ML20126B457

Insp Rept 50-352/84-48 on 840827-30.No Violation or Deviation Noted.Major Areas Inspected:Allegations Re Structural & Piping Design Activities Performed by Bechtel

ML20126B457
Person / Time
Site:	Limerick
Issue date:	06/05/1985
From:	Collins S, Gallo R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20126B428	List: ML20126B424 ML20126B457
References
50-352-84-48, NUDOCS 8506140108
Download: ML20126B457 (24)

See also: IR 05000352/1984048

Text

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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

R. M. Gallo, Chief, Reactor Projects

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Section 2A, Team Leader

Approved by: (MULkh///tzd ~

6/5I85

S./J. Collins ( Chibf, Projects Branch date

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 Pechtel 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

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

PECO

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

LC. 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

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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 4

Limerick, Susquehanna, and Hope Creek were inadequate and 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:

--

visual inspection of the south stack for its relationship to

other plant structures, and its "as-built" geometry.

--

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 standards.

--

review of construction and quality control procedures for

installation.

--

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", S'outh 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).

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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 Nc. 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 Standards

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 requirements. 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.

--

The south stack is analyzed and designed as a seismic category

IIA structure as defined in Section 3.2.1 of the LGS-FSAR.

--

Because the south stack is not Q-listed and is designated

seismic category IIA, it is not required to withstand blast

overpressure load.

--

The scuth 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. l

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 (PSAR)

LGS - Final Safety Analysis Report (FSAR)

Quality Control Inspection Reports (QCIR) 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, 191A-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 engineeririg personnel to determine the technical

validity and adequacy of design approach, design assumptions,

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 applicabili_ty to the functional

objectives of the slabs in the Reactor 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

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 connected 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 cross-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 simplify the analysis, it was assumed by Bechtel that in

transferring the earthquake load to the shear walls, the load

will be resisted only 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'.

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depending on the boundary conditions, the beam may be

fixed, or fixed at one end and simply supported at the

other.

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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 hori-

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 Sectinn

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

. 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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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 veri fy the splices in installed

rebars. The installation conformance wr.s 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 using 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:

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Review of the methodology of the plant design staff.

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Review of the calculations performed to determine the location

and design of the different types of restraints.

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Discussions with cognizant engineering and management personnel.

4.1.2 References

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Limerick -

Pipe break valve operability -

Main Steam Line

Inside Containment. Calculation No. S/8031/P-013.

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Limerick - Pipe break dynamic analysis for isolation valve-

Operability of Main Steam Outside Containment. Calculation No.

S/8031/PB-001, Revision 2.

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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.

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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 piping isometrics. Once the

other groups completed their analysis, the civil group would verify

and make appropriate 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

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performed, the final check performed by the various groups involved

was assurance that no problems existed. The allegation was not l

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 valfd, :he impact of this error on the safety

of the plant. The effort to pursue the inspection objective

consisted of the following:

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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

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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".

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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

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File No. 1085.

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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.

No violations were identified.

_ _ .

. _ _ -

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16

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 I 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 . 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.

_

_ - _ _ - _ _ _

. -

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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 A_11egation - 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 1984."

(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

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.

18

design process, and to assess, if the allegation were valid, the

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 objective consisted of the following:

--

Review of 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 with cognizant engineering and management

personnel.

4.4.2 References

--

Report No. NSC-1-78-027, Pipe Break Dynamic Analyses for

Mainsteam and Feedwater Lines at Susquehanna.

--

Stress Report No. S/8856/PB-049. Rev. O, Forcing Function Study

i

forRCIC(outsidecontainment)SystematSusquehanna,

--

Stress Report No. S/8856/PB-046, Rev. O, Pipe Whip Analysis with

EAC Design 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.

--

Hope Creek FSAR Question 210.19.

--

Calculations 101 File No. SR-114, Comparison of GAFT Code with

RELAP 5 Calculations for the HPCI System outside Containment at

Susquehanna.

--

Calculation No. 4001, HPCI Line's Susquehanna.

--

Stress Report SR 8856-1800 Appendix E.

4.4.3 Conduct of Inspection

The review of calculations and other pertinent documents was carried

out in Bechtel's office in San Francisco, California. The group

reviewed the calculations and held discussions with cognizant

engineering personnel to assess the validity and technical adequacy

of calculations design assumptions and approach. The group compared

the pipe break locations identified in the FSAR Section 3.6.1 with

those actually analyzed in the pertinent stress report (First and

third references from Section 4.4.2).

.

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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

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

Based on the above reviews of documentation and discussions with the

licensee and A/E engineers, the group determined the following:

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

with the locations shown in Susquehanna FSAR Section 3.6.1 entitled

" Postulated Piping Failures in Fluid Systems" and the accompanying

figures. However, discussions with the licensee's design engineers

indicated that the breaks were conservatively postulated at each

r

'

fitting and terminal ends. Rationale was also provided to justify

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 that 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 nineteen 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 Bechtel 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.

.

I -

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.

_ _ _ . _ _ _ _ - _ _

. . . . . .. . . . . . . _ .._ .

. . _ . . - . . . . _ . . , . . - . -

.

. -

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20

N

l The pipe break analyses for the remaining systems were verified to be

! available by selection of a sample of four systems listed in FSAR

4 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 exceotion 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-

e

sequentiv, 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.

I-

The review of the calculations for the pipe break analyses for the

standby liquid control system revealed no inconsistencies or non-

compliance with applicable guidelines and standards. The group,

therefore, concluded that allegation (b) stated in paragraph 1 above

1

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 1 or less than 1.8

o 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

f

r

,

_ . _

_ _ _ _ _ ___ . _ _ _ _ _ _ _ _ .

.

.

21

function study. The results of these calculations demonstrated that

the hand calculated (1.26 Pressure x Area) step forcing 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)

i The inspectors did not identify any inconsistencies and/or noncom-

i al %nce with applicable guidelines and standards in the analyses and/

f

or calculations for HPIC or Standby Liquid Control System. Therefore,

i 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 r;: e conservative than calculations by the

RELAP computer code. Momver, the frequencies associated with pipe

__

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.

.

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 Check 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 3echtel 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

__

( .. _.

O

.

23 l

l

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

'ae 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 usumed 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.

._.

m

.

.

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 ult: mate

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 Feedwatar 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 reviewed. Also, part of one allegation was substan-

tiated, but it was determined that no safety concerns existed (p.19).

This allegation pertained to Susquehanna.

)

1

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l