ML20126B397
| ML20126B397 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 06/10/1985 |
| From: | Starostecki R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | Kenyon B PENNSYLVANIA POWER & LIGHT CO. |
| References | |
| NUDOCS 8506140092 | |
| Download: ML20126B397 (2) | |
See also: IR 05000352/1984048
Text
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Docket Nos. 50-387/50-388
10585
Pennsylvania Power & Light Company
ATTN: Mr. Bruce D. Kenyon
Vice President
Nuclear Operations
2 North Ninth Street
Allentown, Pennsylvania 18101
Gentlemen:
Subject: Inspection Report No. 50-352/84-48
This refers to the special team inspection, headed by Mr. R. Gallo, NRC Region
I, on August 27-30, 1984 at the Bechtel Corporation's offices at 50 Beal Street,
San Francisco, California. The inspection was conducted in response to allega-
tions related to structural and piping design activities performed by Bechtel
for Susquehanna, Limerick and Hope Creek. The findings of this special inspec-
' tion are presented in the enclosed NRC Region I Inspection Report No. 50-352/
84-48. The inspection findings were discussed with Mr. E. Hosterman of your
staff during the inspection.
The inspection team found that one of the allegations regardir1 Susquehanna was
substantiated, i.e., the pipe break location , analyzed for the main steam and
feedwater systems did not correspond to the respective locatiors described in
the FSAR. Although the team concluded that the actual analysis conservatively
envelopes the FSAR commitment, and therefore, no safety concern exists, you are
requested to respond to this letter within thirty days. In preparing your
response, you should address the apparent discrepancy between the actual
Bechtel analysis and the FSAR, and what steps you are taking to ensure that the
FSAR accurately reflects the analysis that was performed. The response is
exempt from the Office of Management and Budget's clearance procedures under
the Paperwork Reduction Act of 1980, PL 96-511.
Your cooperation with us is appreciated.
Sincerely,
%ned 4:
Richard W. Starostecki, Director
Division of Reactor Projects
Enclosure:
As Stated
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Pennsylvania Power & Light 2
Company
cc w/ encl:
Norman W. Curtis, Vice President, Engineering and Construction - Nuclear
A. R. Sat;ol, Manager, Nuclear Quality Assurance
W. E. Barberich, Licensing Engineer
H. W. Keiser, Superintendent of Plant
A. J. Pietrofitta, General Manager, Power Production Engineering and
Construction, Atlantic Electric
William Matson, Allegheny Electric Cooperative, Inc.
Public Document Room (PDR)
Local Public Document Room (LPDR) .
Nuclear Safety Information Center (NSIC)
NRC Resident Inspector
Commonwealth of Pennsylvania
R. J. Benich, Service Project Manager, General Electric Company
bec w/ encl:
. Region I Docket Room (with concurrences)
Senior Operations Officer (w/o encl)
DRP Section Chief
J. Grant, DRP
J. Durr, DRS
L. Bettenhausen, DRS
R. Bosnak, NRR
J. Rajan, NRR
M. Campagnone, NRR
Allegation File: RI-84-A-0093
RI-84-A-0099
6
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Stro dder Kister Sta s ecki
G clti Sh' f 6 \ (r
0FFICIAL RECORD COPY SUSQUEHANNA ALLEGATION - 0002.0.0
05/01/85
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g the nah UNITED STATES
g NUCLEAR REGULATORY COMMISSION
a S c E0loN I
E f $31 PARK AVENUE
KING OF PRUSSI A. PENNSYLVANI A 19406
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JW 0 71985
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Docket No. 50-352
Philadelphia Electric Company
ATTN: Mr. S. L. Daltroff
Vice President, Electric Production
2301 Market Street
Philadelphia, PA 19101
Gentlemen:
Subject: Inspection Report No. 50-352/84-48
This refers to the special inspection conducted by a team headed by R.M. Gallo
of this office on August 27-30, 1984, at Bechtel Corporation's offices at 50
Beal Street, San Francisco, California, of the activities authorized by NRC
Construction Permit No. CPPR-106, and to the discussion of our findings held by
Mr. Gallo with Mr. Walters of your staff at the conclusion of the inspection.
The inspection was conducted in response to allegations related to structural
and. piping design activities performed by Bechtel for Limerick, Susquehanna
and Hope Creek. The findings of this special inspection are presented in the
enclosed inspection report.
Based on the results of this inspection, none of the allegations pertaining to
Limerick were substantiated, nor were any violations identified. One issue,
relating to the Limerick Feedwater Check Valve Slam Analysis, was identified.
Although the inspection concluded that the integrity of the feedwater piping
was not a concern, certain discrepancies as described in Section 4.5 of the
enclosed report were identified related to the analysis conducted by Bechtel.
You are requested to respond to this issue within thirty days of the date of
this letter. In preparing your response, you should address the correction of
the discrepancies and the steps you are taking to ensure that similar
analyses conducted by your contractor incorporate as-built plant information.
The response is exempt from the Office of Management and Budget's clearance
procedures under the Paperwork Reduction Act of 1980, PL 96-511.
In accordance with Section 2.790 of the NRC's " Rules of Practice," Part 2,
Title 10, Code of Federal Regulations, a copy of this letter and your reply
will be placed in the Public Document Room.
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. Philadelphia Electric Company 2
Your cooperation with us_is appreciated.
Sincerely,
O ,
P
Richard W. arostecki, Director
Division of Reactor Projects
Enclosure: Insepetion Report No. 50-352/84-48
cc w/ encl:
V. S. Boyer, Senior Vice President, Nuclear Power
John S. Kemper, Vice President, Engineering and Research
G. Leitch, Station Superintendent
Troy B.. Conner, Jr., Esquire (Receives All 2.790 Information)
Eugene J. Bradley, Esquire, Assistant General Counsel
Limerick Hearing Service List
Public Document Room (PDR)
Local Public Document Room (LPDR)
Nuclear Safety Information' Center (NSIC)
-NRC Resident Inspector
Commonwealth of Pennsylvania
- - . . . . _ - , - - _ _ . , - . _ __ , , -- -.
~ Operations
. 50-352 24
_ ,$,-, Lirerick Hearing Service List
Judge Helen F. Hoyt Mr. Marvin I. Lewis
Atomic Safety and Licensing 6504 Bradford Terrace
Board Philadelphia, PA 19149
U.S. Nuclear Regulatory
Commission
Washington, D.C. 20555
Judge Richard F. Cole Phyllis Zitner
Atomic Safety and Licensing LEA
Board P. O. Box 761
U.S. Nuclear Regulatory Pottstown, PA 19464
Commission
Washington, D.C. 20555
Judge Jerry Harbour Docketing and Service Station
Atomic Safety and Licensing Office of the Secretary
Board U.S. Nuclear Regulatory Commission
U.S. Nuclear Regulatory Washington, D.C. 20555
Commission
Washington, D.C. 20555
Mr. Frank R. Romano Counsel for NRC Staff
61 Forest Avenue Office of the Executive Legal Director
Ambler, Pennsylvania 19002 U. S. Nuclear Regulatory Commission
Washington, D.C. 20555
Mr. Robert L. Anthony Philadelphia Electric Company
P. O. Box 186 ATTN: Edward G. Bauer, Jr.
103 Vernon Lane ~'
Vice President,and
Moylan, Pennsylvania. 19065 General Coun. sal
2301 Market Street.,'a.
.. Philadelphia, -PA 1~9$b1
.. : --
David Wersan, Esq. T
Charles W.1Elliott, Esquire
Assistant Consune-r Advocate Brose and Postwistilo
Office of, Consumer Advocate 1101 Buildinc T
1425 Strawberry Square lith anc Northamston Street,
Harrissurg, PA 17120 Easton, PA 18042
, Steven P. Hershey, Esquire Zori G. Ferkin '2.
-
' CoedurW.y Legal Services , Inc . Governor's Energy, Council
law Genter West P. O. Box 8010 7
5219 Chestnut Street Harriseurg, PA 17105
PNGadel onia , PA 19139
- t ~; .,_. . ,
Martha W;irBusn, ,Escui re Troy B'. tonn'sn. JFfFEscui.re
' Kathryn $,..Lewi s , Escuire Mark J. Weindrann, Esc.uire
' Municipal 5 Services 51dg. Conner &<Wetieshnnn^ 1 -
15th anc JFK Elvd. 'i157;PE.6nsylva'nia Avenue
Philaceiphia, PA 19107 Washington, D. C. 20006
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5'-352
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,, Angus Love, Esquire Robert J. Sugarman, Esq.
g'- , 101 East Main Street Sugarman, Denworth & Hellegers
N, Norristown, PA 19401 16th Floor Center Plaza
101 North Broad Street
Philadelphia, PA 19107
~
Spence W. Perry, Esquire Mr. Joseph H. White, III
Associate General Counsel 15 Ardmore Avenue
Federal Emergency Management Agency Ardmore, PA 19003
500 C Street, S.W. Room 840 l
Washington, DC 20472
Thomas Y. Au, Esquire )
Assistant Counsel i
Commonwealth of Pennsylvania l
DER
505 Executive House
P. O. Box 2357
Harrisburg, PA 17120
Thomas Gerusky, Director
' Bureau of Radiation Protection
Department of Environmental '
Resources
5th Floor, Fulton Bank Bldg.
Third and Locust Streets
Harrisburg, PA 17120
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No. 50-352/84-48
Docket No. 50-352
License No. CPPR-106 Category B
Licensee: Philadelphia Electric Company
-2301 Market Street
Philadelphia, PA 19101
Facility Name: Limerick Generating Station Unit 1
Inspection At: Bechtel Corporation Offices
San Francisco, California
Inspection Conducted: August 27-30, 1984
NRC Personnel: S. K. Chaudhary, Senior Resident Inspector, Limerick
S. Kucharski, Reactor Engineer
P. D. Milano, Vendor Inspector, IE
C. P. Tan, Structural Engineer, NRR
J. Rajan, Mechanical Engineer, NRR
Reviewed By: Mk b
R. M. Gallo, Chief, Reactor Projects
b
dat'e
BE
Section 2A, Team Leader
Approved by: 77)Lilk/1/ //*f 65 P.5
date
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S./J. Collins ( Chief, Projects Branch
No. 2
Inspection Summary:
A special announced inspection by two Region-Based Inspectors, two NRR Techni-
cal Reviewers, one IE Vendor Program Branch Inspector and one Region-Based
Supervisor of allegations related to structural and piping design activities
performed by Bechtel Engineering for the Limerick Generating Station. The
inspection involved 180 hours0.00208 days <br />0.05 hours <br />2.97619e-4 weeks <br />6.849e-5 months <br /> at the Bechtel ' offices in San Francisco and 20
,
hours onsite by the NRC inspectors and reviewers.
Results: Of the two major areas inspected, one issue regarding the Limerick
Feedwater Check Valve Slam Analysis was identified which was not relevant to
r the_ specific allegations. (Para. 4.5). Part of one allegation was substan-
tiated, but no safety concern was identified. (Para. 4.4)
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DETAILS
1. Persons Contacted
PEC0
J. S. Kemper, Vice President; Engineering & Research
J. Corcoran, Head, Field QA Branch
H. R. Walters, Resident Project Manager, San Francisco
G. Szonntagh, Engineer
Bechtel Power Corporation
H. Hollinghause, Manager of Engineering
R. E. Jagels, Chief, Mechanical Engineer
G. Ashley, Mechanical Analysis Group Engineer
C. Soppet, Project Manager
R. Schlueter, Assistant Project Engineer
A. Wong, Group Leader, Civil Engineering
G. Duncan, Nuclear Group Leader
H. Safwat, Mechanical Analysis Group Supervisor
E. R. Nelson, Manager, QA Division
In addition to the above, the inspectors interviewed and held discussions
with many more members of engineering and management staff of PECO and
Bechtel during the course of inspection.
2. Background
On June 28, 1984, NRC Region V office in Walnut Creek, California, re-
ceived an allegation that structural design problems, involving blast
loads of the Reactor Building south stack and strength calculations for
the 201 and 217 Reactor Building elevations, exist at Limerick. Respon-
sibility for follow-up on this allegation was transferred to Region I on
June 29, 1984.
On July 19, 1984, another allegation was received by Region V that the
pipe break forcing functions per the RELAP Code used in analyses for
Limerick, Susouehanna, and Hope Creek were inadequate anc did not conform
to FSAR commitments. Responsibility for follow-up on this allegation was
. transferred to Region I on July 20, 1984.
Both allegers were contacted by Region I staff for additional information.
On July 18, 1984, NRC representatives met with the first alleger at the
Region V office. On July 23, 1984,. Region I representatives spoke, via
telephone, with the second alleger. Based on the preliminary information
obtained, Region I conducted a one-week inspection at the Bechtel
San Francisco Office with assistance from NRR and IE Vendor Inspection
Branch. (Bechtel is the Architect / Engineer (A/E) for Limerick, Susquehanna
and Hope _ Creek).
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The inspection included a review of (1) the structural design applicable
to-the Limerick Reactor Building and (2) the pipe break analyses used at
Limerick with sampling inspection done on Susquehanna and Hope Creek in
order to address all the concerns of the second alleger.
3.0 Structural Design Allegations
3.1 Allegation - South vent stack was not designed to resist blast loads.
The first alleger's primary concern, as understood by Region I, was:
"The Reactor Building south stack was not designed to resist blast load
due to sudden air compression resulting from a nearby railroad accident.
Also, the design calculations were in error."
3.1.1 - Scope of Inspection
The inspection was directed to ascertain the technical ' validity of
the allegation, to determine if designers had engaged in any improper
or inadequate design process, and to assess, if the allegation were
valid, the impact of this error on the safety of the plant operations
and the safety margin in the plant as a whole. The effort to pursue
the inspection objective consisted of the following:
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visual inspection of the south stack for its relationship to
other plant structures, and its "as-built" geometry.
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review of engineering calculations to determine technical
adequacy and validity of ' design approach and basic assumptions.
--
review- of technical specifications, design and construction
drawings, and referenced codes and stancards.
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review of construction. and quality control procedures for
installation. o
-- discussion with cognizant engineering and management personnel.
3.1.2 References
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LGS South Stack Calculations:
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No. Title Date Revision
21.7 South Stack Truss 11/2/78 Original
8/30/84 8
21.7.2 NRC Inquiry on Reactor 8/25/84 0
Building South Stack Due
to Blast Load (68 sheets)
21.5~ Precast Panel Design - 8/22/80 0
Reactor Building (175 sheets)
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Field Changes ,and Supporting Calculations:
No. Title Date
C-9668F Reactor Building #2 South Stack 7/28/82
Truss; Calc. #21.7, Revision 6,
Sheet 34-1
C-9505F Platforms Elev. 332'- 2 3/8" through 5/13/82
378'- 9", South Stack; Calc. #21.7,
Revision 6, Sheet 34-7
C-1059F Reactor Building #2 South Stack 8/31/83
Truss; Calc. #21.7, Revision 6,_
Sheets 34-8 to 34-10
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Bechtel Specification 8031-A-1, " Specification for Furnishing,
Fabrication and Delivery of Precast Concrete."
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Bechtel Specification 8031-G-41, " Specification for Furnishing,
Detailing, Fabrication and Delivery of Structural Steel for the -
Reactor Building and Control Complex Super Structure and Rad
Waste Building."
--
Bechtel Meeting Notes; Document Control No. 182634, dated
5/16/84.
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Limerick Generating Station " Fire Protection Evaluation Report"
(FPER).
--
Bechtel Drawings:
C-484; C-660, Rev. 16
C-667, Rev. 10; C-791, Rev. 4
C-795, Rev. 15; C-845, Rev. 14
C-847, Rev. 12;
QAD-108, Rev. 16; QAD-109, Rev. 10
QAD-110,'Rev. 9,-QAD-111, Rev. 9
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3.1.3 Conduct of Inspection
a'. Visual Inspection of the South Stack
The Limerick Senior Resident Inspector visually examined the
south stack for any obvious defects in material, construction,
or workmanship. The "as-built" configuration was also compared
with design and construction drawings. The orientation and
geometry of the stack was reviewed to ascertain its exposure to
the postulated blast load from a nearby railroad car accident.
This-inspection was carried out at the Limerick site.
b. Review of Engineering Calculations
The review of calculations and other pertinent documents was
carried out at Bechtel's Home Office in San Francisco,
California. The inspectors reviewed the calculations and held
discussions with cognizant engineering personnel to assess the
validity and technical adequacy of calculations, design assump-
tions, and approach. The original design calculations for the
structural steel framing for the south stack were performed in
early 1978. The calculations were further upgraded and expanded
to include precast concrete panels in 1980. The complete design
of the south stack was also subjected to an in depth interdis-
ciplinary group review in May 1984. This review is documented
in Bechtel meeting notes (Document Control No. 182634) of May
16, 1984.
The inspectors also reviewed additional calculations for the
south stack that were performed by Bechtel, in response to this
NRC inquiry, to verify the validity of the original calculations.
This effort was completed by the design engineers at the time of
inspection, but all documents were not formally approved. The
approval of all additional calculations was accomplished during
the inspection period.
c. Review of Technical Specifications, Design and Construction
Drawings, and Referenced Codes and Stancards
The inspectors reviewed the applicable design and construction
specifications, and other documented requirements pertinent to
the design and construction of the south stack. The specifica-
tions were reviewed to determine if they contained adequate
technical requirements, evoked pertinent codes and standards
. directly or by reference, and 'were sufficiently detailed and
}" free of ambiguities to convey the technical - requi rements. The
drawings were examined to determine if they contained sufficient
information and adequate details to permit acceptable construc-
tion / installation and inspection. The referenced codes and
industry standards were reviewed to assess pertinence 'and ap-
plicability to the functional objectives of the construction /
installation of the south stack.
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3 .1 ~. 4 Findings
Based on the above reviews of documentation, and discussions with the
licensee and A/E, the inspectors determined the following:
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The south stack is not a safety-related structure as defined in
the Limerick-Project Q-List.
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The south stack is analyzed and designed as a seismic category
IIA structure as defined in Section 3.2.1 of the LGS-FSAR.
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Because the south stack is not Q-listed and is designated
seismic category IIA, it is not required to withstand blast
overpressure load.
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The south stack is not required for post-accident monitoring
because HVAC exhaust to this stack containing accident effluents
is automatically isolated (LGS-FSAR, Section 11.5.2.2.2).
On the basis of the above findings, the inspectors examined the
technical validity and adequacy of the south stack design-basis and
assumptions.
The south stack is located on the south side of the reactor enclosure
building between column lines 21.5 and 24.5 (east-west), and C and D
(north-south). The stack is designed to meet the requirements of
seismic category IIA (FSAR Sec. 3.2.1). The stack is a tall tubular
passage created by structural steel framing enclosed by precast
concrete panels attached to the structural members by high strength
bolted connections. The structural framing itself is attached to the
reactor enclosure south wall through welded connections to steel
embedments in the wall. The principal code used in the design of the
structure is AISC for framing and - connections and ACI for precast
concrete panels. The precast panels are doubly reinforced on both
faces, and are six and one-half inches (6h") thick.
Because the stack is not safety related, and its failure in a seismic
event, in itself, will not jeopardize the safe shut-down of the
plant, it is not required to withstand other than normal structural
loads. However, due to the adjacent location of the diesel generator
building that is safety related, the stack has been analyzed and
designed as a seismic category IIA structure.
3.1.5 Conclusions
The stack is not required to withstand the blast overpressure load
because its' failure due to blast overpressure will not affect the
safe shut-down of the plant. The allegation was not substantiated.
No violations were identified.
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3.2 Allegation - Deficiencies in the Design of Concrete Floor Slab of
Reactor Enclosure
This allegation as understood by the NRC consisted of several subparts as
follows:
a. The amount of reinforcing steel used in the floor slabs around the
containment structure is not in conformance with ACI 318-71 Code,
Section 10.5 1, specifically the formula (P min =200f fy),
b. The distribution of the reinforcing steel is not in accordance with
the requirements in Section 10.6 of ACI 318-71 Code.
c. The shear reinforcement requirements as contained in Section 11.1.1
of ACI 318-71 Code are not complied with,
d. The bundling and lap splicing of rebar is not in conformance with ACI
318-71 Code.
Due to the close interrelationship of the above subparts, they were re-
viewed as one concern.
3.2.1 Scope of Inspection
The inspection effort was directed to ascertain that proper design
techniques and assumptions were made, and that requirements of the
code were correctly interpreted and/or applied, and that the slab, as
designed and constructed, would fulfill the safety function it was
designed to perform.
The above inspection, objectives were pursued by review and examina-
tion of documentation, discussions with cognizant engineering per-
sonnel, and an evaluation of the existing design of the slab. Addi-
tionally, the adequacy and validity of design bases and associated
assumptions used in the design and analysis were also evaluated.
3.2.2 References
LGS - Preliminary Safety Analysis Report (PSAP)
LGS - Final Safety Analysis Report (FSAR)
Quality Control Inspection Reports (OCIR) for preplacement
,
inspections of Reactor Building floor-slabs at Elv. 201 and 217.
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Bechtel Design Drawings:
C-122, Rev. 22, Reactor Building Unit 1 Floor Plan,
Elevation 217' 0", Area 11
C-123, Rev.17, Reactor Bldg. , U-1, Elv. 217, Area 12
C-126, Rev. 25, Reactor Bldg., U-1, Elv. 217, Area 15
C-127, Rev. 23, Reactor Bldg., U-1, Elv. 217, Area 16
C-113, Rev. 21, Reactor Bldg. , U-1, Elv. 201, Area 11
C-114, Rev. 17, Reactor Bldg., U-1, Elv. 201, Area 12
C-117, Rev. 26, Reactor Bldg. , U-1, Elv. 201, Area 15
C-118, Rev. 22, Reactor Bldg. , U-1, Elv. 201, Area 16
Bethlehem Steel Drawings:
8031-C-39-140-2, 134-2, 115-2, 116-2, 191-2, 191 A-2, 173-3, 173A-3,
174-3, 174A-3, 169-2, 166-2, 163-2, 163A-2, 158-2, 158A-2
Bechtel Drawings:
'C-601, Rev. 31
C-602, Rev. 25 Project Civil Standards
. C-606, Rev. 9
Bechtel Calculations
VOL. FILE NO. SHEET NO. TITLE
(Calc. No.)
24 23.3 1 thru 12, 12-1, Reinforced Concrete
13 thru 19, 19-1, Slab design at.
20 thru 25, 25-1 El. 201 - 0",
.
thru 25-4, 26, Reactor Building
26-1 thru 26-3,
27 thru 50, 50-1
thru 50-26, 51
thru 53, 53-1,
54 thru 63, 63-1
thru 63-6, 64
thru 68, 68-1, 69
thru 73, 73-1 thru
73-2, 74, 75, 75-1
thru 75-6, 76, 77
(Total = 127 sheets)
24 23.5 1 thru 10 _ Recap of reinforce-
plus attachment ment requirement of
(Total = 11 sheets) slab edge at reactor
building
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3.2-3
. Conduct of Inspection
a. Visual Inspection
The inspector visually examined floor slabs for general confor-
mance with design and for any obvious defects in workmanship.
The _"as-built" configuration of the slabs was compared with de-
sign and construction drawings. This inspection was carried out
at the Limerick site.
b. Review of Engineering Calculations
The review of engineering analysis and design calculations was
performed at Bechtel's Home Office in San Francisco, California.
The inspector reviewed calculations and held discussions with
cognizant engineering personnel to determine the technical
validity and adequacy of design approach, design assumptions, j
applicability of codes and standards, and the governing con-
struction drawings and specifications.
The referenced codes and standards were also reviewed and
evaluated to assess their applicability- to the functional
objectives of the r, labs in the Peactor Building.
3.2.4 Findings
a. Description of Reactor Building
The Reactor Building (RB) is a reinforced concrete structure,
designed as a shear wall building. It is a rectangular building
about 324' long, 138' wide and 238' in height. Most outside
walls are-three feet in thickness, and are. continuous around the i
building. The interior walls are intermittent. There are about '
nine floors in the building; some of which are continuous around
the containment structure (drywell/wetwell), and others extend
to limited areas. A one-inch gap separates the RB floors from
the containment wall.
The RB floors consist of a concrete floor slab resting on a
horizontal steel framework consisting of girders, beams and
joists. In some areas, the concrete floor slab is interrupted
by steel grating. The slab thicknesses range from 12" to 36".
The concrete floor slab and the supporting steel members are
built as a composite structural element by the use of shear
connectors. Metal decks are used to support the slab during
construction. The concrete slab is rigidly conne'cted to the RB
concrete wall where the wall and slab meet.
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b. Design Approach and Assumptions
In the design of shear wall buildings it is conservatively
assumed that the horizontal earthquake load is resisted by the
walls parallel to the direction of load. The distribution of
load in the walls is proportional to the rigidities of these
walls at a horizontal crossi section of the building. The floor
system acts as a diaphragm to enhance the rigidity of the shear
walls, and is not considered to resist any horizontal earthquake
load by itself. Because horizontal earthquake loads imposed on
the floor system will be transmitted to the shear walls, the
purpose of designing a concrete slab for horizontal load is to
ensure that the slab will act as a diaphragm when transferring
load.
In view of the above, the following are the major simplifying
assumptions used by Bechtel in the design:
1. The horizontal earthquake loads are resisted by the
concrete walls in the direction of the earthquake forces.
2. The floor system consisting of the floor slab and
horizontal steel framework is designed mainly for vertical
loads.
3. In designing the floor system as a diaphragm, only the
concrete floor slab is taken into consideration, and the
horizontal steel framework and metal decking is neglected.
4. The floor slab is idealized as a beam with fixed or simply
supported end conditions depending on the actual support
condition of the slab, with the horizontal earthquake
forces acting parallel to the plane of the slab.
c. Design and Analysis
In view - of the foregoing design assumptions, the inspectors
reviewed the analysis and the associated design-calculations to
assess technical, procedural, and analytical details, and to
evaluate the design output specified in drawings and specifica-
tions for construction.
To simpli fy the analysis, it was assumed by 3echtel that in
transferring the earthquake load to the shear walls, the load
will be resisted o_nly by the concrete floor acting as a dia-
phragm even though the floor-slab is a part of a composite
structural system with the supporting steel members.
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the horizontal earthquake load will be resisted only by the
concrete floor even thought the floor-slab is a part of a
composite structural system with the supporting steel
members.
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the floor-slab is a beam with the thickness of the slab as
the width of the beam.
--
depending on the boundary conditions, the beam may be
fixed, or fixed at one end and simply supported at the
other.
--
the horizontal load acting on this idealized beam is the
product of the sum of the unit weight of vertical wall sec-
tion plus the unit weight of floor system, and the hort-
zontal acceleration due to an earthquake at the floor under
. consideration.
Based on the above, bending moments at different sections of the
. beam are computed, and the required area of reinforcing steel is
determined. Because there are openings of various sizes in the
floor, additional reinforcement around these openings is also
determined.
The above observations are based on the review of documented
analysis, a random verification of computations during the
review process, and extensive discussions with structural
engineers at Bechtel involved in the design and analysis,
d. Review and Evaluation of Code Requirements
In order to ascertain the applicability of any requirement of
the code, the intent of the code provisions must be established.
The minimum reinforcement requirement in ACI 318-71 Code Section
10.5 is to prevent sudden failure due to too low a steel ratio
'
.
in a flexural member. The provision in Section 10.6 of the ACI
Code is to prevent the formation of large concentrated cracks in
r
flexural members. The provision in Section 11.1 is to increase
the ductility, that is, to prevent sudden failure of flexural
-
members. As indicated in preceding sections, the floor slab is
designed for vertical loads and does not carry any load in the
horizontal direction, not even its own weight. Its function, to
act as a diaphragm to distribute the horizontal earthquake load,
is incidental. Moreover, the reinforcing steel designed for
vertical loads and placed orthogonally at the top and bottom of
the floor slab will enhance its ductility and resistance to
cracking when the floor slab is subjected to horizontal earth-
quake load. Disregarding the contribution from composite
- action of the supporting steel framework and the metal deck
acting as permanent framework for the floor slab is conservative,
,
e. Bundling of Reinforcing Steel Bars and Lap Splice Length
.The inspectors reviewed the QC documentation of inspections, and
rebar placement / detail drawings (Bethlehem Steel), and compared
__
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.
.
12
them to corresponding Bechtel design drawings to verify the con-
formance of rebar detailing to the design and code requirements.
The major part of this inspection and review was performed in
conjunction with the review of floor-slab design due to its
close interrelationship to each other. The inspector observed
that there was adequate instruction provided on Drawing C-601,
detail 8, regarding lap splices and other details to design,
fabricate and install, and verify the splices in installed
rebars. The installation conformance was verified and documented
by QC on Quality Control Inspection Reports (QCIRs). The in-
spector also verified that there were very few bundled rebars
(not exceeding 3 bars to a bundle) in the floors in question,
and they met code requirement for bundling.
3.2.5 Conclusions
On the basis of the above examination, review, discussions with en-
gineers and independent evaluation of available data, it is concluded
that:
a. the ACI Code 318-71, Sections 10.5, 10.6, and 11.1.1 provisions
are not applicable for the design of the RB floor-slab against
horizontal earthquake loads.
b. the _ bundling and lap' splicing of rebars in floor-slabs are in
compliance with the project specifications and Code requirements.
This allegation was not substantiated. No violations were identified.
4.0 Pipe Break Analysis Allegations
4.1 Allegation - Analyses were performed with incorrect revisions of
isometric drawings
Through discussions and interviews, the second alleger's first concern as
understood by Region I was:
" Analyses were performed uslng original revisions of isometric drawings
when there_were many later revisions available".
4.1.1 Scope of Inspection
The inspection effort was directed to ascertain the technical
validity of the allegation, to determine if the design staff has
performed the proper or improper analysis, and to assess, if the
allegation was valid, the impact of this error on the safety of the
plant. The effort to pursue the inspection objectives consisted of
the following:
--
Review of the methodology of the plant design staff.
_
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.
13
--
. Review of the calculations performed to determine the location
and design of the different types of restraints.
--
Discussions with cognizant engineering and management personnel.
4.1.2 References
--
Limerick -
Pipe break valve operability -
Inside Containment. Calculation No. S/8031/P-013.
--
Limerick -
Pipe break dynamic analysis for isolation valve-
Operability of Main Steam Outside Containment. Calculation No.
S/8031/PB-001, Revision 2.
--
Limerick -
Pipe Break Dynamic Analysis for Energy Absorbing
Catacomb (EAC) design for Main Steam Lines A.& B outside
containment. Calculation No. S/8031/PB-031, Revision 0.
--
Limerick -
Pipe Break forcing function calculation for Main
Steam Lines. Calculation No. 25-76.
4.1.3 Conduct of Inspection
The inspector reviewed Bechtel's methods of analyses for the location
and design of pipe whip restraints and valve operability restraints.
There are three groups involved in the process which include plant
design staff, plant design project, and the civil group. The plant
design staff is responsible for the analysis of Class 1 piping. The
plant design project is responsible for the analysis of Class 2 and ~3
piping, and the civil group, based on the results from the other two
groups, is responsible for locating and designing the pipe whip and
valve operability restraints.
4.1.4 Findings
Bechtel's civil group varied the process of waiting for the results
of the analysis that was to be performed by the other groups (design
staff, and plant design project) and performed its own hand calcula-
tion (which after considerable investigation, was found to be con-
servative) and designed and located the restraints as needed in the
original and later revisions of the pipiq; isometrics. Once the
other groups completed their analysis, the civil group would verify
and make appropria e changes where needed.
4.1.5 Conclusions
Although the procedure for the calculational method of designing and
locating the pipe whip restraint and valve operability restraints was
somewhat unorthodox, because of the order in which the analysis was
,
+, --
..
'
14
performed, the final check performed by the various groups involved
was assurance that no problems existed. The allegation was not
substantiated.
No violations were identified.
4.2 Allegation - Analysis was performed with incorrect Computer Code
The alleger's concern in this area as understood by Region I is as follows:
"For the Hope Creek project, the forcing function calculations were
performed by the plant design project staff using the ' Jet' Code which is
-not the proper tool to use."
4.2.1 Scope of Inspection
The inspection effort was directed to ascertain the technical
validity of the allegation, to determine if the plant design project
staff has performed the proper or improper analysis, and to assess,
if the allegation were valid, the impact of this error on the safety
of the plant. The effort to pursue the inspection. objective
consisted of the following:
--
Review of the methodology of the plant design project staff.
--
Review the purpose of the " JET" computer code
--
Discussion with cognizant engineering and management personnel.
4.2.2 References
--
Hope Creek - HPCI System Inside Containment Pipe Whip Design.
Calculation No. 625-10Q
--
F. J. Moody, " Prediction of Blowdown and Jet Thrust Forces",
ASME Paper 69 HT-31, August 6, 1969.
--
Standard Review Plan 3.6.2, " Determination of Rupture locations
and Dynamic Effects Associated with the Postulated Rupture of
Piping"
--
Branch Technical Position MEB 3-1, " Postulated Rupture Locations
in Fluid System Piping Inside and Outside Containment".
--
Hope Creek - JET impingement effects. Calculation No.12-128,
Revision 0.
--
Hope Creek Pipe Whip Restraint and Isolation Valve Operability
File No. 1085.
h
._.
4
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15
4.2.3 Conduct of Inspection
The inspector reviewed the plant design project staff role in
the process for analyzing the forcing function. The plant
design project staff is responsible for the forcing function
calculations for Class 2 and 3 piping. In their analysis of jet
impingement effects, two methods are used. 'First, Bechtel
performs a hand calculation to determine the total impingement
force. acting on any cross-sectional area of the jet. . This
magnitude is equivalent to the jet thrust force as defined in
SRP 3.6.2, Subsection III.2.c(4), i.e.,
T = KpA
where
P = System pressure prior to pipe break
A = Pipe break area
K = thrust coefficient
For the thrust coefficient, Bechtel uses a " Moody multiplier".of
no less than 1.2 - for steam and water-steam mixtures (See SRP.
3.6.2, Subsection III.3.f.), and a factor not greater than 2.0
for subcooled, nonflashing water (See SRP 3.6.2, Subsection
III.2.c(4).). This method is in accordance with MEB.3-1 and SRP
3.6.2, and is considered conservative. Bechtel used this design
method for the majority of . pipe whip and valve operability
restraints.
When an interference problem occurs, due to cumbersome design
based on the conservative calculations, the staff will use a
.
second, more realistic method that involves computer code
calculations. The codes used in this method were RELAP4 ' and
REPIPE, which give a smaller forcing function that is still
within the design safety limit.
4.2.4 Finding
The JET code was never used by the plant ' design project staff for a
forcing function calculation for the design of the restraints.
4.2.5 Conclusion
The allegation was not substantiated. i
No violations were identified.
_
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16
1
.
4.3 Allegation - Inconsistent Pipe Break Analysis for Similar Plants
The alleger's concern as understood by Region I was:
"There has been many more complicated pipe break analyses performed for
Limerick than Susquehanna, for one similar piping system 14 break
locations were analyzed for Limerick where as only 6 locations were
performed for Susquehanna". The alleger does not understand how this
could be valid for systems so similar.
4.3.1 Scope of Inspection
The inspection effort was directed to ascertain the technical
validity of the allegation, to determine if the design staff has
performed the proper or improper analysis, and to assess, if the
allegation were valid, the impact of this error on the safety of the
plant. The effort to pursue the inspection objectives consisted of
the following:
--
Review of the methodology of the plant design staff
,
--
Review of the design and construction drawings
--
Review of the Stress report to determine usage factors.
--
Discussions with cognizant engineering and management personnel.
4.3.2 References
--
Limerick - Stress Report Calculation No. S/8031-1803
--
Susquehanna Unit 1 and 2 Calculation No. SR8856-1800
Appendix E.
4.3.3 Conduct of Inspection
The inspector reviewed the pipe location drawings for the main steam
system, main feedwater system, high pressure coolant injection
system, and the standby liquid control system (SBLC) for Limerick and
Susquehanna. All of the systems were primarily the same for both
plants except for the SBLC system. The SBLC system for Limerick -is
connected to the core spray .line whereas for Susquehanna it is
connected to the reactor vessel.
4.3.4 Findings
For each piping system a stress analysis is performed. Based on the
stress analysis a number of break locations have to be analyzed. A
break location is designated by a usage factor. This factor is based
on the location .of piping and stress due to loads. For values
greater than 0.1, a break location is designated for that region of
pipe.
__.
Gm
r _
.
. 17 .
4.3.5 Conclusion
Due to the similarity of Limerick and Susquehanna, the same number of
breaks were analyzed for the majority of systems. Based on the
differences in the location of the SBLC system (even though the SBLC
systems are similar), more locations had to be analyzed for Limerick
because of the greater number of break locations that were identified
based on the stress analysis report.
The allegation was not substantiated.
No violations were identified.
4.4 Allegation - RELAP Analyses
The alleger's concerns as understood by Region I were:
(a) "Susquehanna FSAR Section 3.6.1,
entitled " Postulated Piping
Failures in Fluid Systems", and the accompanying figures specify
the pipe break locations that should have been analyzed;
however, the calculations done by Bechtel do not correspond with
the locations committed to in the FSAR."
A similar allegation was stated relative to the Hope Creek
project. The allegation as understood by the NRC was:
"For Hope Creek, of the systems listed in FSAR Table 3.6-1, pipe
break analyses were done on only five systems. The other
analyses had not been done as of April 198c."
(b) "The main steam and feedwater lines for Susquehanna were
originally done with hand calculations but, subsequently, RELAP
analyses were carried out for these systems. For the HPCI
system no blowdown forces were calculated." The Standby Liquid
Control System was also mentioned as being unsatisfactory with
respect to pipe break analyses.
(c) "At Susquehanna, Hope Creek and Limerick, hand calculations
should not have been used for determining valve operability."
The alleger believes that, instead of hand calculations, RELAP
Code analysis should have been used because this analysis
identifies the frequency of vibration in the pipe so that cor-
rective valve supports can be determined. The alleger believes
that the hand calculations cannot trace a transient wave in the
pipe and, therefore, may cause a problem determining correct
valve supports.
4.4.1 Scope of Inspection
The inspection was directed to ascertain the technical validity of
the allegation to determine if designers had employed an appropriate
-
_
.
.
18
design process, and to assess, if the ' allegation were valid, the
-
I' impact of this error on the safety of the plant operations and/or the
' safety . margin in the plant as a whole. The effort to pursue the
inspection ob.icctiv'e consisted of the following: 1
l
--
Review Iof ' engineering calculations to determine technical
adequacy and' validity of design approach and basic assumptions.
--
Review. of technical specifications, design ~ drawings, . and
,
referenced codes and standards.
--
Discussion'hithcognizantengineeringandmanagement
personnel.
4.4.2 References
^'
L
-
- - -
Report No. NSC-1-78-027, Pipe Break Dynamic Analyses ~ for
L-
Mainsteam and Feedwater Lines at Susquehanna.
V ,
[ i
--
Stress Report No. S/8856/PB-049, Rev. O, Forcing Function Study
j; for RCIC (outside containment) System at Susquehanna.
f: b --
Strt:ss Report No. S/8856/PB-046, Rev. O, Pipe Whip Analysis with
EAC 'Cesign for. Mainsteam Lines at Susquehanna.
--
Susquehanna FSAR, Fig. 3.6.1A, Rev. 17.
--
Hope Creek FSAR, Table 3.6.1 and FSAR Section 3.6.2.
/r.
' ; JHope Creek FSAR Question 210.19.
--
. .,
Csiculations 101 File No. SR-114, Comparison of- GAFT Code with
- 6 RELAP 5 Calculations for the HPCI System outside Containment at
8 Susquehanna.
,-- Calculation No. 4001, HPCI Line's Susquehanna.
--
Stren Report SR 8856-18G9, Appendix E.
,
4.4.3 Conduct of: Inspection -
- /
The review of calcul'ations and other pertinent documents was carried
out in Bechtel? s df fice in. San Francisco, California. The group
reviewed. the 'ca'i cuTa tion s and held discussions with cognizant
,
' engineering fersonnel to assess the valioity and technical acequacy
of calculations design assumptions and apprcach' . The group compared
the' pipe break locations identified in the FSAR Section 3.6.1 with
,
'
ji 'those actually analyzed in the pertinent strets report (First and
.w third references from Section 4.4.2).
J'
' * .#
r-
r .} f
a
__
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - - - - -
_ _ _ __ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ __ --
f ,
f [ 19
!-
}- The FSAR-was reviewed to determine if it contained requirements
relative to _ the method of pipe break analysis and whether those
requirements evoked pertinent codes and standards directly or by
reference and were sufficiently detailed and free of ambiguities. The
dynamic analysis procedures were reviewed for transient pipe break
l . loading to verify that postulated pipe breaks at intermediate
locations and terminal ends were in accordance with MEB-3-1 and valve
operability criteria.
4.4.4 Findings
f
Based on ~the above reviews of documentation and discussions with the
j_ licensee and A/E engineers, the group determined the following:
i
l With respect to allegation (a), the pipe break ' locations postulated
and analyzed for jet impingement. effects and valve operability for
the main steam and feedwater systems at Susquehanna do not correspond
n with the locations shown in Susquehanna FSAR Section 3.6.1 entitled
l' " Postulated Piping Failures in Fluid Systems" and the accompanying
I figures. However, - discussions _with the licensee's design engineers
[ indicated that the breaks were conservatively postulated at each
Rationale was also provided to justify
'
fitting and terminal . ends.
that the breaks analyzed in the stress reports envelope the loadings
for restraint design and jet impingement effects resulting from the
postulated breaks committed to in the FSAR.
.The group concluded that analysis of breaks at each fitting and at
terminal ends is an acceptable alternative procedure which conserva-
tively envelopes the FSAR commitment. The allegation concerning the
discrepancy between the pipe break locations shown in the Susquehanna
FSAR to those actually analyzed, is valid but alternative analyses,
discussed above, make the allegation moot.
The listing of systems thM are considered in the Hope Creek
Generating Station FSAR as critical in terms of High Energy Line
Breaks are listed in FSAR Table 3.6-1. This tabulation provides the
listing of rineteen systems that are within this category. Within
FSAR section 3.6.2, each of these systems is described with relation
to this condition, and the actions that have been taken to resolve
the issue.
USNRC Mechanical Engineering Branch requested from Becntel in its
FSAR Question number 210.19 identification of any unrestrained
whipping pipes located inside containment. The Bechtel response to
this question, which is described in the Bechtel Meeting, held on May
9, 1984, with NRC relative to FSAR questions, indicates that five
systems are in this category. On the final page of these meeting
notes, the signature showing acceptance for this question as
" acceptable-as-is" was verified.
_
_ _ _ _ . _ - - _ _ _ _ _ _ _ _
_
.
20
The pipe break analyses for the remaining systems were verified to be
available by selection of a sample of four systems listed in FSAR
Table 3.6-1. These systems were: (1) main steam; (2) reactor water
clean-up (RWCU); (3) core spray injection, and (4) emergency diesel
generator starting air. With the exception of the emergency diesel
generator starting air line, calculations for pipe whip and/or valve
operability restraints and line break analyses were available. For
the emergency diesel generator air line, FSAR Section 3.6.2 provided
the detailed justification for the lack of need for these analyses.
With respect to allegation (b), the loadings from postulated pipe
breaks for the main steam and feedwater lines were initially hand
calculated, using the largest break area and the operating pressure
in the line resulting in a conservative design of restraints. Sub-
sequently, the RELAP-5 Code was used for the same calculations, which
-
resulted in elimination of some restraints. The group did not find
anything unacceptable in this procedure. In regard to the HPCI
System, the group reviewed the blowdown calculations (fifth and sixth
reference in Section 4.4.2) and found that the blowdown forces had
been properly calculated.
The review of the .alculations for the pipe break analyses for the
standby liquid contre 1 system revealed no inconsistencies or non-
compliance with applicable guidelines and standards. The group,
therefore, concluded that allegation (b) stated in paragraph I above
is invalid and without technical merit.
Relative to allegation (c), the group determined that the A/E's
approach on Limerick, Susquehanna and Hope Creek with regard to pipe
whip and valve operability is to satisfy the requirements stated in
NRC Standard Review Plan 3.6.2, BTP MEB 3-1. During the pipe whip
event, the stress in the pipe, between the containment and isolation
valve due to dynamic loads, dead weight and pressure remains less
than 2.25 Sm (design stress intensity) for Class I or less than 1.8
Sh (allowable stress at maximum hot temperature) for Class 2 and
Class 3 piping. Between the isolation valve and the moment limiting
pipe whip restraints, the stresses are maintained low enough to
prevent the formation of a plastic hinge. The pipe stress at both
interfaces with the isolation valve is limited to 1.1 times the yield
stress, thus satisfying MEB 3-1 requirements.
The A/E performed a special set of calculations to evaluate various
hand calculated and RELAP forcing functions generated for valve oper-
ability analyses. Five reduced linear transient analyses were per-
formed using two detailed RELAP forcing functions and three simpli-
fied hand calculated forcing functions. Pipe stresses and restraint
loads obtained from these analyses were compared. Differences in the
input forcing functions and differences in the results obtained were
evaluated. The Limerick RCIC piping system was used in the forcing
o
e
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21
function study. The results of these calculations demonstrated that
the hand calculated (1.26 Pressure x Area) step fcrcing function com-
pletely envelopes the results of all other forcing functions, and the
pipe stresses resulting from hand calculated forcing functions are
more conservative as compared to the stresses due to RELAP forcing
functions.
In addition, the licensee made a qualitative assessment of the ac-
celerations and frequencies resulting from valve operability calcu-
lations. The bending moment in the valve supports is inversely
proportional to the square of the excitation frequency in pipe and
directly proportional to the pipe acceleration. As the frequencies
associated with pipe whip are usually much higher than the frequency
content of a seismic event (for which valves are qualified), it was
concluded that for a given allowable bending moment at the valve
supports, higher. accelerations expected during a pipe whip event are
acceptable. The group concurs with the assessment and concludes that
, allegation (c) is without merit.
4.4.5 Conclusions
Allegation (a)
The inspectors concluded that although the pipe break location
analyzed for jet impingement and valve operability for the mainsteam
and feedwater systems at Susquehanna did not correspond with the FSAR
break locations, the actual analysis of breaks at each fitting and at
piping system ends was an acceptable procedure that conservatively
envelopes the FSAR commitment. Therefore, although_ the allegation
was substantiated no safety concern exists with regard to the
operability of the main steam and feedwater systems.
The inspectors did not identify any discrepancies relative to the
pipe break analyses listed in Hope Creek FSAR Table 3.6-1. There-
fore, it was concluded that this portion of the allegation was not
substantiated.
Allegation (b)
The inspectors did not identify any inconsistencies and/or noncom-
pliance with applicable guidelines and standards in the analyses and/
or calculations for HPIC or Standby Liquid Control System. Therefore,
it was concluded that the allegation was not substantiated.
Allegation (c)
The inspectors concluded that the Bechtel stress analyses and calcu-
lations were acceptable because the Bechtel hand calculations for
forcing functions are more conservative than calculations by the
RELAP comouter code. Moreover, the frequencies associated with pipe
r
.
o-
22
whip were much higher than the frequencies associated with a seismic
event. The inspectors, therefore, concluded that for a given allow-
able building moment at valve supports, higher acceleration expected
during a pipe whip event was acceptable. The allegation was not
substantiated.
No violations were identified.
4.5 Allegation - Feedwater Check Valve Slam
The alleger's concern as understood by the NRC is as follows:
"The feedwater system piping analysis for Susquehanna Steam Electric
Station ignored the effects of feedwater check valve slam. This problem
had been worked on by Sargeant and Lundy; but the problem (check valve
slam) was still unresolved."
4.5.1 Scope of Inspection
The inspection effort was directed to determine the validity of the
allegation, to determine ' if feedwater check valve slam had been
considered in the feedwater pipe break analysis, and to identify any
continuing unresolved problems. The inspection effort consisted of:
--
Review of appropriate documentation for the Susquehanna
Feedwater Check- Valve Slam Analysis.
--
Discussion with cognizant engineering and management personnel.
--
Review of the Feedwater Cneck Valve Slam Analysis for Limerick
Generating Station. (This review was conducted for comparison
of Bechtel activities for Susquehanna and Limerick. The
allegation did not mention any concerns relative to the Limerick
-Feedwater Check Valve Slam Analysis.)
4.5.2 References
--
Sargeant and Lundy Report - Susquehanna Steam Electric Station
Feedwater Check Valve Analysis - dated March 11, 1983.
--
NRC letter to Pennsylvania Power and Light Company, Feedwater
Check Valve Analysis, dated April 24, 1984.
--
Bechtel Report - Evaluation of Feedwater Containment Isolation
Check Valves for a Hypothetical Pipe Rupture Condition for
Limerick Generating Station, dated July 1983.
4.5.3 Conduct of Inspection
The inspector reviewed the Bechtel contractor analysis for the effects
of a postulated break of the feedwater system in a line outside con-
tainment. The analysis was conducted to verify (1) the integrity of
-.
.
.
.
23
the intact portion of the feedwater system from the containment iso-
lation valves to the reactor and (2) the integrity of the check valve
for performance of its containment isolation function. The analysis
for Susquehanna was conducted by Sergeant and Lundy for Pennsylvania
Power and Light Company (PP&L). The contractor report and responses
to NRC requests for additional information were provided to the NRC
as discussed in the referenced NRC letter to PP&L dated April 24,
1984.
For comparison,- the analysis of this corresponding postulated break
in the - feedwater system for Limerick was reviewed. The Limerick
analysis was performed by Bechtel and provided in the referenced re-
port dated July 1983. The Limerick analysis was reviewed to deter-
mine if _ integrity of both the feedwater check valve and the intact
portion of the feedwater piping were maintained.
4.5.4 Findings
The Feedwater Check Valve Slam Analysis for Susquehanna had been
conducted.by Sargeant and Lundy and subsequently submitted to the NRC
for review. No discrepancies were identified relative to the
Susquehanna analysis.
Based on the NRC . review of the conduct of the Limerick Feedwater
Check Valve Slam Analysis four discrepancies were identified. The
discrepancies identified were related to the analysis of the
integrity of the check valve. The discrepancies were:
a. The material properties assumed in the analysis were for SA-216;
however, the manufacturer's (Atwood and Morrill Co.) drawings
indicated the valve body and disc materials to be SA-352, Grade
LCB. This inconsistency in the analysis was not properly
reviewed and evaluated for acceptability.
b. The material properties used in the analysis were based on room
temperature; whereas, the system was designed for 425 F service
temperature. The higher service temperature resulting in lower
tensile properties of the material was not properly evaluated
and accounted for in the analysis.
c. No calculation or other documented evidence was available to
establish the integrity of hinge / hinge pin during valve closure
and disc / seat deformation at the time of seating.
d. The evaluation did not address the use of welded-in seat seal in
place of direct disc contact with the weir / orifice of the valve
body.
e
.
24
Licensee -review and revision of the analysis considering the noted
discrepancies is an unresolved item. (352/84-48-01). Based on NRC
review of the analysis and the margins with respect to ultimate
strain limits demonstrated by the analysis it was determined that the
structural integrity of the valve would be retained.
The Limerick analysis was reviewed to determine if the feedwater
piping would withstand the pressure surge resulting from the check
valve slam. No discrepancies were identified for this portion of the
report.
4.5.5 Conclusions
On the basis of the above review, it is concluded that:
a. The Susquehanna Feedwater Check Valve Slam Analysis identified
no unresolved problems and no discrepancies. The allegation was
not substantiated,
b. The corresponding Limerick Feedwater Check Valve Slam Analysis
included discrepancies identified by the NRC. The discrepancies
identified were not associated with any allegation.
5.0 Exit Interview
An exit interview was held with representatives of the Bechtel Corporation
on August 30, 1984 at the Bechtel Engineering Office in San Francisco,
California. The results of the inspection were discussed at that meeting.
One. issue pertaining to Limerick was identified during the inspection as
described in this report (p.23). The issue did not result from any
allegation but was discovered during comparison of Bechtel activities for
the three projects revieweo. Also, part of one allegation was substan-
tiated, but it was determined that no safety concerns existed (p.19).
This allegation pertained to Susquehanna.
. . . . ..