ML20210N950
| ML20210N950 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 04/30/1986 |
| From: | Danielson D, Fair J, Yin I NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20210N927 | List: |
| References | |
| 50-346-86-04, 50-346-86-4, CAL-85-13, IEB-79-14, NUDOCS 8605050334 | |
| Download: ML20210N950 (18) | |
See also: IR 05000346/1986004
Text
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U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Report No. 50-346/86004(DRS)
Docket No. 50-346
License No. NPF-3
Licensee: Toledo Edison Company
Edison Plaza
300 Madison Avenue
Toledo, OH 43652
Facility Name:
Davis-Besse Nuclear Power Station, Unit 1
Inspection At:
Dasis-Besse Site, Oak Harbor, OH
Bechtel Power Corporation, Gaithersburg, MD (Bechtel)
Inspection Conducted: January 14-16, February 5-6 and 19-20, 1986 at the site
January 29-30 and April 1-2, 1986 at Bechtel
oh
Inspectors-
. T. Yin
Da~te
W
duin
(April 1, 1986 only)
Date
e-
Approved By:
D. H. Danielson, Chief
d[3o[/t
Materials and Processes Section
Date
Inspection Summary
Inspection on January 14 through April 2, 1986 (Report No. 50-356/86004(DRS))
Areas Inspected:
Special, announced inspection of the auxiliary feedwater pump
turbine steam supply (AFPTSS) piping modifications; the Facility Change Request
(FCR) system; the implementation of Region III (RIII) Confirmatory Action Letter
(CAL) 85-13 actions; actions on Licensee Event Reports (LER); the status of
completion of IE Bulletin (IEB) 79-14; the Bechtel control of High Energy Line
Break (HELB) analyses; and followup on previous inspection findings.
Results: Of the areas inspected, two violations were identified; (failure
of the Bechtel staff to follow procedures and failure of TED to follow site
procedures - Paragraphs 4.b and 8.b(1); failure of the licensee to take
adequate corrective action on identified problems - Paragraph 8.b(2)).
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DETAILS
1.
Persons Contacted
Toledo Edison Company (TED)
- +*T. J. Bloom, Senior Licensing Specialist
- T. Chowdharm, Manager, Engineering Services Department
- +*P. H. Straube, Senior Engineer
J. Dunne, Senior Engineer
T. J. Myers, Nuclear Safety and Licensing Director
- L. Ramset, Quality Assurance Director
F. R. Miller, Staff Engineer
P. W. Jacobsen, Senior Engineer
J. F. Helle, Nuclear Facility Engineering Director
- S. J. Osting, Senior Assistant Engineer
D. R. Wyokko, Regulatory Affairs Supervisor
- D. Kies, Manager, Mechanical / Structural Engineering
- H. Brinkmann, Director, Nuclear Facility Engineering
C. Merkbel, Civil and Structural Systems Engineer
Bechtel Associates Professional Corporation, Ohio (Bechtel)
J. W. Brothers, Chief, Quality Engineering
N. Tolani, Senior Engineer
+M. S. Wasserman, Mechanical Engineer Supervisor
- M. L. Murphy, Senior Engineer
- W. C. Lowery, Project QA Engineer
A. T. Vieira, Engineering Technical Specialist
- +D. C. Kansal, Deputy Division QA Manager
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J. M. Ogle, Civil Engineer Supervisor
- +D. L. Gill, Project Quality Engineer
+E. J. Ray, Project Engineer
C. H. Abutaa, Senior Engineer
R. Lee, Engineer Supervisor
+T. I. Gillespie, QA Manager, Projects
+S. R. Kalavar, QA Manager, Audit
S. A. Bernsen, Division Manager of QA
J. B. Wallis, Senior Engineer
- V. R. Marathe, Assistant Project Engineer
K. I. Patel, Engineering Supervisor
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U.S. Nuclear Regulatory Commission, Region Ill (RIII)
- W. Rogers, Senior Resident Inspector
- D.
Kosloff, Resident Inspector
+ Denotes those attending the management exit meeting on January 30, 1986
at Bechtel.
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- Denotes those attending the management exit meeting on February 20, 1986
at the site.
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- Denotes those attending the management exit meeting on April 2,1986 at
Bechtel.
2.
Licensee Action on Previous Inspection Findings
a.
(Closed) Unresolved Item (346/83-17-04):
Several new vintage
Grinnell Corporation hydraulic snubbers with Miller cylinders,
including PSP-1-H4 and PSP-1-H6 installed on the Pressurizer Spray
Piping System, were observed installed without fluid reservoir
breather and filter units.
The NRC inspector reviewed the site
Temporary Modification Request, dated January 10, 1986, and
Section 8.2.5 of Procedure MP1410.02.04, " Maintenance of Hydraulic
Snubbers," and considered the licensee's measures for reinstalling
the filter units to be acceptable. A purchase order procuring
50 new filter units was issued on January 10, 1986.
b.
(Closed) Violation (346/85013-01):
The licensee failed to document
nonconformances in accordance with procedure requirements.
The NRC
inspector reviewed Item IV.A.2 of the TED response letter (Serial
No.1-604) to the NRC, dated January 27, 1986, and considered it
acceptable.
TED corrective actions are documented in RIII Inspection
Reports No. 50-346/85013, Paragraph 8; No. 50-346/85033, Paragraphs 2
through 5; No. 50-346/85035, Paragraph 4.b; and Paragraph 4.a of this
report.
c.
(Closed) Violation (346/85013-02):
After the AFPTSS problems were
identified, the TED evaluations did not investigate the cause of the
problem and consequently measures to prevent recurrence were not
developed.
The NRC inspector reviewed Item IV.A.3 of the TED response
letter (Serial No. 1-604) to the NRC, dated January 27, 1986, and
considered it acceptable.
TED corrective actions are documented in
RIII Inspection Reports No. 50-346/85013, Paragraph 10;
No. 50-346/85035, Paragraphs 4.a and 6; and Paragraph 6 of this
report.
d.
(Closed) Violation (346/85013-03):
Inadequate piping suspension
system QC inspection and ineffective implementation of the IEB 79-14
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walkdown inspection program.
The NRC inspector reviewed Item IV.A.1
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of the TED response letter (Serial No. 1-604) to the NRC, dated
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January 27, 1986, and considered it acceptable.
TED corrective
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actions are documented in RIII Inspection Reports No. 50-346/85033,
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Paragraphs 2 to 7; No. 50-346/85035, Paragraphs 4.b and 5; and
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Paragraph 4 of this report.
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e.
(Closed) Violation (346/85013-06):
The licensee failed to report
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AFPTSS component deficiencies in accordance with 10 CFR 50.73
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requirements.
The NRC inspector reviewed Item IV.B of the TED
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response letter (Serial No. 1-604) to the NRC, dated January 27,
1986, and considered it acceptable.
TED corrective actions are
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documented in RIII Inspection Reports No. 50-346/85033, Paragraph 2;
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No. 50-346/85035, Paragraph 4.b; and Paragraph 4.a of this report.
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f.
(Closed) Violation (346/85031-01):
Lack of a design interface
procedure between Bechtel and Grinnell for evaluating pipe hangers
in accordance with IEB 79-02 and IEB 79-14.
The NRC inspector
reviewed TED response letters, Serial No. 1-593, dated November 25,
1985, and Serial No. 1-616, dated February 27, 1986, and considered
the matter resolved. The need to review Grinnell calculations is
discussed in Paragraph 10 of this repo:t.
g.
(Closed) Violation (346/85031-02):
TED did not effectively implement
its FCR system in that a number of safety-related supports were not
restored to their FSAR condition in a timely manner.
The NRC
inspector reviewed TED response letters, Serial No. 1-593, dated
November 25, 1985, and Serial No. 1-616, dated February 27, 1986,
and considered the licensee actions to be acceptable.
The near term
support modification work to assure FSAR conditions were met, was
conducted in accordance with RIII CAL 85-13, Item 1.a(4)
requirements (see Paragraph'4.d of this report).
The licensee's
long term upgrade of the FCR program are being reviewed by RIII and
NRC Headquarters personnel.
See Paragraph 2.k of this report for
details concerning piping design and support modifications.
(Closed)Viofation(346/85035-02): The licensee failed to use the
h.
appropriate allowable stresses specified in Bechtel Evaluation
Procedure MGP-04 for evaluating stresses at weld attachments to the
piping pressure boundary.
The NRC inspector reviewed the TED
response letter, Serial No.1-314, and Bechtel Procedure CGP-04,
" Procedure for Evaluating Nonconformance Reports Related to Pipe
Supports, Pipe Anchors, and Seismic Restraints at Davis Besse
Nuclear Power Station, Unit 1," Revision 1, dated January 28, 1986.
The NRC-inspector noted that Procedure CGP-04 allowed higher
allowable stresses for the SSE load combination than Bechtel
Procedure MGP-04.
Bechtel reanalyzed all affected piping using
the revised allowable stresses.
The results were documented in
a Bechtel letter to TED, BT-16555, " Procedure GCP-04; Faulted
Condition," dated April 2, 1986.
Procedure CGP-04 was subsequently
revised to reflect the lower allowable stresses for the SSE load
combination on April 4, 1986 as Revision 2.
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(0 pen) Unresolved Item (346/85035-03): -The NCR evaluations for weld
deficiencies designed to the AISC specification do not require meeting
the specification minimum weld sizes which correspond to the base
material thicknesses.
The NRC inspector reviewed the licensee's
response contained in an intracompany memorandura (File 0093, T-0294)
dated January 24, 1986.
The NRC inspector will discuss this matter
with NRC-NRR to determine if it represents a position acceptable to
the NRC staff.
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J.
(Closed) Unresolved Item (346/85035-04):
Bechtel exhibited
questionable design control for conducting HELB analyses and whip
restraint designs.
See Paragraph 7 for details of the followup
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inspection.
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k.
(0 pen) Open Item (346/85035-05):
The TED MWO and FCR systems require
further evaluation and improvement.
A temporary organization named
Engineering Services Department (ESD) was formed to focus the
company's attention on closing out open FCRs and to establish better
ways of handling future FCRs.
ESD is presently manned by a full time
technical staff of ten and occasionally by engineers of various
disciplines depending on specific needs. The NRC inspector met with
the Manager of ESD and reviewed the ESD organization chart and the
" Project Proposal for Closeout Backlog Evaluation" to evaluate the
scope and provisions of the project and had no adverse comments.
A more detailed review of ESD will be conducted by the NRC inspector
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to assess the effectiveness of the new system.
In addition to the
above effort, an audit was conducted by Stone and Webster Engineering
Corporation of the FCR system to identify system deficiencies and to
recommend improvements.
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3.
Licensee Action on Licensee Event Reports (LERs)
a.
(Closed) LER (346/85019-LL):
"PORV Discharge Line Overstressed Due to
Inadequate Heat Trace," reported on November 6, 1985.
See Paragraph 8
for inspection details,
b.
(Closed) LER (346/85023-LL):
" Error in the High Energy Line Break
Analysis in the Auxiliary Building," reported on December 28, 1985.
During a TED review of the environmental qualification (EQ) and single
failure analysis for a proposed modification to the AFPTSS piping,
TED discovered that portions of the system upstream of the MS
admission valves 106, 106A, 107 and 107A would not be isolated during
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a postulated high energy line break event.
A break in these pipe
sections would affect rooms 500, 501 and other connected rooms. The
licensee's corrective actions are included in RIII Inspection Report
No. 50-346/85035, Paragraph 6.c.
The NRC inspector noted that similar situations could exist in other
safety-related high energy piping systems. TED stated that efforts
to expand the scope of the HELB review had been initiated, and that
deficiencies had been discovered in the main feedwater lines.
TED
also indicated that nonconforming conditions will be documented in
either an amendment to LER 85023 or in a new LER.
4.
Implementation of RIII CAL 85-13_[ction Items
As a result of a meeting con @;cte at the site on October 9, 1985 (RIII
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Inspection Report No. 50 A ' 150
, Paragraph 4) RIII CAL 85-13 was
issued on October 17, 19t5
The licensee's implementation of the actions set forth in the CAL was
reviewed by the NRC inspector.
The status of CAL Item 1 (action items
prior to plant restart) is as follows:
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a.
Item 1.a(1)-(Closed)
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Reinspections per_ CAL 85-13 Item 1.b
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All 2365 hangers have been inspected; 921 require evaluation and
,f, , clospout prior to restart.
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-IStatus of Engineering Evaluations as of March 25, 1986
858 NCRs', were writteE ,af ter evaluation of the 921 inspected
hangers
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,Of the 858 NCRs, 656 were dispositioned to "Use-As-Is"
Of the 858 NCRs,202 required corrective actions; rework has
been completed for 180 of the NCRs
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b.
Item 1.a(2) (Closed)
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The following piping stress analyses were rerun by Bechtel to
determine system operability:
Systems
Affected Bechtel Calculation Nos.
188, 180, 80A, and T-010A
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LPI/ Core Flood
T-008
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/ HPI
56D, 56F, and T-009B
22F
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32E
' Hydrogen Dilution
119L
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MS-Exhaust from AFPT
161
The NRC inspector selected the following calculations for review at
Bechtel and concluded that the evaluations were technically sound
and conservative:
No. 56F, " Davis-Besse High Pressure Injection System,"
Revision C2, dated December-20, 1985.
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No. 188, " Davis-Besse Low Pressure Injection System,"
Revision C3, dated January 3, 1986.
In Calculation No. 18B there were two restraints that did not meet
FSAR commitments.
They were not reported to TED in accordance with
Bechtel Procedure MGP-04, " Procedure for Control of Interim /Short-Term
Allowable Stress Criteria for Seismic Category I Piping Systems at
Davis-Besse Nuclear Power Station Unit 1," Revision 1, dated
September 27, 1985.
Procedure MGP-04, Paragraph 6.1.a and 6.1.b
states:
"For each nonconformance determined by calculation to require the use
of interim stress criteria per Section 4.0, TED Facility
Engineering shall be notified as follows:
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(1) Bechtel shall notify by telephone the Director, Nuclear
Facility Engineering, or General Supervisor, Facility
Engineering.
(2) The notification by telephone shall be followed up by a
written confirmation.
The written confirmation shall
include the NCR number, specific deviation evaluated,
and the specific analyzed results. Modifications required
to meet design /SAR requirements will be included.
The
written confirmation shall also include a comparison of
the interim versus SAR allowables involved with the design
and the available safety margin."
During the inspection, the NRC inspector noted that the following
systems and pipe restraints required the use of interim stress
criteria (exceeded FSAR allowables):
System
Stress Calculation No.
Restraint No.
22F(C2)
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T-010A(C6)
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18B (C3)
- GCB-1-H13
- A-59
Hydrogen Dilution
119L
29-HBB-15-H7
Containment Sump
to ECCS
32E
- A-43
The failure to follow the approved procedure is a violation of
10 CFR 50, Appendix B, Criterion V (346/86004-01A).
Prior to the conclusion of the NRC inspection, the licensee
initiated a number of corrective actions which are included in
the following documents:
TED letter to Bechtel, "NCR Resolutions," advanced copy dated-
February 11, 1986; formal letter dated February 20, 1986.
The
letter requested a listing of all pipe support evaluations
where FSAR commitments were exceeded but interim requirements
were met.
Bechtel letter to TED, BT-16335, dated February 14, 1986
provided the list requested by TED.
Bechtel Interoffice Memorandum from Project Engineer to the
Group Supervisors, "NRC Violation - NCRs Interim /Short-Term
Allowables," dated March 15, 1986 provided additional
instructions and training for the performing of evaluations,
c.
Item 1.a(3) (Closed)
See RIII Inspection Report No. 50-346/85035, Paragraph 4.b.
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d.
Item 1.a(4) (0 pen)
The FCRs that could impact safety-related piping system operability
are listed in RIII Inspection Report No. 50-346/85035, Paragraph 4.b.
An update of this list (as of January 15, 1986) is as follows:
(1) FCR No.77-213, 77-398,80-221, and 80-276:
No Maintenance
Work Orders (MW0s) were issued for these FCRs.
Since related
modification work will not affect system operability, the MW0s
will be developed after restart.
(2) FCR 78-360: This FCR was voided.
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(3) FCRs78-126, 79-308,83-151, 85-086,85-010, 85-126,85-163,
85-176, and 85-224 were closed.
(4) Status of the remaining FCRs:
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General Description
FCR No.
(No. of Supports Involved)
Status- 79-421
AFW pump turbine modification (15)
Work completed,
in process of
being closed
out.
83-136~
Replace FW pump governor (2)
Work in progress
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,.83-138
Change of 14 valves (4)
Work in progress85-025
Motor driven FW pump (103)
Work in progress
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85-143
Relocate steam admission
Work in progress
valves (6)85-160
Work in progress
e.
Item 1.a(5) (0 pen)
The documentation in RIII Inspection Report No. 50-346/85035,
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regarding Paragraph 4.b remains unchanged.
5.
Status of Completion of IE Bulletin 79-14
RIII Inspection Report No. 50-346/85031, Paragraph 6.a, documents that
TED's ineffective utilization of the FCR system resulted in some support
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component rework not being completed as stated in a TED letter to RIII.
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In response to the RIII findings, TED reported in a letter to RIII (Serial
No. 1-598, dated December 20, 1985) that several FCRs which were originally
issued as a result of the I&E Bulletin walkdown were identified as still
being open, and for several of the work items which were originally
identified as closed, the work was not yet fully completed.
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Due to NRC questions concerning whether or not TED had made misleading
statements to the NRC, the NRC inspector discussed the matters with the
TED licensing and QA directors. Through their cooperation, the NRC
inspector obtained the following information to aid in his assessment
of the matter:
a.
TED Letters to RIII Documenting the Status of IEB 79-14
Implementation
Scope or Status
Serial No. (Date)
Connitment Date
of Work
1-137 (6/16/80)
3/1/81
Approximate 210 supports for
accessible areas
1-177 (12/20/80)
12/31/81
207 supports for accessible
areas
3/82 outage
15 supports for inaccessible
areas (only 60% analyses
completed)
1-187 (2/13/81)
3/82 outage
Modification will be made
for inaccessible areas
1-201 (5/22/81)
3/82 outage
45 supports for ipaccessible
areas
1-223 (11/13/81)
12/82
133 supports for inaccessible
areas
1982 outage
Remaining supports inside
containment
1-289 (8/19/82)
Next outage
11 support modifications
inside containment
1983
Remaining supports outside
containment
1-429 (5/31/84)
12/31/83
All supports " mechanically
completed"
b.
Causes of Support Modifications not Being Completed As Stated in TED
Letter 1-429
Actual Completion
Document No.
Support No.
Date
Causes
FCR 80-87
FSK-M-CCB-8-25-H
12/84
Bechtel failed
to provide TED
with modification
package
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Actual Completion
Document No.
Support No.
Date
Causes
FCR 80-87
FSK-M-HCD-15-18-H
9/85
Same as above
FCR 80-91
A 399/A 402
4/10/84
TED Licensing
(NCR393-79)
(instead of
oversight
Feb.'84)
NCR 396-79
EBD-12-19
4/30/84
TED Licensing
and EBD-12-20
(instead of
oversight
Feb.'84)
FCR 80-125
SR 33, 35, and
5/1/84
TED FMD
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oversight
FCR 80-125
SR37
3/30/84
TED FMD
oversight
FCR 80-125
A171
1/16/84
TED FMD
oversight
NCR 524
A8
11/85
TED FMD and
QC oversight
As a result of the review, the NRC inspector concluded that (1) from 1980
to 1982 TED reported the status of work as requested in IEB 79-14, (2) it
appears that substantial funds were spent to implement the actions set
forth in IEB 79-14 and IEB 79-02, and (3) despite the identified
deficiencies, less than 4% of the components were not modified as reported
in the TED letters to RIII.
No further action is planned by Region III
at this time.
6.
AFPTSS System Modification
Actions taken by TED to modify the AFPTSS system are discussed in RIII
Inspection Report No. 50-346/85035, Paragraph 6.
During the site
inspection of January 14, 1986, the NRC inspector performed a waltdown
of the AFPTSS crossover leg piping connecting steam generator No. 1-2 to
auxiliary feedwater pump turbine No.1-1.
This crossover leg, with its
many piping expansion loops, is located in the " Fan Alley" (area having a
large amount of HVAC equipment) on floor elevation 623'-0" inside the
auxiliary building.
a.
Inspection of Piping Restraints
An inspection of piping restraints and anchor modifications, to
evaluate their functionability, was conducted by the NRC inspector.
The piping section observed was between valve HV 106 and valve
HV 106A. No deficiencies were identified.
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b.
Questions Concerning High Thermal Lockup Stress
During the walkdown, the NRC inspector observed excessive restraining
within the region bounded by restraints 3A-EBD-19-H 104 and
3A-EBD-19-H 109.
This region involved four directional changes in
less than 22 feet of 6" diameter pipe. The restraints are:
H 104
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X and Y restraints
Z restraint
H 105
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H 106
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X restraint
H 107
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Y restraint
Z restraint
H 108
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H 109
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X and Y restraints
The primary reason for all the thermal expansion loops installed in
the crossover leg is to minimize piping thermal stresses.
The
placement of so many rigid restraints within the pipe loop is a
contradiction of this design intent.
In January 1986 the NRC inspector reviewed Bechtel piping stress
analysis Problem 40A, " Main Steam," Revision D2, dated September 25,
1985, and observed:
(1) a high ANSI B31.1 secondary stress (RIII
Inspection Report No. 50-346/85035, Paragraph 6.d.(2)(a) reported a
maximum ASME Section III secondary stress only), and (2) a piping
dimensional deviation at a critical area between rigid restraint
H 105 (Data Point 122) and pipe elbow (Data Point 124).
The Bechtel
analysis showed 25", the Bechtel Hanger Location Drawing HL-203J
showed 15", and the field remeasurement taken during the inspection
was 14".
During the NRC inspection the piping stress analysis was
rerun in a simplified configuration based on corrected dimensions.
The results were:
Data Point
Thermal and Existing /New
Code allowable
Existing /New
SAM Stress (psi)
% Change
(psi) reference
124/72
33,942/33,298
-1.9%
37,500
128/80
8,858/12,690
+43.3%
37,500
132/90
15,410/24,043
+56%
37,500
Subsequent to the review, the NRC inspector noted:
~(1) The present design met code requirements but system relief
should be provided.
(2) The present TED Procedure IP-M-002, "The Piping Support Inspec-
tion and Verification Program:
Verification of Support / Component
Location and Quantity," Revision 1, dated November 7, 1985 . states
" Denote all dimensional differences on the walkdown HL drawings."
The dimensional differences are then evaluated by Bechtel to
determine if there will be any effect on the pipe stress analyses.
The NRC inspector stated his view that priority should be given
to critical /high stressed areas.
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7.
Bechtel Design Control for HELB Analysis
A number of unresolved matters were raised during a previous NRC inspection
conducted at Bechtel on December 4-6, 1985 (RIII Inspection Report
No. 50-346/85035, Paragraph 6.e(4)).
A followup review conducted at
Bechtel on January 29-30 and April 2, 1986 revealed the following
conditions:
a.
The HELB analysis for the AFPTSS modification was based on desk top
design guides.
Bechtel Procedure, MGP-05," Procedure for
Interdiscipline Coordination for High Energy Line Break Evaluation
at Davis-Besse Nuclear Power Station Unit No.1," Revision 1. was
issued for use on March 14, 1986.
b.
The desk top design guide, " Pipe Whip Details," was determined to be
unacceptable by Bechtel upon reevaluation.
Among the 21 postulated
AFPTSS break locations, five indicated that the plastic hinges will
form at the third elbow instead of the second elbow.
However, new
whip restraints were not required due to the location of the impact
areas.
The impact areas are as follows:
Break No.
Area of Impact
6
ceiling
21
wall penetration
36
wall
38
wall
69
no safety-related equipment
A Bechtel QA Management Audit, No. 12501-05, issued QA Finding No. 1
on February 28, 1986, documenting a similar finding.
c.
Design tables utilized in the AFPTSS HELB component design were not
approved for the specific applications.
These design tables also had
not been evaluated for applicability.
Subsequent Bechtel evaluation
determined that the tables presented in RIII Inspection Report
No. 50-346/85035, Paragraph 6.e(4)(c), were conservative for the
application.
d.
The remaining two unresolved matters documented in RIII Inspection
Report No. 50-346/85035, Paragraphs 6.3(4)(d), and 6.e(4)(f) were
adequately addressed in MGP-05.
e.
The NRC inspector questioned whether or not all the previous Bechtel
HELB analyses were adequate.
Bechtel's response was as follows:
The original design criteria were very conservative.
The FSAR,
Revision 13, dated June 1975, states in Paragraph 3.6.2.2.2,
" Pipe Restraint Design Criteria to Prevent Pipe Whip Impact
Outside the Containment - The basic philosophy used to prevent
high energy pipes from whipping is to provide restraints of
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sufficient capacity and with such spacing that pipe whipping
cannot develop.
These restraints are independent of operating
and seismic supports.
Consequently, they are designated for
pipe rupture loads only.
Pipe rupture of either guillotine
or side-split type are postulated to occur in accordance
with Table 3-4b for the 36 inch main steam and 18 inch main
feedwater lines.
Allowable pipe spans are calculated, assuming
that the force developed during the accident experience is
transferred to the pipe restraint.
The pipe restraints are
then designed to withstand this force.
The restraints are
further designed to prevent the pipe from shearing off and
generating missiles.
Jet effects from ruptured pipe are
considered in designing pipes, walls, and shield.
The
,
concurrent effects of jets and pressure differentials are
also considered in designing walls and shields."
Bechtel reevaluated the Steam Generator Blowdown System modifica-
tion conducted in 1982.
Of the approximate 20 postulated pipe
break locations, one was found having a plastic hinge location
at the third elbow instead of the originally determined second
elbow.
Field inspection observed no safety-related equipment
in the force impact areas.
The NRC inspector considers the Bechtel response acceptable.
8.
Followup on Licensee Event Report (LER)
The inspector reviewed the TED LER 85019, "PORV Discharge Line
Overstressed Due to Inadequate Heat Trace," dated November 6, 1985, and
questioned the TED actions taken to identify and resolve the issue.
a.
TED Inspection Findings
During recent hanger reinspections the following damage and
deficiencies were identified on the PORV discharge piping:
Hanger No.
NCR No. (date)
Damages / Deficiencies
30 GCC-9-H10
0920 (10/3/85)
Concrete cracked; plate
separated from wall.
30 GCC-8-H6
0921 (10/3/85)
Concrete cracked and
fell; plate separated
from wall.
30 GCC-8-H5
0914 (10/4/85)
Concrete spalling.
30 GCC-8-H7
0822 (10/1/85)
Pipe clamp slipped off
and binded.
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Upon removal of the H10 and H6 baseplates for repair, the following
abandoned drilled anchor bolt holes were discovered.
Hanger No.
Abandoned Holes
H10
Three holes were outside the three
bolt diameter limit (will not
affect strength of installed bolts)
One hole was at the 2.25 diameter
limit.
Evaluation determined
the spacing to be acceptable.
H6
One hole was outside three bolt
diameter limit.
Chip void in concrete.
Evaluation
determined this void to be
acceptable.
Since two out of two baseplates that were removed for repair revealed
abandoned anchor bolt holes, the NRC inspector indicated to the
licensee that additional inspections of the areas behind other
baseplates in the vicinity of H10 and H6, including Hanger
6C-EBB-4-H12 (see Paragraph 9 for justification), should be
considered.
Depending on the hole / void configuration and relative
distance to the affected anchor bolts, these holes / voids could mean
reduction in support strength or fracturing of concrete at a loading
much lower than full load capacity.
The TED engineering department
did not share the NRC inspector's view.
TED's justification for not
inspecting for possible abandoned drilled holes existing in other
baseplates will be discussed further during a future inspection.
This is ar. unresolved item (346/86004-02).
b.
TED Correr_t13 e Actions
As a resul' ;f the TMI event, the NRC issued NUREG-0737 in 1980
requesting vtilities to reevaluate the pressurizer safety and relief
valve operations.
The Teledyne Engineering Services (TES) issued
Technical Report (TR), TR-5639-2, " Davis-Besse Analysis and
Evaluation of the Safety / Relief Valve Discharge System per NRC
NUREG-0737," Revision 0, dated January 1983, to address this issue.
This report indicated that the PORV loop seal temperature should be
maintained at 500 F.
As a result of the damaged supports observed
to date, the NRC inspector reviewed the operation records and
procedures and concluded that the TED action to ensure design
implementation was ineffective.
The bases for the determination were:
(1) From 1976 to 1979 there was a total of 82 PORV lifts where
the loop seal temperature was less than 500 F.
Of these, 46 were
assumed by TED to be 130 F.
The 46 lifts with the temperature
less than 400 F is in violation of Davis-Besse Periodic Test
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Procedure, PT 5164.03, " Pressurizer Relief Valve Heat Trace
Test," Revision 5, dated August 20, 1982, which indicates that
the test acceptance criteria for heat trace (T772) must be
above 400 F, to allow PORV (RC 2A) to be lifted 650 times and
25 times for a heat trace below 400 F.
This is a violation of 10 CFR 50, Appendix B, Criterion V
(346/86004-01B).
(2) Contrary to the conclusion stated in TES TR-5639-2, PT 5164.03
was not revised to reflect the latest loop seal temperature of
500 F and no attempt was made to inspect and evaluate the
condition of the piping and supports.
This is a violation of 10 CFR 50, Appendix B, Criterion XVI
(346/86004-03).
c.
TES Design Control
The NRC inspector reviewed the following TES reports docun.enting
their evaluations of effects on PORV inlet and discharge piping and
supports:
TR-5639-2, " Analysis and Evaluation of the Safety / Relief Valve
Discharge System per NUREG-0737," dated January 1983.
TR-6388-1, " Analysis of Davis-Besse, Unit 1 Pressurizer Relief
Line 400 F Loop Seal Blowdown," dated September 26, 1985.
TR-6388-2, " Davis-Besse Nuclear Power Station Reconciliation of
ASME Section III Evaluation of Class 1 Pressurizer Relief
Piping," dated November 15, 1985.
TR-6388-3, " Davis-Besse Nuclear Power Station Evaluation of
Class 3 Pressurizer Relief Piping," dated January 21, 1986.
Subsequent to the review, the NRC inspector had the following
concents:
(1) The maximum transient dynamic loading locations where piping
restraint danage was observed, differed from the TES analytical
prediction.
(2) Review of the reference lists revealed what appeared to be an
analytical bases that were either preliminary or interim in
nature.
(3) Several support design loads, based on the present dynamic
transient without loop seal (at 400 F subcooled water condition),
were many magnitudes higher than the original design with loop
seal. The present support loads should have been bounded by
the original design.
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_
_- _ _ _
,
_
_ _ _ _ . . _ _ _
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(4) The support design load combination for rigid restraints should
be the larger of (Thermal + Weight + SSE) or (Thermal + Weight
+ Blowdown).
The snubber design load should not consider the
thermal and weight loads.
Some of the TES support design loads
could not be verified using the above criteria.
(5) The PORV discharge modes of operation include:
(a) PORV opens on 2450 psig saturated steam.
(b) PORV opens on 2450 psig saturated steam followed
by a transition to subcooled water.
(c) PORV opens on 2450 psig 640 F subcooled water.
(d) PORV opens on 2450 psig 400 F subcooled water.
The present design is based on mode (d).
Based on the comment
stated in (3) above, it is not clear that the mode (d) support
loading will bound all modes of operation.
Further review of the subject matter is planned.
This is an
unresolved item (346/86004-04).
d.
PORV Operability Without Loop Seal
The TED decision to remove the PORV loop seal could affect PORV long
term operability due to a continuous steam leak (small amount) and
hydrogen attack of valve disc and seats.
TED letter (A85-30681) dated
August 30, 1985, to Crosby Valve and Gage Company, the PORV designer
and manufacturer, requested evaluation of this condition.
The Crosby
response was documented in a letter to TED, dated February 28,
1986.
The NRC inspector reviewed the Crosby letter and observed the
valve leak detection devices installed on the piping system.
The
measures taken to ensure system safe operation due to the design
modification were determined to be acceptable.
9.
Auxiliary Feedwater (AFW) Line Transient
During the inspection conducted inside the containment on February 6, 1986,
the NRC inspector observed concrete spalling and grout cracking at load
intensified locations on the baseplate for AFW Train 1-2 hanger
These conditions are indications of excessive transient
loads that may not been accounted for in the Bechtel design and analysis.
The NRC inspector's review of NCR 85-0198, issued on September 3, 1985, and
evaluated by TED engineering on October 5, 1985, identified no indication
that the causes and measures taken to correct and prevent recurrence had
been evaluated.
During the inspection conducted on February 19-20, 1986, the NRC inspector
further discussed his observations.
The TED engineer stated that the
concrete damage most likely occurred during construction.
The NRC
inspector re-entered the containment on February 20, 1986 for a closer
examination of hanger H12, and observed the following additional
adverse conditions:
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One of the anchor bolts appeared to be bent.
One of the baseplates was partially separated from the wall.
Markings that could have resulted from thermal displacement and
possible loads in the direction the bolt appeared to be bent, were
observed where the restraint contacted the pipe.
During the NRC inspection conducted at Bechtel on April 1, 1986, TED
presented the following documents to address the NRC inspector's
concern:
Bechtel letter to TED, BT-16470, " Evaluation of Support 6C-EBB-4-H12
Auxiliary Feedwater System," dated March 17, 1986.
This letter
documented the Bechtel inspection and evaluation.
Bechtel concluded
that there was no evidence of transient loads affecting the support.
Bechtel letter to TED, BT-16552, "NRC Inspection AFW System Dynamic
Transients," dated April 2, 1986.
This letter compared the steam
condensation with the check valve leak rate and concluded that a
severe dynamic transient will not occur.
The TED representative further committed to reinspect the piping system
including Support SC-EBB-4-H12 during the next refueling outage after
restart.
The NRC inspector reviewed the above documents and considered
the TED actions to be acceptable.
10.
TED Review of Grinnell Calculations
Due to the lack of a formal design interface control etween TED, Bechtel,
and Grinnell (now a Tyco company) and the fact that G 'nnell did not have
final design responsibility for the adequacy of the or., nal 4,000
(estimate) safety-related support calculations and subsequeat IEB 79-02
and IEB 79-14 evaluations, the calculations performed by Grinnell were
in question.
Since completion of the IEB 79-02 and IEB 79-14 work, some
of the Grinnell hanger calculations were replaced through the disposition
of Bechtel NCRs and FCRs.
The licensee is in the process of obtaining
all hard copies of the Grinnell original calculations and IEB 79-02 and
IEB 79-14 calculations so they can complete their review of this work.
The NRC inspector discussed the matter with the licensee and indicated
that the following areas warrant additional review:
a.
In January 1986 TED requested Grinnell transmit all Davis Besse 1
original hanger calculations and subsequent calculations for
IEB 79-02 and IEB 79-14 to TED.
No documents have been received to
date.
Further efforts are required to obtain these calculations if
they are available,
b.
Upon receipt of Grinnell calculations, TED will develop a program
to review these calculations to assure compliance with design
procedures.
The NRC inspector concurred with TEDS recommendation
that this work be completed prior to the next refueling outage.
17
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d
.
c.
The TED QA Department recently conducted an audit in the area of
design interface of vendors who provide engineering service to TED.
The NRC inspector plans to revier the audit report during a
subsequent inspection.
This is an unresolved item (346/86004-05).
11.
Unresolved Items
An unresolved item is a matter about which more information is required
in order to ascertain whether it is an acceptable item, an open item, a
deviation, or a violation.
Three unresolved items disclosed during this
inspection are discussed in Paragraphs 8.a. 8.c, and 10.
12.
Exit Interview
The NRC inspector met with licensee representative (denoted in Paragraph 1)
at the conclusion of the inspection.
The inspector summarized the scope
and findings of the inspection.
The inspector also discussed the likely
informational content of the inspection report with regard to documents
reviewed by the inspector during the inspection.
The licensee
representatives did not identify any such documents as proprietary.
18