reinforcement running in the l
north south direction. The information described in DCA No. 7041 was I
substantiated by Gibbs & Hill calculations. The DCA approval was
based on the understanding that only the uppermost No. 18 reinforce-
ment would be cut. If the 10 holes were actually drilled 9 inches
deep, then the allegation that reinforcement was cut without proper
authorization may be valid.
The DCA indicated that the holes were drilled to accommodate 1/2-inch
Hilti bolts, which require a minimum embedment of 5-1/2 inches (as
noted in Fig. 39, Sh. 5 of 5, attached to DCA-7041). Since there was
p
,
no need to drill the holes deeper than 5-1/2 inches, the alleger may
not be correct in stating that the holes were drilled 9 inches deep.
3
6 l
i
Although the allegations discussed above, with the exception of AC-15
I{ which requires further action, cannot be substantiated, the fact that
such allegations were made indicated that there was no effective
quality assurance program to oversee the issuance and use of diamond ,
.
core drill bits. l
.% {
$
j
The TRT interviewed the individual concerned about the loan of rebar
j
drills without proper documentation and the unauthorized cutting of i
l, rebar at nonspecific locations to inform him of the TRT's finding. l
l
i
This individual did not agree with certain TRT findings. In l
l
.
K 'B9
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,
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_ .. ._ . ._ _, _ _ . . . - - - - -
*
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particular, the alleger felt that the TRT's estimate of approximately
- 120 unauthorized rebar cuts was much too low. He believes that the
number of drill bits ordered by him was in the thousands and that as
much as 20 percent of the drill bits may have been used in an
- unauthorized manner. It was also his opinion that the unauthorized
cutting of rebar was not limited to his period of employment,' but
occurred for the duration of the project.
- As a result of these additional discussions with the alleger, the TRT
searched TUEC's files relating to the purchase of diamond drill bits
and found that 1170 drill bits were purchased between January 13,
1978 and January 14, 1980. This number is more in agreement with the
a
4
alleger's assessment and is higher than the previously reported
number of 415 (IE Report 83-27). The TRT also found that there were
} -
> a total of 3368 drill bits ordered from one manufacturer between '
January 13, 1978 and March 18, 1983. After this period, other
S manufacturers supplied the drill bits. Based on the usage through
March 10, 1983, the TRT estimates that approximately 5000 diamond
drill bits have been used to date on the project. Assuming that 20
percent of these drill bits were used in an unauthorized manner and
that each drill bit could be used to cut _up to five rebars, the TRT ,
estimates that there could be approximately 5000 alleged unauthorized
, rebar cuts.
L
The TRT estimated that, depending upon the average ~1ength of rebar
assumed, there are approximately 800,000 to 1,200,000 bars installed
in all of the concrete structures. Thus, if 5000 bars were cut
without authorization, they would represent approximately 0.6% of the
total rebar in the plant. Even if all 5000 drill bits were used in
an unauthorized manner it still would only represent 3% of the total
rebar in the plant. Thus the percentage of rebar that could have ,1
'
been cut without proper authorization is low. Since no information
was supplied to the contrary,.the TRT assumed that these unauthorized
cuts, if they did occur, were scattered throughout the plant and not
' concentrated in one localized area. In addition, as noted earlier, a
large number of rebar cuts are not necessarily synonymous with an
identical number of rebar actually being cut. It is also noted that
i
nuclear structures are very conservatively designed. In addition to
the conservative loads, load combinations, and safety factors fj
f'
> utilized in the design, it is the common practice of the design { I engineer
required by to specify 5 to 10 percent more rebar than is actually
his calculations. d
.
E This occurs because it is difficult to' i
obtain the exact area of reinforcement required using standard bar
sizes and standard bar spacing. The area of reinforcement is
I
selected from charts which show the area provided for each bar size (t
<
{ at a given spacing. Rather than underdesigning, the designer selects
I
an area of reinforcement from the charts which is higher than that $
I
which is actually required. In addition, because critical structures T
*
contain a large number of bars, they are not generally vulnerable to j
the random cutting of a small number of bars. i
4l
5. Conclusion and Staff Positions: The TRT concludes that allegations AC-13, i
'
.
AC-14, AC-18 and AC-40 have no structural safety significance.
,
-
K-90 -
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[. -
- _ _ _ _ . - ___________ ___ . - ________ _ _
/*
.' .-
,
3
.
. . .
a. The allegations were not specific as to who made unauthoriged cuts of
rebar er where the cuts took place.
,
b. The cumber of unauthorized rebar cuts alleged, if true, would have an
I
inconsequential effect on the safety of the structures.
However, the results of these evaluations will be further assessed as a
part of the programmatic review concerning procedures addressed under
QA/QC Category 6, "QC Inspection." Therefore, the final acceptability of
these evaluations will be predicated on the satisfactory results of the
programmatic review of this subject. Any adjustments to the existing
conclusion of this evaluation resulting from the programmatic review will
be reported in a supplement to this SSER.
Allegation AC-15 will remain open until the information requested of Texas
Utilities Electric Company (TVEC) in " Actions Required" is provided.
The TRT attempted to contact the individuals who made allegations AC-18
,
and AC-15 to inform them of the TRT's findings. The individual who made
allegation AC-18 will be informed of the TRT's findings by letter. One of
the two individuals involved in allegation AC-15 cannot be located; the
TRT is still attempting to contact the other. The TRT also contacted the
individual who made allegations AC-13, AC-14, and AC-40 to discuss the
TRT's findings pertaining to the concerns he raised in the first closure
interview. An interview was arranged; however, later the alleger indicated
he did not want to meet with the TRT. A letter will be sent to him
informing him of the TRT's findings.
6. Actions Required: TUEC shall provide:
A
1.
3 Information to demonstrate that only the No.18 reinforcing steel in
j
s
the first layer was cut, or
2.
Design calculations to demonstrate that structural integrity is
maintained if the No. 18 reinforcing steel on both the first and
y
, third layers was cut.
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1. Allegation Catecory: Miscellaneous 11, Polar Crane Shimming .
2. Allegation Number: AM-15, AM-16
3. Characterization: It is alleged that the shimming of the Unit'l polar
crane rail system supports was improper and that the polar crane system is
- improperly installed.
4. Assessment of Safety Sionificance: The NRC Technical Review Team (TRT)
.tried to contact the alleger during its inspection to learn more details
about the allegation. On August 14, 1984, two TRT members visually
inspected the shims from the polar crane. During the first 180* rotation
of the crane, the TRT members stood on the platform above the operator's
booth to view the radial restraint brackets and the seismic restraint
brackets. Several brackets at different locations appeared to have gaps
in excess of 1/16 inch. However, this only confirmed what had been ,
previously observed by an NRC Region IV Resident Inspector and was docu- l
mented in NRC Inspection _ Report (IR) No. 50-445/84-08. which required
corrective action which was not yet completed.
The TRT then moved below the operator's booth to view the polar crane rail
system from another vantage point. The TRT observed the shims used for
shimming 28 crane girder to girder-support brackets. During this 180*
rotation, the TRT observed large gaps, particularly on the inside edge
(looking from the inside of the Containment Building to the outside),
g
E The TRT met with the Texas Utilities Electric Company (TUEC) project civil
C engineer, the Brown & Root (B&R) project control manager, the B&R subcon- l
tracts supervisor, and a representative from Chicago Bridge and Iron '
(CB&I) to determine the gap-tolerance specification between be'aring plate
"A" .(Dwg. 2323-SI-0515, Revision 4) and the girder to girder-support
4 bracket. Neither Gibbs & Hill (G&H) specification SS-14 ncr the Crane
4 Manufacturers Association of America Manual (CHAA-70) addressed this issue.
j The meeting failed to produce a specific answer; however, copies of two
M letters related to the issues were provided. The first, a B&R letter
,
(No. BRF-7404), dated November 8, 1977, contained the as-built measure-
ments of gaps at all shim locations and a request for G&H to evaluate this
information and provide direction. At 28 locations, the as-built drawings
showed gaps that ranged up to 0.581 inch. In the second letter, G&H
(GHF-2207, dated November 28, 1977) responded as follows:
. Girder Seat Connections
[ These scated connections will not require shimming since the area in
bearing is at least the width of the bottom flange of the crane girder.
The gap dimensions indicated in the Brown & Root survey exist only at
the extreme edges of plate A, Section 3-3, Dwg. 2323-S1-0515, Revi-
sion 4.
..
Q The TRT noted that the bottom flange of the girder referenced in the
j G&H letter (the bearing surface) is 20 inches wide.
l? !J On August 30, 1984, an NRC inspector, accompanied by a TUEC quality con- 'p trol (QC) inspector, inspected the 28 crane girder to girder-support hMr i
'
K-121
.
'
Jl
__ - - ____ . _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ . - _ _ _ _ _ _
.
.
- .
.
.
L
k bracket shims. Nine girders, identified as A7-6 right-end (RE), A7-8 (RE) A
k
A7-12 (RE), A7-14~(RE), A7-18 left-end (LE), A7-19 (RE), A7-20 (RE). A7-24 '
(RE), and A7-25 (RE) .had gaps in excess of 1/16-inch extending under the
bottom flange. This observation invalidated the G&H assumption of 20 in- ,
ches of bearing surface. 1
9
The TRT closely observed girder A7-20 (RE) as the crane wheels passed
directly
of over the support bracket and saw no visible compression (closure)
the gap. ..
In addition, a visual inspection of the complete rail system ?
revealed that the rail has-moved or is moving circumferential1y,-as-indi- '
cated by the fact that some of the 1-inch-diameter stabilizing rods are
* bent from the force of this movement. The 3/8-inch designed gap between (y
the ends of the rail section also varied from 0.000-inch to 0.875-inch, h 4
when measured at the inside edge of the rail. In addition, three of the ;
rail-to-rail ground wires and two Cadwelds were broken, and at least two l
rail shim plates had partially worked out from under the rail, r
:
The TRT interviewed the polar crane operator and asked if he knew of any
existing problems with the crane or its operation. He replied that the
.
crane operates satisfactorily and has experienced no apparent problems.
He also stated there are no " dead spots" (i.e., no loss of electrical
energy at spots) in the bus bars.
The TRT found additional shimming problems and additional types of problems,
described above, that had not previously been identified. These deficien- ,
cies appear to be safety significant and generic. >
5.
Conclusion and Staff Positions: Based on the above inspections, the TRT
concludes
j
significant.that this allegation is substantiated and is potentially safety !
The problem with shimming and inspection of safety related
work was first identified in 1982. Because problems still existed in 1984,
this matter appears to be generic. !
>
j
On November 8, 1984, the TRT interviewed the alleger to provide the above il
findings and conclusions. The alleger stated that his concerns were ; !
y I , s , N L, resolved. 1
..
6. Actions Required:
TUEC shall inspect the polar crane rail girder seat
%1 - 27 connections for the presence of gaps which reduce the twring surf ace to
less than the width of the bottom flange.
TUEC shall perform an analysis which will determine whether existing gaps
are acceptable or if. corrective actions are required. TUEC shall determine
20
'
Ok - if additional rail mova=-at-is accurrino and, if so, provide an evaluation'
-
l
of safety significance and the need for corrective action.
TUEC shall perform a general inspection of the polar crane rail and the
30 " fo rail support system, correct identified deficiencies of safety signifi-
<
cance, and provide an assessment of the adequacy of existing maintenance 9
!
and/or surveillance programs. j.
{
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ve een evaluated and
f .
a response'sade to the referenced violations, TUEC shall consider this
matter as a part of the inspection of the polar crane system.
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