IR 05000352/1984048
| ML20126G874 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek, Limerick, 05000000 |
| Issue date: | 06/10/1985 |
| From: | Starostecki R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | Martin T Public Service Enterprise Group |
| References | |
| NUDOCS 8506180172 | |
| Download: ML20126G874 (2) | |
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JUti 101985 Docket No. 50-354
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Public Service Electric & Gas Company ATTN: Mr. T. J. Martin -
Vice President Engineering and Construction 80 Park Plaza - 17C Newark, New Jersey 07101 Gentlemen:
Subject: Inspection Report No. 50-352/84-48 This refers to the special team inspection, headed by Mr. R. Gallo of NRC Region I, on August 27-30, 1984 at Bechtel Corporation's offices at 50 Beal Street, San Francisco, California. The inspection was conducted in response to <
allegations related to structural and piping design activities performed by Bechtel for Hope Creek, Limerick and Susquehanna. The findings of this special inspection are presented in the enclosed NRC Region I Inspection Report No' .
50-352/84-48. The inspection findings were discussed with Messrs. R. Kirk and F. Danbra of your staff during the inspectio Based on the results of this inspection, the inspection team found that none of the allegations pertaining to Hope Creek were substantiated, nor were any violations identified. One issue, pertaining to the safety relief valve dis-charge line analyses contained in the Hope Creek Plant Unique Analysis Report, is still under review. The results of that review will be forwarded when the review is complete.~
No response to this letter is required. Your cooperation is appreciate
Sincerely, M I*eds w my Richard W. Starostecki, Director Division of Reactor Projects
Enclosure:
As Stated
REGION I==
Report N /84-48
. Docket N License N 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 i
NRC Personnel: S. K. Chaudhary, Senior Resident Inspector, L..nerick S. Kucharski,-Reactor Engineer P. D. Milano, Vendor Inspector, IE C. P. Tan, Structural Engineer, NRR
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J. Rajan, Mechanical Engineer, NRR Reviewed By: Mk R. M. Gallo, Chief, Reactor Projects ( dat'e BE Section 2A, Team Leader Approved by: 77)ULk/2/1/tvf 65,N5 S. C. Collins ( Chief, Projects Branch date No. 2 Inspection Summary:
A'special announced inspection by two Region-Based Inspectors, two NRR Techni-
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cal Reviewers, one IE Vendor Program Branch- Inspector and one Region-Based
- Supervisor of allegations related to structural and piping design activities
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performed by Bechtel Engineering for the Limerick Generating Statio The
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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 reviewer 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 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 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 i Bechtel-during the course of inspectio . 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 Limeric Respon-sibility for follow-up on this allegation was transferred to Region I on June 29, 198 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, 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, 198 Both allegers were contacted by Region I staff for additional informatio 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 allege .0 Structural Design Allegations 3.1 Allegation - South vent stack was not designed to resist blast load 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 acciden Also, the design calculations were in error."
3. 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" geometr review of to determine technical
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engineering calculttions adequacy and validity of design approach and basic assumption review of technical specifications, design and construction drawings, and referenced codes and standard review of construction and quality control procedures for installatio discussion with cognizant engineering and management personne . References
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LGS South Stack Calculations:
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N Title Date Revision 2 South Stack Truss 11/2/78 Original 8/30/84 8 21. NRC Inquiry on Reactor 8/25/84 0 Building South Stack Due to Blast Load (68 sheets)
. 2 Precast Panel Design - 8/22/80 0 Reactor Building (175 sheets)
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Field Changes and Supporting Calculations:
N Title Date t C-9668F Reactor Building #2 South Stack 7/28/82 Truss; Calc. #21.7, Revision 6, Sheet 34-1 C-9505F Pla.foims 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 Trurs; 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."
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Bechtel Meeting Notes; Document Control No. 182634, dated 5/16/8 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; 0AD-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 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 drawing The orientation and geometry of the stack was reviewed to ascertain its exposure to
. the postulated blast load from a nearby railroad car acciden This inspection was carried out at' the Limerick sit 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
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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, 198 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 calculation This effort was completed by the design engineers at the time of inspectir'n, but all documents were not formally approved. The approval of all additional calculations was accomplished during the inspection perio Review of Technical Specifications, Design and Construction Drawings, anc Referencea 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 inspectio 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 stac ..
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3. 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-Lis The south stack is analyzed and designed as a seismic category IIA structure as defined in Section 3.2.1 of the LGS-FSA Because the south stack is not Q-listed and is designated seismic category IIA, it is not required to withstand blast overpressure loa 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 assumption 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 0 (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 i
bolted connections. The structural framing itself is attached to the i 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 (61")
s thic Because the stack is not safety related, and its failure in a seismic event, in itself, will not jeopardize the safe shut-cown 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 structur . 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 substantiate No violations were identified.
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1 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: 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 = 00ffy), The distribution of the reinforcing steel is not in accordance with the requirements in Section 10.6 of ACI 318-71 Cod The shear reinforcement requirements as contained in Section 11. of ACI 318-71 Code are not complied with, The bundling and lap splicing of rebar is not in conformance with ACI 318-71 Cod Due to the close interrelationship of the above subparts, they were re-
> viewed as one concer . 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 perfor 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 evaluate . References
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LGS - Preliminary Safety Analysis Report (PSAR)
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, 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 VO FILE N SHEET N TITLE (Calc. No.)
24 2 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 2 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 Visual-Inspection The inspector visually examined floor slabs for general confor-mance with design and for any obvious defects in workmanshi The'"as-built" configuration of the slabs was compared with de-sign and construction drawings. This inspection was carried out at the Limerick sit Review of Engineering Calculations The review of engineering analysis and design calculations was performed at Bechtel's Home Office in San Francisco, Californi The inspector reviewed. calculations and held discussions with cognizant engineering 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 specification The referenced codes and standards were also reviewed and
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evaluated to assess their applicability to the functional objectives of the slabs in the Reactor Buildin .2.4 Findings 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 wal The RB floors consist of a concrete floor-slab resting on a horizontal steel framework consisting of girders, beams and
. joist In some areas, the concrete floor slab is interrupted by steel gratin 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 connector 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 mee __
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. 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 loa 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
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ensure that the slab will act as a diaphragm when transferring loa 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
t concrete walls in the direction of the earthquake forces.
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! The floor system consisting of the floor siab and horizontal steel framework is esigned mainly for vertical load . In designing the floor system as a diaphragm, only the concrete ficor slab is taken into consideration, and the
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horizontal steel framework and metal decking is neglecte l The floor slab is idealized as a beam with fixed or simply supportad end conditions depending on the actual support condition of the slab, with the horizontal earthquake
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forces acting parallel to the plane of tne sla c. Design and Analysis j In view of the foregoing design assumptions, the inspectors j reviewed the analysis and the associated design calculacions to assess technical, precedural, anc analytical details, and - to evaluate the design output specified in drawings and specifica-
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To s in:pli fy the analysis, it was assumed by Bechtel tnat in transferring the earthcuake 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
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member he
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the floor-slab is a beam with the thickness of the slab as the width of the bea depending on the boundary conditions, the beam may be fixed, .or' fixed at one end and simply supported at the othe 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 consideratio ' 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 determine 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, Review and Evaluation of Code Requirements In order to ascertain the applicability of any requirement of the coce, the intent of the code provisions must be establishe 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 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 incidenta 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 loa Disregarding the contribution from composite action of the supporting steel framework and the metal deck acting as permanent framework for the floor slab is conservativ 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 requirement 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 bundlin . Conclusions On the basis of the above examination, review, discussions with en-gineers and independent evaluation of available data, it is concluded that: 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 I horizontal earthquake load ' the bundling and lap splicing of rebars in floor-slabs are in compliance with the project specifications and Code requirement This allegation was not substantiate No violations were identifie .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 unoerstood by Region I was:
" Analyses were performed using original revisions of isometric drawings when there were many later revisions available".
4. 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 staf .
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Review of the calculations performed to determine the location and design of the different types of restraint Discussions with cognizant engineering and management personne . References
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Pipe break valve operability - Main Steam Line Inside Containment. Calculation No. S/8031/P-01 Limerick - Pipe break dynamic analysis for isolation valve-Operability of Main Steam Outside Containment. Calculation N S/8031/PB-001, Revision Limerick -
Pipe Break Dynamic Analysis for Energy Absorbing Catacomb (EAC) design for Main Steam Lines A &B outside containmen Calculation No. S/8031/PB-031, Revision Limerick -
Pipe Break forcing function calculation for Main Steam Line Calculation No. 25-7 .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 restraint There are three groups involved in the process which include plant design staff, plant design project, and the civil group. The ' plant i 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 restraint .1.4 Findines 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 aesign project) and performed its own hand calcula-tion (which af ter 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 isometric Once the other groups completed their analysis, the civil group would verify and make appropriate changes where neede .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 existe The allegation was not substantiate No violations were identifie .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. 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 plan The effort to pursue the inspection objective consisted of the following:
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Review of the methodology of the plant design project staf '
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Review the purpose of the " JET" computer code
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Discussion with cognizant engineering and management personne . References
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Hope Creek - HPCI System Inside Containment Pipe Whip Desig Calculation No. 625-100
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F. J. Moody, " Prediction of Blowdown and Jet Thrust Forces",
ASME Paper 69 HT-31, August 6, 196 Standard Review Plan 3.6.2, " Determination of Rupture locations and Dynamic Ef fects 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".
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Hope Creek - JET impingement effect Calculation No.12-728, Revision Hope Creek Pipe Whip Restraint and Isolation Valve Operability File No. 108 _
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'4. Conduct of Inspection i
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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.,
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P = System pressure prior to pipe break
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A = Pipe break area
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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 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 i
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 i
' calculation The -codes used in this method were RELAP4 and REPIPE, which give a smaller forcing function that is still within the design safety limi ! 4. Finding
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' The JET code was never used by the plant design project staff for a
forcing function calculation for the design of the restraints.
i 4. Conclusion l The allegation was not substantiated.
- No violations were identified.
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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 simila . 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:
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Review of the methodology of the plant design staff
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Review of the design and construction drawings
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Review of the Stress report to determine usage factor Discussions with cognizant engineering'and management personne . References
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Limerick - Stress Report Calculation No. S/8031-1803
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Susquehanna Unit 1 and 2 Calculation No. SR8856-1800 Appendix . Conduct of Inspection The inspector reviewed the pipe location drawings for the main steam system, main feedwater system, high pressure coolant injection system, Susquehann and the standby liquid control system (SBLC) for Limerick and All of the systems were primarily the same for both plants except for the SBLC syste The SBLC system for Limerick is connected to the core spray line whereas for Susquehanna it is connected to the reactor vesse . 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 i
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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4. Conclusion Due to the similarity of Limerick and Susquehanna, the same number of 4 breaks were analyzed for the majority of system 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 repor The allegation was not substantiate No violations were identifie .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 projec 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 system 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 system For the HPCI system no blowdown forces were calculated." The Standby Liquid Control System was also mentioned as being unsatisf actory with respect to pipe break analyse (c) "At Susquehanna, Hope Creek and Limerick, hand calculations should not The allegerhave been used for determining valve operability."
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 pipe and,hand calculations therefore, cannot may cause a trace a transient wave in the valve support problem determining correct 4. 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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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:
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Review of engineering calculations to determine technical adequacy and validity of design approach and basic assumption Review of technical specifications, design drawings, and referenced codes and standard Discussion with cognizant engineering and management personne . References
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Report N NSC-1-78-027, Pipe Break Dynamic Analyses for Mainsteam and Feedwater Lines at Susquehann . Stress Report No. S/8856/PB-049, Rev. O, Forcing Function Study for RCIC (outside containment) System at Susquehann Stress Report No. S/8856/PB-046, Rev. O, Pipe Whip Analysis with EAC Design for Mainsteam Lines at Susquehann Susquehanna FSAR, Fig. 3.6.1A, Rev. 1 Hope Creek FSAR, Table 3.6.1 and FSAR Sect'on 3. Hope Creek FSAR Question 210.1 Calculations 101 File No. SR-114, Comparison of GAFT Code with RELAP 5 Calculations for the HPCI System outside Containment at Susquehann Calculation No. 4001, HPCI Line's Susquehann Stress Report SR 8856-1800, Appendix . 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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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 criteri . 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 fitting and terminal end 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 t,etween the pipe break locations shown in the Susquehanna FSAR to those actually analyzed, is valid but alternative analyses, discussed above, make the allegation moo The listing of systems that are considered in the Hope Creek Generating Station FSAR as critical in teres of High Energy Line Breaks are listed in FSAR Table 3.6- This tabulation provides the listing of nineteen systems that are within this categor 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 issu 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 categor On the final page of these meeting notes, the signature showing acceptance for this question as
" acceptable-as-is" was verified.
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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 analyse 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 restraint 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 procedur 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 calculate 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 I above is invalid and without technical meri 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- 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 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 requirement The A/E performed a special set of calculations to evaluate various hand calculated and RELAP forcing functions generated for valve oper-ability analyse 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
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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 function 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 meri .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 system The inspectors did not identify any discrepancies relative to the pipe break analyses listed in Hope Creek FSAR Table 3.6- There-fore, it was concluded that this portion of the allegation was not substantiate 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 substantiate 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 computer cod Moreover, the frequencies associated with pipe
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whip were much higber than- the frequencies associated with a seismic event. The inspectors, therefore, concluded that for a given allow-able building moment atjvalve supports, higher acceleration expected during .a~ pipe whip event was acceptable. The allegation' was not substantiate '
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No violations were identifie L 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. 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 problem The inspection effort consisted of:
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Review of appropriate documentation for the Susquehanna Feedwater Check Valve Slam Analysi ,
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Discussion with cognizant engineering and management personne Review of the Feedwater Check Valve Slam Analysis for Limerick Generating Station. (This review was conducted for comparison of Bechtel activities for Susquehanna and Limeric The allegation did not mention any concerns relative to the Limerick Feedwater Check Valve Slam Analysis.)
4. References
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Sargeant and Lundy Report - Susquehanna Steam Electric Station Feedwater Check Valve Analysis - dated March 11, 198 NRC letter to Pennsylvania Power .and Light . Company, Feedwater Check Valve Analysis, dated April 24, 198 Bechtel Report - Evaluation of Feedwater Containment Isolation Check Valves for a Hypothetical Pipe Rupture Jondition for Limerick Generating Stati,on, dated July 198 . Conduct of Inspection t
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 a
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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 l to NRC requests for additional information were provided to the NRC .
as discussed in the referenced NRC letter to PP&L dated April 24, l 198 For comparison, the analysis of this corresponding postulated break in the feedwater system for Limerick was reviewe 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 maintaine .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 revie No discrepancies were identified relative to the Susquehanna analysi 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: 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 LC This inconsistency in the analysis was not properly reviewed and evaluated for acceptability, 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 analysi 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 seatin 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.
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l 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 retaine 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 repor . Conclusions On the basis of the above review, it is concluded that: The Susquehanna Feedwater Check Valve Slam Analysis identified no unresolved problems and no discrepancies. The allegation was not substantiate ' The corresponding Limerick Feedwater Check Valve Slam Analysis included discrepancies identified by the NRC. The discrepancies identified were not associated with any allegatio .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,
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California. The results of the inspection were discussed at that meetin 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 reviewe Also, part of one allegatior was substan-tiated, but it was determined tha't' no ' safety concerns existed (p.19).
This allegation pertained to Susquehann .
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