September 18, 1997

SES 97-276

 

 

Mr. Robert Nicholls

Quality Assurance Manager

GE Nuclear Energy

175 Curtner Avenue

San Jose, CA 95125-1088

Dear Mr. Nicholls

Enclosed is a copy of Audit G-97-120 which documents the results of the Audit conducted at the San Jose, CA facility on August 18 through 22, 1997. The Audit resulted in identifying nine Findings and three Unresolved Items to GE Nuclear Energy (GENE) in which one deficiency was identified as a Level I Finding, seven deficiencies as Level II Findings, and one deficiency as a Level III Finding. In addition, this audit resulted in the identification of one Level II finding to the ComEd BWR Site Engineering Departments and Nuclear Fuel Services (NFS) Department.

The Level I finding encompasses the other eight deficiencies and reflects an overall assessment that GE Nuclear Services is ineffectively implementing its Quality Assurance Program in the design area. Due to the extensive nature and severity levels of the subject audit findings, ComEd Supplier Evaluation Services (SES) issued a Stop Work to GENE Nuclear Services for the safety related engineering and design activities performed at San Jose, CA for all ComEd BWR Stations (Dresden, Quad Cities, and LaSalle County) on August 29, 1997.

GENE will be required to provide a written response to the Findings and the Unresolved Items by October 10, 1997. (CAR Nos. G 97-120-01 through 08 and CAR Nos. G-97-120-10 through 13). The subject CARs were transmitted to the GENE QA Manager on September 8, 1997 for the purpose of an expedited corrective action.

The cooperation and assistance extended to the Audit Team by GENE personnel during the Audit was greatly appreciated.

September 18, 1997

SES 97-276

Mr. Robert Nicholls

Page 2

 

 

If you have any questions please contact Oscar Shirani, ComEd Supplier Evaluation Services (SES), at Tel.No. 630-663-7934.

Sincerely,

 

 

Oscar B. Shirani, PE

Audit Team Leader

OS:jkw/g:\qv\97audits\G-97-120.rpt

cc: J. Hosmer

L. Waldinger

S. Perry- Dresden

W. Subalusky - LaSalle

E. Kraft

D. Winchester - Dresden

J. McDonald - LaSalle

F. Famulari - Quad Cities

G. Poletto - LaSalle

R. Renuart

R. Freeman - Dresden

J. Hutchinson - Quad Cities

W. Pearce - Quad Cities

T. Rieck

E. Netzel

Audit File G-97-120

Duplicate File

Memorandum

 

 

Date: September 18, 1997

SES-97-276

To: Mr. J. Hosmer

Engineering Vice President

From: Oscar Shirani

Subject: ComEd Audit of GENE - G-97-120

 

Enclosed is a copy of Audit G-97-120 which documents the results of the Audit conducted at the San Jose, CA facility on August 18 through 22, 1997. The Audit resulted in identifying nine Findings and three Unresolved Items to GE Nuclear Energy (GENE) in which one deficiency was identified as a Level I Finding, seven deficiencies as Level II Findings, and one deficiency as a Level III Finding. In addition, this audit resulted in the identification of one Level II finding to the ComEd BWR Site Engineering Departments and Nuclear Fuel Services (NFS) Department.

The deficiency issued to ComEd Engineering was for the failure to utilize the Nuclear Design Information Transmittal (NDIT) process to transfer design information to GENE.

Due to extensive nature and severity levels of the subject audit findings, ComEd Supplier Evaluation Services (SES) issued a Stop Work to GENE Nuclear Services for the safety related Architect/Engineering and design activities performed at San Jose, CA for all ComEd BWR Stations (Dresden, Quad Cities, and LaSalle County) on August 29, 1997.

ComEd BWR Site Engineering Departments and the Nuclear Fuel Services (NFS) Department will be required to provide a written response to the Finding CAR No. G-97-120-09 by October 10, 1997. The subject CAR was transmitted to ComEd’s Engineering Vice President on September 8, 1997 for the purpose of an expedited corrective action.

 

September 18, 1997

SES-97-276

Mr. J. Hosmer

Engineering Vice President

Page 3

 

If you have any questions please contact Oscar Shirani, Supplier Evaluation Services (SES), at ext. 8-347-7934 Downers Grove.

cc: L. Waldinger

S. Perry- Dresden

W. Subalusky- LaSalle

E. Kraft- Quad Cities

D. Winchester- Dresden

J. McDonald- LaSalle

F. Famulari- Quad Cities

G. Poetto- LaSalle

R. Renuart

R. Freeman- Dresden

J. Hutchinson- Quad Cities

W. Pearce - Quad Cities

T. Rieck

E. Netzel

Audit File G 97-120

Duplicate File

 

ComEd

A Unicom Company

Audit Report

for

 

Special Audit G-97-120

 

GE Nuclear Energy

175 Curtner Avenue

San Jose, CA 95125-1088

 

 

 

 

 

 

 

 

 

 

Prepared By: __________________________________ Date:__________________

Oscar B. Shirani, PE

Audit Team Leader

 

Approved By: __________________________________ Date:__________________

Edward R. Netzel, PE

Supplier Evaluation Services Director

 

 

 

Vendor Name / Address

GE Nuclear Energy

175 Curtner Avenue

San Jose, CA 95125-1088

Audit Dates

August 18-22, 1997

Audit Purpose and Scope

GE Nuclear Energy (GENE) was contracted by ComEd to perform various safety related Architect/Engineering activities. The focus of this Audit was to review the implementation of their quality program for providing engineering services to ComEd.

The focus of this audit was to validate potential similar technical and design control issues that were identified during the Dresden ISI for the engineering activities that GENE is providing to ComEd.

The primary scope of the audit was the design control requirements of 10CFR 50 Appendix B as they apply to engineering services and specifically safety related calculations. This was a performance based audit which reviewed the activities and documentation of GENE QA program in the areas of order entry, design control, design interfaces, software control, technical adequacy of design documents, corrective action identification and internal audits in the design area.

Implementation of all the general Quality Assurance program elements, such as procurement, material control, fabrication, calibration, test/inspection, document control, organization, nonconforming items/part 21, external audits and records was not reviewed during this audit.

Quality Programs

The following GENE Quality Assurance Programs was in place at the time of this Audit:

GE Nuclear Energy Quality Assurance Program Description NEDO-11209-04A Revision 8, March 31, 1989

 

References

10CFR50 Appendix B - Criterion II, III, XVI, and XVIII

ANSI N45.2

GE Nuclear Energy Quality Assurance Program Description NEDO-11209-04A Revision 8, March 31, 1989 and various Engineering Operating Procedures

 

Checklist

An augmented NUPIC Revision 7 checklist was utilized during the performance of the Audit.

Program Effectiveness Statement

GENE has established a QA Program based upon the requirements of ANSI N45.2, NRC Regulatory guide 1.28, and 10CFR50 Appendix B. This audit was focused on the design control process with emphasis on a rigorous technical review of calculations. The audit also evaluated the GE/ComEd interface with regard to the transfer of design information.

The audit examined fifteen (15) General Electric Nuclear Services Design Record Files (DRF’s). The audit team performed a rigorous technical review of fifty four (54) calculations contained within the fifteen (15) G.E. DRF’s. There were nine findings and three unresolved items identified against GE in the audit. These results revealed a general lack of formality in the documentation supporting the design review process. Also technical questions regarding assumptions, references, design inputs, software verification/validation and insufficient detailed analysis were raised during the audit. Due to the quantity and nature of issues identified, the independent design review process was deemed ineffective. Furthermore, the QA independent oversight was found to be ineffective at reviewing the design portion of the DRF’s. Consequently ComEd Supplier Evaluation Services (SES) issued a finding to GENE Nuclear Services for not effectively implementing its QA Program in the area of design. Also ComEd SES issued a stop work order on August 29, 1997, to GENE Nuclear Services for safety related Engineering and Design activities performed for ComEd BWR stations.

In order to facilitate an expedited corrective action review, the ComEd Corrective Action Records (CAR’s) were transmitted to GENE on September 8, 1997. ComEd Problem Identification Forms (PIF’s) were issued to address the technical issues raised from the audit. The ComEd BWR site engineering organizations, ComEd Nuclear Fuel Services (NFS), and ComEd Nuclear Engineering Services (NES) reviewed the issues with GE for potential operability. No operability issues were identified.

Specific details regarding the evaluations performed during the audit and the types of deficiencies identified for each calculation can be found in the Audit Summary Section of this report. More detail regarding all of the deficiencies can be found in the Audit Findings Section of this report and in the attached Corrective Action Records (CAR’s).

 

 

ComEd Definition & Classification of Finding and Unresolved Item

 

Definition

Finding - a condition affecting the safety and/or reliability of the unit(s) or an identified noncompliance with regulations which requires corrective action.

Classification

Level I - a condition which does affect the safety and/or reliability of the unit(s) or a significant breakdown in the QA Program.

Level II - a condition which may affect the safety and/or reliability of the unit(s) or a major noncompliance in the QA Program.

Level III - a condition that probably does not affect the safety and/or reliability of the unit(s) but is a substantive deviation from implementing procedures.

Unresolved Item: A condition that is not a finding, may be an opportunity for improvement; or if left unaddressed, could potentially result in an unacceptable condition. This item may require an evaluation and response by the auditee or may require further audit investigations to determine its status.

Audit Findings:

One Level I, Seven Level II, one Level III Findings and three Unresolved Items were issued to GENE, and one Level II was issued to ComEd Engineering as a result of this audit.

 

Finding Level II (CAR G 97-120-01)

Numerous administrative and editorial errors were found in GENE design documents. Examples of these errors include document legibility, page numbering, record identification, changes made improperly and suitable identification of the preparer & reviewer. These discrepancies reveal a lack of formal control in the GENE design control process.

Finding Level II (CAR G 97-120-02)

Due to the numerous design control deficiencies being identified during this audit, the GENE independent design review process was determined to be ineffective.

Finding Level II (CAR G 97-120-03)

Numerous GENE calculations were found to have design control deficiencies such as unjustified assumptions, references lacking, design input errors, and inadequate detailed analysis.

Finding Level III (CAR G 97-120-04)

Design Record Files (DRFs) had missing contractual agreements as required per GENE procedures.

Finding Level II (CAR G 97-120-05)

ComEd Engineers performed and reviewed design analysis calculations under GENE QA Program without being employed and indoctrinated to GENE procedures.

Finding Level II (CAR G 97-120-06)

GENE's internal audits are ineffective in independently overviewing the design analysis area.

Finding Level II (CAR G 97-120-07)

Computer software frequently used at GENE lacked evidence of being verified and validated.

Finding Level II (CAR G 97-120-08)

GENE was provided design input data by Sargent & Lundy for ComEd projects without a formal design interface

 

Finding Level II (CAR G 97-120-09)

ComEd Engineering failed to use the Nuclear Design Information Transmittal (NDIT) process to transfer information to GENE. Letters or facsimiles were utilized to transmit design input documents to GENE calculations.

Unresolved Item (CAR G 97-120-10)

GENE NEDE-31744, Procedure No. 10-27 needs to reference 10CFR50 Appendix B since its scope includes safety related work.

Unresolved Item (CAR G 97-120-11)

Documentation for computer programs were unavailable for review during the audit.

Unresolved Item (CAR G 97-120-12)

The cognizant engineer was unavailable to answer questions regarding specific design documents.

Finding Level I (CAR G 97-120-13)

GENE has not effectively implemented its Quality Assurance Program in the area of design.

 

 

AUDIT SUMMARY

Introduction

ComEd has contracted with GENE to perform safety related Architect/Engineering services.

Order Entry/Specifications:

A review of the GENE incoming orders revealed that the incoming orders are processed into the GENE system through the use of Engineering Operating Procedures EOP-25-2.00 rev. 3, EOP 42-10.00 rev. 7, and GENE Policies & Procedures (P&P) NEDE-31746 Procedure No. 10-27 issued 9/94.

Procedure 10-27 requires that the responsible business manager ensures that a completed, reviewed and approved proposal authorization and/or proposal change authorization as appropriate is in place to support a proposal prior to issuance. The proposal shall adequately describe the product/service and delivery schedule requirements specified by the customer. Before entering into any contract, the QA Manager shall assess the GENE ability to comply with applicable QA Program requirements of NEDO-11209-04A, revision 8. Evidence of the purchase order acceptance reveals the translation of the customer’s purchase order, utility/plant, conversion of marketing record on ISIS number, approvals from appropriate level of management, and other related information.

The ComEd purchase orders address the GENE’s proposal correspondences. The Design Record File (DRF) is intended to be the formal controlled information record for in-process and completed engineering work which is retained and from which information can be retrieved. EOP 42-10.00 Appendix D, rev. 7 indicates that all DRFs require an assignment sheet, a table of contents, and any supporting information required by EOPs. Supporting information includes contractual or commercial documents which supply customer unique requirements (e.g., QA and design inputs). Contrary to the procedural requirements, many GENE DRFs did not include the ComEd purchase orders. This area was found deficient and identified as a Level III Finding CAR-G-97-120-04.

In addition, the scope of the Policy & Procedure NEDE-31744, Procedure NO. 10-27, issued 9/94 "Proposal & Sales Contract" references ISO-9001 and has also been utilized for safety related contracts. There is no reference to 10CFR50 Appendix B. Unresolved Item G-97-120-10 was issued to revise the subject procedure and incorporate a reference to 10CFR 50 Appendix B.

Design and Software

This audit reviewed the translation of design input documentation and the subsequent use of the design inputs, assumptions, methodology, references, summaries and conclusions into the GENE program for the development of engineering calculations. A rigorous technical review was performed on a population of 54 design calculations in 15 DRFs from the identified population of 42 DRFs containing safety related design calculations. Since there were only two electrical I&C DRFs the audit team excluded them from the population of the selected DRFs. The 54 calculations were completed by G.E. within the last three years and covered the Structural, Mechanical, and Nuclear Engineering disciplines. The calculations were a composite of design basis, plant operability, and supporting plant modifications and were based upon PRA significant systems.

There were 49 of 54 design documents selected that had a variety of administrative and editorial discrepancies including document legibility, page numbering, record identification which revealed a lack of formal control in the GENE design control process. Additionally, there were 25 of 54 design documents containing discrepancies involving the documentation of the independent design review. Additionally, there were also 45 of 54 design documents found to have design control deficiencies such as unjustified assumptions, references lacking, design input errors, and inadequate detailed analysis. No design documents were found to be free of any deficiencies. As a result of all the numerous deficiencies found in the design documents, the audit team assessed that the independent design verification process was ineffective.

The following is a summary of the technical specialists’ comments on the 54 calculations that were reviewed during this audit at GENE in San Jose, CA. See the referenced Corrective Action Records (CAR’s) for specific composite details concerning the deficiencies that were identified. The calculation summaries are provided as follows:

 

1. CORE SPRAY CRACK ANALYSIS FOR QUAD CITIES UNITS 1 & 2: DRF No. 137-0010-7; ISIS No. 1EXB5 (GENE-523-A80-0594, DATED 6/9/94

 

A. INTRODUCTION

This DRF performs evaluations of a crack indication on the "B" core spray line at the Quad Cities Unit 1. This indication was identified during an inspection in response to IE Bulletin 80-13. The crack indication is located outside the shroud, just inside the vessel, where the piping and the junction box meet in the heat affected zone of the weld. The crack is estimated to be approximately 120° of the outer circumference of the pipe. Using the top of the core spray pipe as reference of 0° azimuth, the crack was located from approximately 30° azimuth to 150° azimuth.

 

 

B. DRF CONTENT

This DRF (#137-0010-7) contains seven (7) calculations that perform various evaluations related to the subject crack as follows:

Determination of Loads: Tab B

Allowable Crack Size Tab C

Crack Growth / Leakage Tab D

Vibration Analysis Tab E

Fatigue Analysis Tab F

Thermal Mismatch Analysis Tab G

Displacement Calculation Tab N

Tab M includes the summary report based on the above seven calculations.

 

 

C. EVALUATION

C.1 GENERIC FINDINGS APPLICABLE TO ALL TABS

(CAR No. G-97-120-4)

Specific findings for applicable calculations are provided separately in the subsequent paragraphs here.

 

C.2 TAB B : DETERMINATION OF LOADS

In this calculation, a finite element model of the Quad Cities 1 & 2 core spray line was developed to calculate the stresses at the crack location due to dead weight and seismic loadings. ANSYS computer code was used to perform the analysis. This calculation is deficient in the following respects:

 

 

C.3 TAB C: ALLOWABLE CRACK SIZE

In this tab, the allowable through-wall crack size is calculated based on plastic hinge formation methodology given in a published reference. This methodology is acceptable, however, the reference is not fully documented and various variables used in the analysis are not defined in this calculation. More complete documentation, however, is provided later for the plastic hinge formation methodology in Tab M (CAR No. G-97-120-3).

 

C.4 TAB E: VIBRATION ANALYSIS

GE performed vibration analysis for Monticello core spray crack configuration considering vibration data from Kuosheng I reactor (NEDE 22146). For the Monticello analysis, justification for using the Kuosheng I data was based on the similarity of the geometry of the Monticello and Kuosheng I core spray lines and the vortex shedding frequencies of the two lines. Results found the Monticello core spray line unaffected by vibrational loading.

The present calculation, on page 2 of Tab E states that "Since Quad Cities appear at least as rigid as Monticello based on geometry and since the Monticello analysis very conservatively demonstrated that vibrational loading is not a factor, vibrational loading is not expected to affect the Quad Cities 1 & 2 core spray lines."

The above statement, in reality, is based on two levels of similarity - viz. Kuosheng I Vs Monticello and Monticello Vs Quad Cities. This makes the evaluation very approximate. A specific comparison based on the geometry, stiffness and supports of the Monticello Vs Quad Cities piping systems is not performed to justify the applicability of the Monticello vibration analysis to the Quad Cities vibration analysis.

It will result in a better evaluation if a modal analysis of the Quad Cities Piping is performed since a finite element model already exists for this system in Tab B. Based on the frequencies from the modal analysis and the vortex shedding frequency, a better evaluation for vibration can be performed for the Quad Cities crack configuration.

It may further be noted that no reference (such as report number, document number etc.) is provided in the calculation for the Monticello vibration analysis. (CAR No. G-97-120-3).

 

C.5 TAB F: FATIGUE ANALYSIS

Fatigue crack growth behavior of austenitic stainless steels is affected by a number of parameters such as environment, material variability, geometry, mean stress and R ratio (Kmin/Kmax). In Tab F, a Fatigue Crack Growth Analysis for the Quad Cities Units 1 & 2 core spray system is documented. This analysis is based on the effects of fatigue due to thermal shock loading on crack growth rate for Monticello core spray line crack. However, there is no documentation provided which shows that the material, environment, geometry, and R ratio parameters of the Quad Cities core spray crack are bounded by the Monticello core spray crack. Objective evidence should be provided by GE by comparing the various material and crack geometry/stress parameters to establish the applicability of the Monticello core spray crack fatigue analysis to Quad Cities core spray crack fatigue evaluation

It may further be noted that no reference (such as report number, document number etc.) is provided in the calculation for the Monticello fatigue crack growth analysis. (CAR No. G-97-120-3).

 

C.6 TAB G: THERMAL MISMATCH

In Tab G, on p. 1 it is stated that "Based on similar geometries of the core spray lines and similar locations of the cracks, the Monticello and Quad Cities Units 1 & 2 systems are considered comparable with regard to this analysis." No detailed basis for this consideration (such as crack parameters and orientation / location comparison between Monticello and Quad Cities cracks) are provided by GE for the thermal mismatch analysis.

It may further be noted that no reference (such as report number, document number etc.) is provided in the calculation for the Monticello thermal mismatch analysis. (CAR No. G-97-120-3).

C.7 TAB N: DISPLACEMENT CALCULATION

Tab N includes the displacement calculation for the core spray line crack. In this calculation, an evaluation is made for postulated pipe break of the Quad Cities core spray line. The ANSYS finite element model of Tab B is modified to include a pipe break at the crack location. Based on this modified model, displacements of the pipe ends are calculated. Based these displacements, a total area through which leakage occurs is calculated as 2.2 in2. Based on this area, the leakage flow is determined to be 656 gpm. This calculation is deficient due to the following:

On page 2, it is stated that "P - Pinf = 64 psid (Source: Luke Jen, Core Spray LSE)." This is an incomplete reference. GE should provide the complete reference for this information.

On page 2, it is stated that "Dr. Fred Moody has reviewed both the methodology and calculations of this analysis and has concurred with its results". However, there is no objective evidence (signature of Dr. Fred Moody) in this calculation. Note that the Q. A. form (included with summary report in Tab M), for Preparer/Reviewer/Approver’s signatures, does not contain Dr. Fred Moody’s name. There is no letter/fax signed by Dr. Moody indicating that he has actually reviewed and concurred with this analysis. Proper letter/fax reference from Dr. Moody should be included in this Tab N (CAR No. G-97-120-3).

Two pages following page 4 have no page numbers. These two pages appear to be scratch pages, informally marked during some discussion, and out of context (CAR No. G-97-120-1).

C.8 TAB M: "CORE SPRAY CRACK ANALYSIS FOR QUAD CITIES UNITS 1 & 2"

 

Tab M provides the summary report for the core spray crack analysis for Quad Cities Units 1 & 2.

In this report, an analysis is presented for the Quad Cities core spray piping, to determine the allowable through-wall circumferential crack for which failure by plastic collapse might occur. In particular the following equations are used to determine the limiting crack size for a circumferential crack not penetrating the compressive side of the pipe:

Pb = [2 s f / p ] [2 sin b - (d/t) sin a ]

b = [p - (a /t)d - (Pm/s f) p ] / 2

where:

Pb = piping bending stress

Pm = piping membrane stress

a = half flaw angle

d = flaw depth

t = thickness of the pipe

 

This methodology is acceptable per ASME Boiler & Pressure Vessel Code, Section XI, Appendix C, (p. 392, 1995 Edition).

This report (Tab M) does not provide references to applicable calculations which are used as bases for the conclusions in the report. For example, the reference to the calculation containing the ANSYS finite element analysis of the core spray line in the vicinity of the crack is not provided. In the summary report (Tab M), it is stated that to assess the potential of flow induced vibrations for the cracked core spray line, additional analyses were conducted assuming a 180° through-wall crack. However, the references for these analyses are not provided in the report. (CAR No. G-97-120-1).

 

2. FINAL REPORT OF THE IMPACT EVALUATION OF USING GE9 80-MIL FUEL CHANNELS FOR THE LASALLE UNITS 1 & 2 - DRF NO.: A12-00098; ISIS NO.: 1ESR5, REPORT NO.: GE-NE-523-A191-1294 (DATED 10/9/95)

 

A. INTRODUCTION

LaSalle Units 1 & 2 originally used 100-mil thick fuel channels for GE9 fuel. These analyses are performed by GE to evaluate impact of using 80-mil thick fuel channels for GE9 fuel in LaSalle Units 1 & 2.

The fuel channel thickness is important to the reactor pressure vessel (RPV) and internals, and the surrounding structure. The overall dynamic response of the entire reactor vessel, including shroud and internals is affected by this important change.

In this DRF, GE performed the horizontal seismic (OBE and SSE) analyses, the vertical fuel-lift analysis, and the life evaluation analysis for the 80-mil channel configuration. The horizontal mathematical models used for seismic analyses were developed using equivalent beam and spring elements to represent the stiffness of reactor pressure vessel (RPV) and major internal components. Masses of the RPV and internals were lumped at nodes connecting the beam and spring elements. The inertial effects that the RPV and internal components incur due to motion of the adjacent fluid was also simulated in the model. Multi-unit components such as fuel assemblies and the control rods were modeled as equivalent beam and spring elements. The seismic analysis was performed using GE program SAP4G07. Response spectra at specified locations were generated from the elevated time histories using GE program SPECA05C.

B. DRF CONTENT

This DRF contains two binders each 4" thick and it includes the following three analyses for the LaSalle 80-mil channels:

 

C. EVALUATION

A review of the analyses contained in this DRF (DRF A12-00098 - ISIS No. 1ESR5), resulted in the following findings:

C.1

Part of the input data for the analyses performed in DRF A12-00098 (ISIS No. 1ESR5), were transmitted to GE by a letter from Sargent & Lundy (Reference 7: Letter from S. Singh of S & L to P. Shah of GE dated July 26, 1994). No NDITs were used, as such the validity of the input - and consequently the output (final analysis results) - is questionable. Since the input primarily consisted of the detailed finite element model of the reactor vessel, including shroud and all internals, all analyses are affected by any potential error in the input data (CAR No. G-97-120-8).

C.2

Part of the input data for the analyses performed in DRF A12-00098 (ISIS No. 1ESR5), were transmitted to GE by a fax from ComEd -Dresden site- (Reference 6: Fax from T. Behringer of ComEd to P. Shah of GE dated 7/22/94). No NDITs were used, as such the validity of the input - and consequently the output (final analysis results) - is questionable. (CAR No. G-97-120-9).

C.3

The vertical fuel-lift analysis was performed by using an in-house computer code for non-linear analysis. This analysis is described on page 6 of the DRF, but no reference is provided for the GE non-linear in-house code used in the analysis (CAR No. G-97-120-3). GE engineer indicated that they used SEISM02 program for this analysis. Validation documentation for the SEISM02 program was not available for review. This is an Unresolved Item (CAR No. G-97-120-11).

C.4

Validation documents for SAP4G07, SPECA05C, and CHANL01V programs used in this DRF were also not available for reviews. This is an Unresolved Item (CAR No. G-97-120-11).

C.5:

In DRF A12-00098 (ISIS No. 1ESR5), changes made by hand on pages A-3, A-5, and page # 224 of computer listing in index 5 are not initialed by the preparer and the reviewer. (CAR No. G-97-120-1).

In DRF A12-00098 (ISIS No. 1ESR5), all the figures and tables have no record identification numbers. (CAR No. G-97-120-1).

In DRF A12-00098 (ISIS No. 1ESR5), all pages of computer output for SAPG07 program (about 6" thick output) have no record identification numbers (CAR No. G-97-120-1).

C.6

In this DRF (DRF A12-00098 - ISIS No. 1ESR5), no analysis is performed for the SRV and other hydrodynamic high frequency loadings for the new (80-mil) RPV finite element model. A detailed, documented comparison of the various structural frequencies should be performed in relation to the SRV spectral peaks for the new (80-mil) and old (100-mil) RPV finite element models to determine the potential impact of the combined hydrodynamic and seismic loadings. This is required to satisfy the original design basis loading criteria for the plant for the 80-mil configuration per the LaSalle UFSAR load combinations (CAR No. G-97-120-3).

C.7

In this DRF (DRF A12-00098 - ISIS No. 1ESR5), the channel lifetime evaluation is performed by GE using an in-house program CHANL01V. This program is used by GE to predict acceptability of the channel design. However, these predictions have unqualified uncertainties. The channel lifetime evaluation consists of determining the possibility of excessive channel to control rod friction, such that interference with normal control rod motion occurs. The interference depends on a number of parameters such as:

Some of the above parameters are accounted for by GE program in a simplified, statistical way. This is not acceptable, since the parameters listed above cannot be represented accurately by analytical models. As such the results by this program could vary considerably and may not represent actual operating situation. Moreover, in this approach, no account is made for the adhesive wear and localized asperity contact deformation at the interface which primarily control friction. GE should not use this analytical method for life prediction but instead develop, controlled tests based on field data and use more deterministic approach to predict life (CAR No. G-97-120-3).

 

3. "Structural Evaluation of Potential Top Guide & Core Plate Cracking at Dresden 2 & 3," DRF No. 137-0010-8, GE-NE-523-A081-0895, ISIS No. 1FQQX, Dated 12/1/95.
    1. Sheet 5, considered OBE is the governing loading case and used the scaling factor = (180/171) = 1.05 for the evaluation. However, per sheet 13, the ratio between SSE and OBE of the TOP Guide of Dresden Seismic Loading is (390/180) = 2.16. The safety factor ratio between OBE and SSE is only 2.0. Since 2.16 > 2.0, therefore, the SSE should be the governing loading case for Top Guide (not the OBE loading case). Re-evaluation of Top Guide is required. (CAR No. G-97-120-3)
    2. Sheets 4 & 9, "Critflaw" computer program is used. After discussion with Mr. H. Mehta (the Author of "Critflaw" computer program), this computer program is not verified and validated per GE procedure EOP42-10.00. Therefore, this computer program should be treated as a hand calculation, i.e., it is necessary to verify and validate every time it is used. In this DRF, the computer program source codes and the references of equations and allowables should be listed, verified and validated. (CAR No. G-97-120-7)
    3. Sheet 13, Note: Vertical Coefficients in paragraph 3.9.3.1.1.2 of the UFSAR are 0.08g and 0.16g, but it is assumed that ComEd and GE have agreed to the above values from reference 6 (reference 6, the vertical coefficient: Top Guide = 0.067). 0.08/0.067 = 1.19. Provide the justification of this assumption to address the 19% difference. The assumption needs to have some solid justification. (CAR No. G-97-120-3)
    4. Sheet 8, line 1, the crack growth rate is based on 304 Stainless Steel. The specific references are required to provide that: (1) The material is 304 stainless steel for the top guide and (2) The crack growth rate is based on the maximum temperature of how much degree F for the top guide. (CAR No. G-97-120-3)
    5. Sheet 13 is from Reference 3 which is only applied to Dresden Unit 2. Since this DRF applies to both units 2 and 3, the justification should be provided for using the information from Reference 3 to indicate that the Reference 3 also applies to unit 3. (CAR No. G-97-120-3)
    6. Sheet 6, BWR/6 loads, Horizontal OBE = 500.0 kips. It can not be verified that this value is correct or not. This value is from sheet 37, however, the definitions of HD, MD, HE and HF should be provided to prove that the correct value is being used. (CAR No. G-97-120-3)
    7. Sheet 53, the dimension of b = 4.464 inches is from Reference 5. Provide the justification to show that this value b = 4.464 inches can be applied to Dresden units 2 & 3. (CAR No. G-97-120-3)
    8. Sheet 60, s b, int = (MY/I) = (388.538) X (25.5-16.68) / 24.17. Since Y = 25.5 - 16.68 = 8.82 is less than 16.68. Provide the justification to prove that the stress calculated at Y = 8.82 is critical (as opposed to the stress at location of 16.68). (CAR No. G-97-120-3)
    9. Sheet 31, line 3, only the maximum axial stress occurs at point B is evaluated. Provide the justifications to prove that the stresses at all other directions and locations are not critical. (CAR No. G-97-120-3)
    10. Provide the justification of the thermal effect is negligible. (CAR No. G-97-120-3)
    11. Sheets 50 and 51 are part of a letter from ComEd providing the design inputs for this DRF. The ComEd NDIT procedure NEP-12-03 (the requirements for the preparation and control safety-related, regulatory related and non-safety-related nuclear design information transmittals) were not utilized. (CAR No. G-97-120-9)
    12. Sheet 1 (cover sheet), line 2, sheet 1 of ?? . The total number of sheets is unknown. The sheet number is labeled up to 66. It can not be determined that there are any other sheets missing, since the total number is unknown. (CAR No. G-97-120-1)
    13. Sheet 1, (cover sheet), item 1A, Application: Dresden 2 reactor assembly. It should be Dresden 2 & 3 reactor assembly. (CAR No. G-97-120-1)
    14. Sheet 1, Item 1F, Responsible Engineers are C. L. Chu/Ed Ng. Only Mr. C. L. Chu signed this sheet, but Mr. Ed Ng did not sign it. Mr. C. L. Chu indicated that Mr. Ed Ng resigned at that time and was not available to sign. (CAR No. G-97-120-1)
    15. Sheet 1, Item 4A should be check marked. (CAR No. G-97-120-1)
    16. No sheet number is labeled on the Reference sheet. Also, References 2 and 4 have no indication of revision number or date. (CAR No. G-97-120-1)
    17. Cover sheet is labeled as sheet 1. Sheet of the Objective section is also labeled as sheet 1. Why do two separate sheets have the same number? (CAR No. G-97-120-1)
    18. Sheet 1 (Item 1C), 4, 5, 8, 57, 58, 59 and 60 have been "Lined-out and Changed" without being initialed and dated. (CAR No. G-97-120-1)
    19. Several sheets have no DRF Numbers to indicate that these sheets belong to this DRF. (CAR No. G-97-120-1)
    20. Letter from C. L. Chu/D. B. Drendle (GE) to J. Williams (ComEd), Subject "Dresden Nuclear Power Plant Units 2 and 3, Structural Evaluation of Potential Top Guide and Core Plate Cracking", Dated 11-17-98. The date 11-17-98 is incorrect (i.e., it is a future date). (CAR No. G-97-120-1)

 

4. "Response to Commonwealth Edison Technical Audit Questions", DRF No. 137-0010-7, GE-NE-523-A69-0594, ISIS No. 1EXB8, Dated 6/20/94.
    1. Sheet 1 indicated that Mr. H. Metha is the independent verifier. However, sheets 17 through 24, Mr. H. Mehta signed as originator. (CAR No. G-97-120-2)
    2. Sheet 17, using the thickness = 3" to calculate the R/t ratio. However, sheets 13, 14 and 23 indicated that the thickness is 2". Based on the thickness of 2", the R/t ratio = (207.125-2)/(2 X 2) = 51.28. The results will be changed based on the R/t ratio being different. (CAR No. G-97-120-3)
    3. Sheet 15 should indicate that the computer output values were generated from the "Critflaw" computer program. After discussion with Mr. H. Mehta (the Author of "Critflaw" computer program), this computer program is not verified and validated per GE procedure EOP42-10.00. Therefore, this computer program should be treated as a hand calculation, i.e., it is necessary to verify and validate every time it is used. In this DRF, the computer program source codes and the references of equations and allowables should be listed, verified and validated. (CAR No. G-97-120-7)
    4. This DRF is for Dresden Unit 3 and Quad Cities Unit 1 (indicated on sheet 1). However, the comparison (sheets 13 through 24) is only from Dresden Unit 3. The justification should be provided to indicate that this comparison is applicable to Quad Cities Unit 1. (CAR No. G-97-120-3)
    5. Sheets 13 and 14 should provide a reference to indicate the source of these values. (CAR No. G-97-120-3)
    6. Sheet 16, the last line, a specific reference is needed for S.F. = 1.4. (CAR No. G-97-120-3)
    7. Sheet 17, ratio = 2.009/1.8567 and constant = 2.7 should have a detailed explanation regarding the meaning of these values. (CAR No. G-97-120-3)
    8. Sheet 23, line 13, a specific reference or explanation is needed for "2 X 0.75". (CAR No. G-97-120-3)
    9. Sheet 22, a specific reference or explanation is needed for this sheet. (CAR No. G-97-120-3)
    10. Sheets 18 and 19 are missing. Were these sheets part of the original DRF ? (CAR No. G-97-120-1)
    11. Several sheets have no DRF Numbers to indicate that these sheets belong to this DRF. (CAR No. G-97-120-1)
    12. Sheet 1, Item 1D, indicated "see sheet 5 of letter". Sheets 3 through 10 are editorial comments on Draft Letter GLS 94-11, Dated 6/8/94. Sheets 25 through 34 are the final letter GLS 94-11. Sheet 1, item 1D shall indicate "see sheets 25 through 34" instead of sheet 5, since sheet 5 is one of the pages of the draft letter. (CAR No. G-97-120-1)
    13. Sheet 1, Item 4A should be check marked. (CAR No. G-97-120-1)

 

5. "KVS a Profile for H5 Weld", DRF #137-0010-7, GE-NE-523-A69-0594, ISIS No. 1EXB8, Dated 6/20/94.
    1. Sheet 4, a specific reference or explanation is needed for Weld Residual Stress Profile. (CAR No. G-97-120-3)
    2. Sheet 5 needs a detailed explanation why the results are the same and which chart is being compared to? (CAR No. G-97-120-3)
    3. Several sheets have no DRF Numbers to indicate that these sheets belong to this DRF. (CAR No. G-97-120-1)
    4. Sheet 1, Item 4A should be check marked. (CAR No. G-97-120-1)

 

6. "Evaluation of the Indications Found at H5 Weld in Dresden Unit 3", DRF No. 137-0010-7, GE-NE-A69, ISIS No. 1EXB8, Dated 6/7/94.
    1. Sheet 4, Line 17, a specific reference is needed for Sm = 16900 psi which includes the material as being 304 stainless steel and the maximum temperature as 550 degree F. (CAR No. G-97-120-3)
    2. Sheet 4, 3" = the wall thickness (2") + fillet weld (1"). Specific reference is needed to explain that the strength of the weld is equal or stronger than the strength of the shroud material. (CAR No. G-97-120-3)
    3. Sheet 1 indicates that this DRF has 14 sheets. However, after carefully counting the sheet numbers, the total number of sheets is 20. (CAR No. G-97-120-1)
    4. Sheet 1, Item 4A should be check marked. (CAR No. G-97-120-1)
    5. Sheet 1, Item 1E, Outputs: Report GE-NE-523-A69-0594, Rev. 0. Report GE-NE-523-A69-0594, Rev. 0 is prepared and verified on 6/7/94. However, this DRF is prepared on 5/17/94 which is early than 6/7/94. The output was not approved yet to be used as an input into this DRF. (CAR No. G-97-120-3)

 

7. "LaSalle Unit 1 and Unit 2, Riser Pipe Flaw Evaluation Handbook", DRF No. B13-01869-009, ISIS No. 1G5WA, Dated 3/26/97.
    1. Sheet 9, line 6, states load combinations are consistent with LSCS UFSAR. After carefully reviewing LSCS UFSAR Table 3.9-16, Rev. 4, Dated April 1988, there are several loading cases that are missing in Emergency/Faulted combination: Load Cases 3 : (N + SRV + SSE), Load Case 5 : (N + SRVads + OBE + SBA/IBA), and Load Case 6 : (N + SRVads + SSE + SBA/IBA). These loading cases should have been evaluated. Also, LSCS UFSAR should be added as a Reference on Section 9 (References Section). (CAR No. G-97-120-3)
    2. Sheet 1, Item 1E indicated that this cover sheet is for Final Draft Report GE-NE-523-B13-01869-009 (Draft which is signed on 3/26/97). However, the final report is prepared and reviewed on May 1997. First, the Engineering Analysis Verification Cover Sheet for the final report is not documented. Second, after comparing the results of the draft report (3/26/97) and the final report (May 1997), it was found that they are different (see Section 7.1, Fatigue Evaluation and Section 7.2, Leakage Calculation). The justification should be provided to explain these two issues. (CAR No. G-97-120-3)
    3. Sheet 13, line 17, a specific reference is needed for Sm = 16900 psi which includes the material as being 304 stainless steel and the maximum temperature as 550 degree F. (CAR No. G-97-120-3)
    4. Sheet 14, the last 2 line, a specific reference is needed for D K of thermal expansion is less than 18 ksi (in) -5. (CAR No. G-97-120-3)
    5. Sheet 9, line 5, a specific reference is needed for the safety factors of 2.77 and 1.39. (CAR No. G-97-120-3)
    7. Sheet 14, line 3, a specific reference is needed for the calculated allowable axial flaw being 7.4" (CAR No. G-97-120-3)
    8. The results summary from computer analyses on sheets 15, 17 and 18 should have a detailed cross reference for these computer analyses. Also, a detailed list of all computer analyses should be provided in this DRF. (CAR No. G-97-120-1)
    9. Sheet 1, Item 4A should be check marked. (CAR No. G-97-120-1)

 

8. "LaSalle Unit 1 and Unit 2, Riser Pipe Flaw Evaluation Handbook, Verify FIV Stress", DRF No. B13-01869-009, GE-NE-523-B13-01869-009/TAB9, ISIS No. 1G5WA, Dated 3/26/97.
    1. Refer to Item 7A stated above for load combination. (CAR No. G-97-120-3)
    2. A detailed list of all computer analyses and a cross reference should be provided in this DRF. (CAR No. G-97-120-1)
    3. Sheet 1, Item 4A should be check marked. (CAR No. G-97-120-1)
    4. Several sheets have no DRF Numbers to indicate that these sheets belong to this DRF. (CAR No. G-97-120-1)

 

9. "Dresden 2 In-Vessel Visual Inspection Flaw Acceptance/Disposition Criteria", GE DRF No. 137-0010-7, ISIS No. 1F3ST
  1. On the Engineering Analysis Verification Cover Sheet, the Independent Verifier signed and dated for the preparer (Responsible Engineer) and sheet nos. 17, 18, 25, 28, 30, 32, 33, 40, 47a, 49, and 50 have changed several numbers, but no preparer’s and verifier’s signatures were evident. It appears that document control has been lost for this DRF and independent review process is questionable. Conversation with the verifier Mr. Chu, Principal Engineer, determined that he signed for the preparer, E. Ng, because he had resigned from GENE, however, the calculation was performed by E. Ng. (CAR No. G-97-120-2)
  2. Outputs from the program "CRITFLAW" were included in this DRF. Being considered as a hand calculation, the input, the parameters, the equations, and the output shall be included in the DRF and shall be reviewed. Being considered as an in-house program, it needs to be validated, verified, and document controlled properly. Without performing those actions, results are not reliable. (CAR No. G-97-120-3 & CAR No. G-97-120-7)
  3. On sheets 34, 35, 36, and 37, a plus b are less than p (3.1416). For this condition Case 1 should have been used, instead of Case 2. Also several actual Pb stresses tabulated on these sheets are higher than the allowables. (CAR No. G-97-120-3)
  4. Sheet 47b was initialed by MKK as a preparer, but he didn’t sign as a preparer on the verification cover sheet. MKK was not a preparer nor a verifier on the cover sheet. Preparer and verifier of this design document may not even be aware of sheet 47b. This may also constitute a change in the design document. (CAR No. G-97-120-1 and CAR G-97-120-2)
  5. Sheet no. 48, " Superseded by...", Signature, initial of person who wrote this statement, and date were missing. (CAR No. G-97-120-1)
  6. Design input data of OD (2") and ID (1.5"), OD (1.9") and ID (1.5"), and M were verbally taken from Dave Drendel. The references for these inputs were not documented. (CAR No. G-97-120-3)
  7. Design input data of 10 ksi , crack growth rate = 2X10-6 in/hr, and 1 fuel cycle » 17000 hrs of operation were verbally taken from H. Mehta. The references for these inputs were not documented. (CAR No. G-97-120-3)
  8. Sm value and input value of 0.01 on sheet no. 5c were not referenced. (CAR No. G-120-97-3)
  9. References on sheet 29, 46, and 48 were taken verbally from Maharaj Kaul. They were not documented and referenced properly. (CAR No. G-97-120-3)
  10. The calculation portion of DRF does not have final page number. or total number of sheets. It also does not have a control on the subpages. In addition, many sheets have no DRF numbers, nor GE titles, etc. to identify which DRF sheets they belong to. (CAR No. G-97-120-1)
  11. The Purchase Order was not included in the DRF. (CAR No. G-97-120-4)

10. "Evaluation and Screening criteria for the Dresden 3 Shroud Indication", DRF No. 137-0010-7, ISIS No. 1EJJ5, Index 2, sheet no. 2-1 to 2-34.

 
  1. The assumption on sheet no. 2 - 11 " The bounding crack growth estimated for the next fuel cycle was included in postulated flaw lengths used for evaluation" need to be verified. (CAR No. G-97-120-3)
  2. On Sheet no. 2-12, the justification for using the crack growth rate of 5X10 -5 in/hr is not documented. Why it is conservative? (CAR No. G-97-120-3)
  3. Letter from ComEd (H. Do) , 2/17/94, " Shroud Seismic Loads for Dresden and Quad Cities ", which was used as the design input for this DRF, was not sent via NDIT to GENE. (CAR No. G-97-120-9 for ComEd)
  4. Need to provide references for the following: (CAR No. G-97-120-3)
  1. Sheet 2-23, " Nevertheless a conservative fracture mechanics evaluation was performed using an equivalent Kjc ... The K jc for the overseas plant shroud was approximately 150 ksi Ö in...". however, the information source was not specified and was not referenced. This data should be verified and documented to show the comparability between these two plants. (CAR No. G-97-120-3)

11. Portion of this DRF, dated 4/12/95, " RCIC System Performance Calculations for Operating Plant", DRF No. E51-00178 Volume 1, Section 6, ISIS No. LS509
    1. The test report of Bingham Pump Co. is used as the design input. However, there is no pump model number, pump ID number, or system number shown on this test report. GENE needs to document the evidence and the reference to support that the correct test report is used for this RCIC pump. During the audit T. Simpson presented a document to support that this test report is for the subject RCIC pump. However this document needs to be signed, verified and documented in the DRF. (CAR No. G-97-120-3)
    2. GENE needs to document the justification that there will be no insignificant flow into the connected branch lines between the RCIC pump to the RCIC spray nozzle when the RCIC pump is operating. (CAR No. G-97-120-3)
    3. Justification for the additional losses, such as the relative power loss in bearing and stuffing box friction, and the hydraulic friction loss, is not documented for using the test report for a full sized pump tested at the reduced speed (3595 rpm) and for using equations, such as H2/N22 = H1/N12 and Q1/N1 = Q2/N2. Also need to document that the 4487 rpm is equal to or less than the full speed for pump operating condition. (CAR No. G-97-120-3)
    4. Need to add "NEDE-22034 (Based on A 251-BWR/5 (LaSalle), Figure 2-4" into the Reference Section. (CAR No. G-97-120-3)
    5. Need to provide the Rx assembly drawing no. and rev. no. for reference of the elevation from HPCS nozzle to RCIC head spray nozzle. (CAR No. G-97-120-3)
    6. Need to explain the reason for listing all information on sheets 2 and 4 within the assumption section. Confusion is caused as to whether those are actually design inputs or assumptions. It is believed that those are design inputs and subsequently need to provide references for such design inputs. (CAR No. G-97-120-3)
    7. On Engineering Analysis Verification Cover Sheet, Section 1D input section, item C, date is August 17, 19972, which is a typographical error. (CAR No. G-97-120-1)
    8. One attached sheet after sheet 12 of 12 is found. Therefore, the item 4A on the engineering analysis verification sheet should have been check marked. (CAR No. G-97-120-1)
    9. Purchase Order was not included in this DRF. (CAR No. G-97-120-4)
    10. Sheets 7, 8, 9 and 9A (Dated 8/15/96) were included in this portion of DRF after the preparer and the verifier had signed (Date of 4/12/95) for the original document. (CAR No. G-97-120-1)

12. Portion of this DRF, dated 8/15/96/, " RCIC System Performance Calculations for Operating Plant", DRF No. E51-00178 Volume 1, Section 6, ISIS No. LS509
    1. The test report of Bingham Pump Co. is used as the design input. However, there is no pump model number, pump ID number, or system number shown on this test report. GENE needs to document the evidence and the reference to support that the correct test report is used for this RCIC pump. During the audit T. Simpson presented a document to support that this test report is for the subject RCIC pump. However this document needs to be signed, verified and documented in the DRF. (CAR No. G-97-120-3)
    2. GENE needs to document the justification that there will be no insignificant flow into the connected branch lines between the RCIC pump to the RCIC spray nozzle when the RCIC pump is operating. (CAR No. G-97-120-3)
    3. Justification for the additional losses, such as the relative power loss in bearing and stuffing box friction, and the hydraulic friction loss, is not documented for using the test report for a full sized pump tested at the reduced speed (3595 rpm) and for using equations, such as H2/N22 = H1/N12 and Q1/N1 = Q2/N2. Also need to document that the 4487 rpm is equal to or less than the full speed for pump operating condition. (CAR No. G-97-120-3)
    4. Need to add "NEDE-22034 (Based on A 251-BWR/5 (LaSalle), Figure 2-4" into the Reference Section. (CAR No. G-97-120-3)
    5. Need to provide the Rx assembly drawing No. and Rev. no. for reference of the elevation from HPCS nozzle to RCIC head spray nozzle. (CAR No. G-97-120-3)
    6. Need to explain the reason for listing all information on sheets 2 and 4 within the assumption section. Confusion is caused as to whether those are actually design inputs or assumptions. It is believed that those are design inputs and subsequently need to provide references for such design inputs. (CAR No. G-97-120-3)
    7. On Engineering Analysis Verification Cover Sheet, Section 1D input section, item C, date is August 17, 19972, which is a typographical error. (CAR No. G-97-120-1)
    8. One attached sheet after sheet 12 of 12 is found. Therefore, the item 4A on the engineering analysis verification sheet should have been check marked. (CAR No. G-97-120-1)

13. DRF T23-00740, "Dresden DBA LOCA Containment Analysis 1 LPCI/ Containment Cooling Pump and 2 CCSW Pumps," Jan 97

15 calculations exist in the DRF

18 miscellaneous items exist in the DRF including letters, reports, digitizing of data, etc.

Reviewed (4) calculations and (1) report:

Section 1.0 Heat Exchanger K value (Calculation)

Section 2.1 SHEX Analysis Cases 2, 2A, 2B (Calculation)

Section 2.7 SHEX analysis with Containment Heat Sinks Cases 2A1, 2B1, 3A1, 3B1 (Calculation)

Section 2.9 SHEX Analysis (Calculation)

Section 4.4 GE-NE-T2300740-2 (Report)
    1. Section 1 of DRF contains calculation pages that do not have sequential or total page numbers or have no DRF# in the header identifying the page belonging to this DRF. (CAR No. G-97-120-1)
    2. All sections of the DRF, the Reviewer had checked the "no comments" box on each of the independent design verification sheets, but contrary to this, GENE engineer stated that there were comments prior to signature that had been resolved on an informal basis. (CAR No. G-97-120-2)
    3. Sections 2.7 of the DRF, Reactor building heat transfer was not included or addressed. ECCS volumetric flow rate was converted to a mass flow rate assuming constant density that was not identified or justified. Non-condensable containment model uses air not the actual post LOCA gases nitrogen and hydrogen that was not identified or justified. Sec 2.9 the Reference 1 teleconference was a design input which should have been transmitted as an acceptable design input with a prepared and approved source. All sections, Break area not identified as a design input in OPL-4a or in final report and did not include Recirculation Piping Replacement diameter for Dresden Unit 3 or the Bottom Head Drain/RWCU additional flow path break area. (CAR No. G-97-120-3)
    4. Section 1, ComEd P.O. not included. (CAR No. G-97-120-04)
    5. Section 2.9, the Reference 1 telecon was a design input but not sent to GE as an NDIT. (CAR No. G-97-120-3)
    6. Sections 1, There was a letter to J. Nash (GE) from W. Dingler (ComEd) dated 10/16/96 which transmitted input data for the containment analysis. Based on discussion with GE principal engineer, this data is called the OPL-4a since GE issues the list of parameters and ComEd fills in the blanks. This document did not include a line item for the DBA LOCA break area to be used. The computer code used for this analysis SHEX-04V requires the break area to be used as a design input. The GE principal engineer stated that the break area from the Mark I containment analysis was used. This was not identified in the OPL-4a. All containment analyses calculations used a LOCA break area which did not include the Recirculation Piping Replacement diameter for D3 or the BHD/RWCU additional flow path break area. The LOCA DBA break area is a fundamental design input for this type of calculation. (CAR No. G-97-120-3)
    7. Section 2.9 SHEX Analysis (Calculation) was typically the same as the other SHEX calculation sections. One difference was that the cases run for this calc were directed from a teleconference with ComEd. Reference 1 identified this teleconference. However, there were no explicit notes from the teleconference or follow-up letter from ComEd or NDIT from ComEd. Design inputs for these calculations were taken from this teleconference, which were not confirmed by a follow-up NDIT prepared and reviewed at ComEd. (CAR No. G-97-120-9 for ComEd)

 

 

14. DRF A00-00648-5, "SHEX-04V Engineering Computer Program," June 23, 1993

14 sections exist in the DRF to control this software

Reviewed all 14 sections
    1. DRF A00-00648-5, SHEX-04V does not have a complete Software Requirements Description as the guidelines show in EOP 40-3.00, section A1.4. This ECP must match the same experimental test data used for the original SHEX-01 (A00 648) as a validation requirement. Since the SRD did not identify this validation requirement, the Software Test Plan & Test Report has not been validated against the same experimental test data used for the original SHEX-01 (A00 648) despite significant code revisions and enhanced models and capability. It was verified and validated with certain plant specific cases but not to the original code requirements. The Independent Design Verification Packet also did not identify this weakness despite guidance given in EOP 40-3.00, section A1.8 directing comparisons of results with experimental data. The sample plant analyses comparisons used for validation and verification software testing had a good detailed discussion but did not have clearly identified nor quantitative acceptance criterion. Criteria should be applied at least to the key results of the code, e.g. temperature and pressure. SHEX-04V cases for validation against the original 4TCO and Monticello SRV data should be run with defined quantitative acceptance criterion for each of the key results. Software testing reviews should include a discussion of the revised code results with respect to a defined quantitative acceptance criterion. Also the sensitivity of the mixing fraction models that was presented in NEDE-30911, the SHEX04 User’s Manual section 4.1 must be verified as applicable to SHEX-04V since the figures presented evidence based on SHEX-01. (CAR No. G-97-120-7)

15. DRF A00-03049, "SAFER-04 Engineering Computer Program," May 26, 1988

15.1 DRF A00-03049-1, April 25, 1991

15.2 DRF A00-03049-2, June 18, 1993

15.3 DRF A00-03049-3, October 5, 1993

12 sections exist in the DRF A00-03049 to control this software

Reviewed 8 sections in DRF A00-03049

Reviewed 1 section of A00-03049-1

Reviewed 2 sections of A00-03049-2

Reviewed 2 sections of A00-03049-3

    1. DRF A00-03049, SAFER-04 does not have a complete Software Requirements Description (SRD) as the guidelines show in EOP 40-3.00, section A1.4. The experimental data, TRAC-G data and sample plant analyses data used for SAFER-04 validation and verification software testing do not have clearly identified nor quantitative acceptance criterion. Acceptance of each code revision was based on the judgement of a review committee. The sample plant analyses comparisons used for validation and verification software testing had a good detailed discussion but did not have clearly identified nor quantitative acceptance criterion. All of the constraints applied by the NRC in the SER for the entire SAFER/GESTR-LOCA methodology are not clearly stated, outlined and activities for software changes were not explicitly addressed with respect to the each NRC constraint. For example, this code revision included model enhancements to the Jet Pump entrainment, Two phase leakage flow and the minimum core pressure drop. The sections of the Licensing Topical Report where these models were described or defined, which the NRC reviewed and approved, were not explicitly identified as a possible constraint of NRC’s approval. Since the SRD did not identify quantitative acceptance criterion or explicit constraints of NRC approved models, the Software Test Plan & Test Report did not address these issues despite significant code revisions, enhanced models and capability. GE should clearly identify and quantify acceptance criterion and NRC approved constraints as part of its SAFER-04 SRD used for software testing validation and verification. GE should explicitly show how each revision of SAFER-04 complies with these requirements in the Software Test Plan & Test Report as guidelines show in EOP 40-3.00, section A1.6. (CAR No. G-97-120-7)

16. DRF B13-01760, "LaSalle SRV Removal"

At least 6 calculations exist in the DRF

Many miscellaneous items exist in the DRF including letters, reports, references, etc.

Reviewed (5) sections representing (5) calculations and the final reports:

Availability Section

ATWS Section

Section 2.4 ODYN analysis

Section 2.6 SRVOOS Impact on Thermal Fatigue

Section 3.0

GE-NE-B13-01760 Report
    1. DRF B13-01760; Availability Section has calculation pages which do not have total page numbers. This should be corrected for the section indicated and all sections of this DRF which apply. DRF B13-01760 PER # TS-97-003, GE self identified significant deficiencies with this DRF. These PER items must be corrected. Items not found in PER were: all final reports delivered to ComEd and their independent design verification were not included in DRF microfiche. All of ComEd comments on draft revisions were not included in DRF. These should be inserted into the DRF. (CAR No. G-97-120-1)

      2.  

    2. In the Availability section of DRF B13-01760, the Preparer did not sign and date answers to Reviewer. The preparer should sign and date the answers to the reviewer’s comments. This should be corrected for the section indicated and all sections of this DRF which apply. (CAR No. G-97-120-2)

      3.  

    3. DRF B13-01760, L2C7 was used as a design input or rather as an assumption but it was not treated as a design input. Although use of L2C7 seems appropriate, no written, verbal or fax authorization from ComEd was evident regarding the use of this input. This should be clearly identified that the L2C7 cycle specific inputs were used for each calculation and have to be verified as appropriate prior to final application for later cycles at the plant. This statement was included in the reports but could not be found in any of the calculations. (CAR No. G-97-120-3)
    4. DRF B13-01760; ComEd P.O. not included. (CAR No. G-97-120-4)

 

17. DRF 137-0010-7, Tab O, "Quad Cities Core Spray Crack"

At least 5 SAFER/GESTR-LOCA calculations exist in the DRF

Reviewed (1) sections representing (5) calculations and the final reports:

Letter with SAFER/GESTR-LOCA Evaluation Section

 
    1. UNRESOLVED ITEM: DRF 137-0010-7, Tab "O", the Quad Cities Core Spray Piping Crack LOCA evaluation source DRF was not available for review. A summary of the LOCA evaluation appeared in a letter "Analysis of Postulated Pipe Break in Quad Cities Unit 1 ‘B’ Core Spray Header, " to M. Richter (ComEd) from V. McCarthy (GE) and D. Shen (GE), dated July 13, 1994. The technical adequacy of the SAFER/GESTR-LOCA evaluation can not be determined until the design calculation documents are provided . (CAR No. G-97-120-12A)

18. MSLB TRACG ANALYSIS-DRF L12-00817 (ISIS No. 1FA4Z, DRF L12-00819)

The MSLB TRACG analysis is an evaluation the provided a justification for continue operation for Dresden Units 2/3 and Quad Cities Units 1 and 2. The acceptance criteria for this analysis is that during a worst case accident (i.e., main steam line break (MSLB)) the shroud and attached channel guides will not lift above the top of the fuel channels, even when the shroud separates. To perform this analysis several programs are used to provide input to TRACG. The TRACG simulated reactor pressure vessel pressure gradients are use as inputs to a spread sheet calculation and subsequently in a computer program, SHRD-LIFT2, to calculate the vertical lift distance. This dynamic lift distance can not exceed the vertical distance between the bottom of the channel guides and the top of the fuel channels.

The Cycle 14 Quad Cities Unit 1 station reactor vessel and core geometries were used and were assumed to bound Dresden Units 2 and 3, and Quad Cities Unit 2. PANACEA, ISCOR, and ODYN-SS results were used as input to a computer program, ATRAC, which produced an input deck for TRACG. Values not provided by the three programs were provided by the analyst. Once the TRACG input deck was prepared, TRACG was used to simulate the MSLB. The output from TRACG was used in a spread sheet and in SHRD-LIFT2 to predict the vertical lift of the shroud.

The TRACG Qualification documented in NEDE-32177P, Rev. 1, June 1993 was reviewed. In Section 3.1.5, the PSTF level swell tests were simulated using TRACG. This simulation represented a steam line break. Water level and break flow appeared to match the test data. It was concluded from this review that TRACG was validated for the MSLB analysis.

 

In reviewing the DRF L12-00817, the following deficiencies were found relating to

CAR G-97-120-01:

"MSLB TRACG analysis": Many of the pages of the DRF did not have page numbers

or the DRF identified. This should be corrected for all sections of this DRF which apply.

The output files that were used for ATRAC could not be determined from the

documentation.

"ISCOR calculation of Quad Cities Cycle 14": This calculation was represented by

computer input and computer output. The output used for input to ATRAC was difficult to follow from the lack of organized documentation.

"PANACEA Calculation of Quad Cities": This calculation was represented by

computer input and computer output. The output used for input to ATRAC was difficult to follow from the lack of organized documentation.

"ODYN-SS Calculation of Quad Cities": This calculation was represented by some

minor calculations and a computer input and computer output. The output was used for input to ATRAC was difficult to follow from the lack of organized documentation.

ATRAC Calculation of Quad Cities, this calculation used input from ISCOR, PANACEA,

OPL-3, and ODYN-SS to develop input to TRACG. ATRAC identified values that were needed to complete the TRACG input. These values were developed as part of the DRF. The output used for input to ATRAC was difficult to follow from the lack of organized documentation.

(See CAR G-97-120-01)

 

In reviewing the MSLB TRACG analysis, DRF L12-00817, the following deficiencies were found relating to CAR G-97-120-3:

An OPL-3 from Quad Cities Unit 1 was used to bound the Quad Cities Unit 2 and Dresden Units 2 &3. The basis for Quad Cities Unit 1 OPL-3 values bounding the Quad Cities Unit 2 and Dresden Units 2 & 3 was that Quad Cities Unit 1 has been analyzed for 108 % core flow and Dresden has not. Therefore the Quad Cities Unit 1 conditions are expected to bound conditions of Dresden. However, If Dresden performed a new design basis calculation to increase core flow to 108%, there does not appear to be a GE process or control to trigger a reassessment of the MSLB TRACG Analysis. This is a Lack of Control of Design Input. This calculation was performed for Quad Cities 1 & 2 and Dresden 2 & 3. This DRF did not represent LaSalle.

Data was taken from a data base identified as LaSalle FDS.CYCLE.CEO and was used as input to the MSLB analysis for Quad Cities and Dresden. Apparently, the FDS.CYCLE.CEO is a GE controlled data base. However if data in FDS.CYCLE.CEO, that was used in the MSLB analysis, is changed, there is no mechanism in place to ensure that the potential impact on the MSLB DRF is evaluated. This is a Lack of Control of Design Input.

ISCOR calculation for Quad Cities Cycle 14, this calculation was represented by computer input and computer output. The output used for input to ATRAC was not clearly organized and was difficult to follow. References were not given which made the inputs not traceable.

PANACEA calculation for Quad Cities Cycle 14, this calculation was represented by computer input and computer output. The output used for input to ATRAC was not clearly organized and was difficult to follow. References were not given which made the inputs not traceable. Cycle 13 input was used instead of Cycle 14. No comparison or justification for use of Cycle 13 data for applicability to a cycle 14 analysis. The validity of this design input was not demonstrated.

ODYN-SS calculation for Quad Cities, this calculation was represented by some minor calculations, computer input and computer output. The output used for input to ATRAC was not clearly organized and was difficult to follow. References were not given which made the inputs not traceable.

ATRAC calculation for Quad Cities, this calculation used input from ISCOR, PANACEA, OPL-3, and ODYN-SS to develop input to TRACG. ATRAC identified values that were needed to complete the TRACG input. These values were developed as part of the DRF. Some of the values did not have adequate references, e.g., separator pitch. Traceable references were not given. The output used for input to ATRAC was not clearly organized and was difficult to follow. References were not given which made the inputs not traceable.

TRACG Calculation of Quad Cities, the decay power used to perform the TRACG calculation was not referenced. References were not given which made the inputs not traceable.

(See CAR G-97-120-03)

 

In reviewing the MSLB TRACG analysis, DRF L12-00817, the following additional deficiencies were found:

The lift calculation is a key aspect of the calculation. The computer program, SHRD-LIFT2, was used to determine the lift of the shroud. The results of this calculation was provided in the final report to ComEd. There is no documentation of the computer program. There is no documented verification or validation of the computer program. (See CAR G-97-120-07)

A Sargent and Lundy engineer provided design information and input to General Electric. A GE engineer indicated that ComEd told him, verbally, that the Sargent and Lundy engineer represents ComEd. In the input section, design input was sent from Sargent and Lundy directly to General Electric without design review by ComEd. (See CAR G-97-120-08)

 

 

19. WATER LEVEL INSTRUMENTATION SUPPORT-DRF B21-0537

This DRF contained many calculations which were used to determine the acceptability of the back fill of cold water from the control rod fill pumps into the hot condensing pots which are used to determine water level in the reactor vessel. The calculations consisted of three parts. The thermal/hydraulic, stress and set point bias calculations. The acceptance criteria is to satisfy the ASME stress allowable, due to the flow of the cold water into the condensing pot and down the steam leg into the reactor vessel. These analyses were performed for Dresden, Quad Cities and LaSalle.

In reviewing the DRF B21-0537, the following deficiencies were found relating to

CAR G-97-120-01:

Water Level Instrumentation Support- "Calculation RVWLLS Condensing Chamber,":

Could not read calculation and drawings.

"Mixed Mean Model Spread Sheet Usage,": Calculation document illegible.

"Calculation of Puddle Depth in the CC at LaSalle,": Calculation document illegible.

"Calculation Heat Transfer Coefficient Estimate,": Calculation document illegible.

"Calculation of Flow Area Used in the 1D Stratification,": Calculation document illegible.

"Calculation of Condensing Chamber Data Flow Split Calculation,": Calculation

document illegible.

"Data Used in EXCEL Spread Sheet (Mixed Mean Temperature),": Calculation

document illegible.

"Steam Leg Depth Calculation,": Calculation document illegible.

(See CAR G-97-120-01)

 

In reviewing the DRF B21-0537, the following deficiencies were found relating to

CAR G-97-120-02:

Water of Cond. Pot section, the sign off sheet was signed by preparer and reviewer,

but not approved. The reviewer had comments, but resolution of the comments were not documented and comments were not resolved. There was an inadequate completion of required design review documentation.

RVWLLS Cond. Chamber section, the sign off sheet was missing. A sheet signed by

the preparer, reviewer, and approver was either never completed or destroyed during microfiching.

This is inadequate design review documentation.

Mixed Mean Model Spreadsheet section, the sign off sheet was missing. A preparer

was identified on the calculation sheets but not a reviewer. A sheet signed by the preparer, reviewer, and approver was either never completed or destroyed during microfiching. This is inadequate design review documentation.

LS Puddle Depth in the CC at LaSalle section, the sign off sheet was missing. A

preparer was identified on the calculation sheets but not a reviewer. A sheet signed by the preparer,

reviewer, and approver was either never completed or destroyed during microfiching. This is

inadequate design review documentation.

H/T Coef. Estimate section, the sign off sheet was missing. A preparer was identified

on the calculation sheets and a reviewer. A sheet signed by the preparer, reviewer, and approver was either never completed or destroyed during microfiching. This is inadequate design review

documentation.

Flow Area, 1D Stratification section, the sign off sheet was missing. A preparer was

identified on the calculation sheets and a reviewer. A sheet signed by the preparer, reviewer, and

approver was either never completed or destroyed during microfiching. This is inadequate design

review documentation.

Cond. Chamber Flow Split section, the sign off sheet was missing. A preparer was identified on the calculation sheets and a reviewer. A sheet signed by the preparer, reviewer, and approver was either never completed or destroyed during microfiching. This is inadequate design review documentation.

Data used in EXCEL Spreadsheet section, the sign off sheet was missing. A preparer

was identified on the calculation sheets and a reviewer. A sheet signed by the preparer, reviewer, and approver was either never completed or destroyed during microfiching. This is inadequate design review documentation.

Steam Leg Depth Calc section, the sign off sheet was missing. A preparer was

identified on the calculation sheets and a reviewer. A sheet signed by the preparer, reviewer, and

approver was either never completed or destroyed during microfiching. This is inadequate design

review documentation.

H/T Coef. DR & QC section, the sign off sheet was missing. A preparer was identified

on the calculation sheets but not a reviewer. A sheet signed by the preparer, reviewer, and approver

was either never completed or destroyed during microfiching. This is inadequate design review

documentation.

Rx Water & Instr. Nozzle Data section, the sign off sheet was missing. A preparer was

identified on the calculation sheets and a reviewer. A sheet signed by the preparer, reviewer, and

approver was either never completed or destroyed during microfiching. This is inadequate design

review documentation.

Length of 2" pipe for QC section, the sign off sheet was missing. A preparer was

identified on the calculation sheets but not a reviewer. A sheet signed by the preparer, reviewer, and approver was either never completed or destroyed during microfiching. This is inadequate design review documentation.

2nd Data used in EXCEL Spreadsheet section, the sign off sheet was missing. A

preparer was identified on the calculation sheets and a reviewer. A sheet signed by the preparer,

reviewer, and approver was either never completed or destroyed during microfiching. This is

inadequate design review documentation.

(See CAR G-97-120-02)


In reviewing the DRF B21-0537, the following deficiencies were found relating to

CAR G-97-120-03:

Dresden Backfill Section, GENE-637-031-1093, dated October 1993, the calculations for the cold liquid flow into the condensing pot. design inputs of 15 lb/hr and 19 lb/hr (found on page 3 of the report) did not have any reference which made the inputs not traceable. This brings the validity of these design inputs into question.

In Report "LaSalle Unit 2 Reactor Vessel Water Level Instrumentation System Backfill Report", GENE # 637-027-0993, many of the design inputs have no references and therefore the basis can not be established. An example of this is on page 10 & 11 of the report. Other examples were found on pages 28, 29, 30, 31, 32, 33, 34, 35 and 36. The lack of references make the inputs not traceable. This brings the validity of these design inputs into question.

"Reactor Water Level Backfill", an Engineering Services Verification Cover Sheet (Ref. EOP 42-6.00 and EOP 25-6.00), related to the " Revised Heat Transfer Coefficients" was prepared on 11/8/93 by Joe Darr and approved by Hank Phefferlen on 11/21/95, but the report included and the DRF were approved 9/9/93. It appears that design analyses were performed after the DRF was approved. It was not clear from the DRF if the revised heat transfer calculation was used as a design input for a 1993 report or for a 1995 report. The heat transfer coefficient design input was changed without proper controls or references.

Report # GENE-637-031-1093, the RPV level instrumentation bias should be evaluated against the setpoint methodology program to ensure that the set point basis was addressed. No evidence or references could be found that this evaluation was performed. The lack of references make the inputs not traceable. This brings the validity of these design inputs into question.

Water of Cond. Pot calculation, the lack of a response to the reviewers comments on the design verification brings the validity of these design inputs into question.

RVWLLS Cond. Chamber calculation, the lack of legibility and design verification brings the validity of these design inputs into question.

Mixed Mean Model Spreadsheet calculation, the lack of legibility and design verification brings the validity of these design inputs into question.

LS Puddle Depth in the CC at LaSalle calculation, the lack of legibility and design verification brings the validity of these design inputs into question.

H/T Coef. Estimate calculation, the lack of legibility and design verification brings the validity of these design inputs into question.

Flow Area, 1D Stratification calculation, the lack of legibility and design verification brings the validity of these design inputs into question.

Cond. Chamber Flow Split calculation, the lack of legibility and design verification brings the validity of these design inputs into question.

Data used in EXCEL Spreadsheet calculation, the spreadsheet itself was not provided in the DRF. The lack of legibility and design verification brings the validity of these design inputs into question.

Steam Leg Depth calculation, the lack of legibility and design verification brings the validity of these design inputs into question.

H/T Coefficient "h" for DR & QC calculation, the lack of references and design verification brings the validity of these design inputs into question.

Rx Water & Instr. Nozzle Data calculation, the lack of design verification brings the validity of these design inputs into question.

Length of 2" pipe for QC calculation, the lack of design verification brings the validity of these design inputs into question.

2nd Data used in EXCEL Spreadsheet (Mixed Mean Temperature) calculation, the lack of legibility and design verification brings the validity of these design inputs into question

(See CAR G-97-120-03)

 

In reviewing the DRF B21-0537, the following deficiencies were found relating to

CAR G-97-120-05:

ComEd Engineering personnel prepared and performed independent verification of calculations under GENE QA Program without being indoctrinated and trained. Furthermore, GENE violated its program by not using its own employees. The following calculations were affected:

Water of Cond. Pot section.

RVWLLS Cond. Chamber section.

Mixed Mean Model Spreadsheet section.

LS Puddle Depth in the CC at LaSalle section.

H/T Coef. Estimate section.

Flow Area, 1D Stratification section.

Cond. Chamber Flow Split section

Data used in EXCEL Spreadsheet section.

Steam Leg Depth Calc section.

H/T Coef. DR & QC section.

Rx Water & Instr. Nozzle Data section.

Length of 2" pipe for QC section.

2nd Data used in EXCEL Spreadsheet section.

(See CAR G-97-120-05)


In reviewing the DRF B21-0537, the following item was noted as an Unresolved Item:

This DRF was approved on 9/8/93. The sign off sheet for this design document changing the heat transfer coefficient was approved on 11/21/95. Why wasn't the design document revised for this change, which constitutes a need for revision? Were the preparer and design verifier of this design document aware of this change to reevaluate the impact? (See CAR G-97-120-12 Item A)

 

Corrective Action:

GENE Engineering Operating Procedure No. EOP 75-3.00 revision 5, dated 7/30/97 addresses self assessments, corrective actions and audits. In the corrective action area, the Problem Evaluation Request (PER’s) form is utilized as a process of identifying a condition which could be adverse to quality. The PER process is identified to assure that conditions adverse to quality, such as potential causes of nonconforming product, failures, malfunctions, deficiencies, deviations, and other quality conditions. There are no restrictions as to who can identify and generate a PER.

PER’s are reviewed for validity and to determine the need for a root cause analysis, committed corrective action, and committed preventive action. Each of the PER’s receives a quality review. Adverse conditions are required to be communicated to the client. A reportability review and notification is intended to be performed per P&P 70-42. A review of four recently issued PER’s during the course of this audit revealed that PER’s are generated in accordance with approved procedures and that they are required to receive the appropriate review by the Engineering Manager, Quality Assurance and the assigned individual. Each of the PER’s reviewed were not dispositioned as yet since the PER process was started on July 31, 1997.

PER No. TS-97-004 was generated during this audit by GENE’s Regulatory Services Project Manager to document the deficiency identified by this audit regarding ComEd personnel performing design analysis on the CECo water level modification analysis project under the GE QA Program (CAR No. G-97-120-05). This was the only PER issued by GENE during this audit. The GENE QA Manager indicated that once GENE receives the audit findings from ComEd they will be evaluated for PER status

Internal Audits:

GENE was found to have an approved internal 1997 scheduled audit plan for Customer Services Asia & Europe, GENE Sourcing and Support, Information Management Systems, Nuclear plant projects, and Nuclear Services Department as administered by Nuclear Services Quality (NSQ). NSQ gets input from a particular business unit, selects specific jobs, develops a plan, schedule and distributes the information to the responsible departments. Checklists are then derived and generated for the specific activity being audited.

The information provided by the GENE QA Organization revealed that internal audits did not perform a technical review of design activities. The audit team after carefully reviewing several internal GENE audits, determined that no evidence exists which demonstrates that QA reviews the programmatic aspects of the design calculation portion of the DRF. After the exit meeting G.E. faxed some additional samples of audits to demonstrate a programmatic review of the design calculation portion of the DRF. However, these examples were found lacking to support a review of the design portion of the DRF. Therefore, the audit team assessed that the GENE's internal audits are ineffective in independently overviewing the design analysis area.

This area was found deficient and identified as a Level II Finding CAR-G-97-120-06

Indoctrination/Training:

Policies & Procedures NEDE-31746 Procedure No. 70-30 issued 8/94 has established the minimum personnel proficiency requirements to be implemented by each manager of employees performing activities affecting the quality of GENE products. Qualification for technical positions are documented to include minimum education, experience and/or special technical requirements. Employees responsible for adherence to Code or regulatory requirements shall maintain their knowledge of current provisions of such Codes, rules, and regulations.

Each person, prior to assignment of work activity affecting quality of products, shall be indoctrinated or instructed in the applicable quality system procedures. Indoctrination and training shall be attained and maintained. Contrary to the above procedural requirements, several ComEd Engineers were found to have performed and reviewed design analysis calculations under the GENE QA Program. In addition to not being employed by G.E. they were not indoctrinated and trained to GENE procedures.

This area was found deficient and identified as a Level II Finding CAR-G-97-120-05

EOP 40-9.00 rev. 10 provides responsibilities and procedures requirements to meet the ASME Boiler & Pressure Vessel Code (BPV) requirements for Certification of Design Specification, Design Drawings and Reports, Load Capacity Data Sheets, Overpressure Protection Reports, and Construction Specifications. Appendix A of the subject procedure indicates that Registered Professional Engineers must meet the requirements of Code and ASME/ANSI N626.3-1993 in order to certify the above stated documents. Interview with the GENE QA and Engineering Managers revealed that no such certification activities has been performed in the past three years. The qualification of inspection personnel, auditors, NDE personnel was not part of the scope of this audit.

There were no ASME Class 1 pressure boundary activities performed by GENE for ComEd for the past three years.

Attachments

Attachment 1 - Entrance/Exit Meeting Attendance

Attachment 2 - Audit Team

Attachment 2 - Personnel Contacted During the Audit

Attachment 1

Entrance / Exit Meeting Attendance

 

Entrance Meeting (8/18/97)

Name Title Organization

Oscar Shirani Audit Team Leader ComEd

Yakub A. Patel Technical Specialist ComEd

Albert Lie-Mien Sheng Technical Specialist EMS

YuHua Chen Technical Specialist EMS

George B. Strambach Regulatory Services, PM GENE

Dave Grim Regulatory Engr./Tech. Account Interface GENE

Robert J. Nicholls Manager-Nuclear Services Quality GENE

Shyam S. Dua Plant Analysis Services Manager GENE

Bradley J. Erbes Engineering Resource Manager GENE

Joe Quirk Industry Programs GENE

Noel Shirley Principal Engineer GENE

 

Exit Meeting (8/22/97)

Name Title Organization

Oscar Shirani Audit Team Leader ComEd

Yakub A. Patel Technical Specialist ComEd

YuHua Chen Technical Specialist EMS

Joe Miller Techncial Specialist EDA

John Freeman Lead BWR Safety Analysis ComEd

Lie-Mien (Albert) Sheng Technical Specialist EMS

George B. Strambach Regulatory Services, PM GENE

Shyam S. Dua Plant Analysis Sevices Manager GENE

Robert J. Nicholls Manager-Nuclear Services Quality GENE

Saul Mintz Senior Engineer GENE

Carl Young Principal Engineer GENE

Har Mehta Principal Engineer GENE

Cherk Chu Principal Engineer GENE

Hwang Choe Principal Engineer GENE

Gerald Hayes Manager, Engineering GENE

D. R. Helwig General Manager Nuclear Services GENE

P. T. Tran Engineer Leader GENE

N. E. Barclay Manager Audits GENE

Noel Shirley Principal Engineer GENE

Gary Plotycia Nuclear Account Executive ComEd

Attachment 2

Audit Team

 

Audit Team

Name Title Organization

O. B. Shirani Audit Team Leader ComEd

Y. A. Patel Technical specialist ComEd

N. Chen Technical specialist EMS

A. Sheng Technical Specialist EMS

J. Freeman Technical Specialist ComEd

J. Miller Technical Specialist EDA

 

Personnel Contacted During Audit

 

Name Title Organization

H. S. Mehta Principal Engineer GENE

P. K. Shah Engineer GENE

C. Chu Engineer GENE

M. Romero Engineer GENE
George B. Strambach Regulatory Services, PM GENE

Shyam S. Dua Plant Analysis Sevices Manager GENE

Robert J. Nicholls Manager-Nuclear Services Quality GENE

Har Mehta Principal Engineer GENE

Cherk Chu Principal Engineer GENE

Hwang Choe Principal Engineer GENE

N. E. Barclay Manager Audits GENE

Noel Shirley Principal Engineer GENE