Text

Insp Repts 50-498/88-07 & 50-499/88-07 on 880125-29 & 880208-12.No Violations Noted.Major Areas Inspected: Structural Steel Welding,Pipe Supports & Restraints,Nuclear Welding,Preoperational Welding & Onsite Design Changes
	ML20148G687
Person / Time
Site:	South Texas
Issue date:	03/16/1988
From:	Gilbert L, Hunnicutt D, Stewart R; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:	;
Shared Package
ML20148G649	List: ML20148G644; ML20148G665; ML20148G687;
References
	50-498-88-07, 50-498-88-7, 50-499-88-07, 50-499-88-7, NUDOCS 8803290203
	Download: ML20148G687 (16)
	v • d • e

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APPENDIX B

U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

NRC Inspection Report: 50-498/88-07 Operating License: NPF-71

50-499/88-07 Construction Permit: CPPR-129

Dockets: 50-498

50-499

Licensee: Houston Lighting & Power Company (HL&P)

P.O. Box 1700

Houston, Texas 77001

Facility Name: South Texas Project (STP), Units 1 and 2

Inspection AT: STP, Matagorda County, Texas

Inspection Conducted: January 25-29andFebrud 8-12, 1988

Inspectors: a 2 d

R. C. Stewarf,-Reactor Inspect 6f, Materials Date '

and Quality Programs Section, Division of

Reactor Safe y

L. D. Gilbert, Rsactor Inspector, Materials

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and Quality Programs Section, Division of

Reactor Safety

bi YtwuenW& 3 Af$

D. M. Hunnicutt, Project Engineer, Reactor Date '

Project Section D, Division of Reactor Projects

Other Accompanying Personnel:

R. V. Azua, Test Programs Section, Division of Reactor Safety

R. C. Haag, Materials & Quality Programs Section, Division of

Reactor Safety

Approved: b%M

I. Barnes, Chief, Materials and Quality Programs

3b7 $$

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Date

Section, Division Reactor Safety

UdO3290203

DR ADOCKG9031e 05000493

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Inspection Sumary

inspection Conducted January 25-29 and February 8-12, 1988 (Report 50-498/88-07)

Areas Inspected: Routine, unannounced inspection of licensee action on

previously identified inspection findings, nuclear welding, and manual reactor trip circuits.

Results: Within the three areas inspected, two violations were identified

(failure to test welding material for different postweld heat treatment

applications, paragraph 4; and failure to provide adequate control of quality

assurance records, paragraph 2).

Inspection Conducted January 25-29 and February 8-12, 1988 (Report 50-499/88-07)

Areas Inspected: Routine, unannounced inspection of structural steel welding,

pipe supports and restraints, nuclear welding, preoperational testing, onsite

design changes, safety-related components, and manual reactor trip circuits.

Results: Within the seven areas inspected, a violation was identified (failure

to test welding material for different postweld heat treatment applications,

paragraph 4).

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DETAILS

1. Persons Contacted

HL&P

G. Vaughn, Vice President, Nuclear Operations

+*J. T. Westermeier, Project Manager

+*J. E. Geiger, General Manager, Nuclear Assurance

+*J. S. Phelps, Supervisor, Project Compliance

+*T. J. Jordan, Project Quality Assurance (0A) Manager

+J. N. Bailey, Engineerir.g & Licensing Manager, Unit 2

+*D. C. King, Construction Manager

M. A. McBurnett, Operations Support & Licensing Manager

Bechtel Energy Corporation

+*R. W. Miller, Project QA Manager

+*R. D. Bryan, Field Construction Manager

+K. P. McNeal, Project QA Engineer

+E. B. Luder, Lead QA Engineer

Ebasco Services Inc.

+*D. D. White, Construction Manager

+*A. M. Cutrona, Quality Program Site Manager

F. G. Miller, Welding Superintendent

+R. E. Abel, Quality Control Site Supervisor .

NRC

+*D. M. Hunnicutt, Project Engineer

+*R. C. Stewart, Reactor Inspector

+*L. D. Gilbert, Reactor Inspector

+*R. C. Haag, Reactor Inspector

+R. V. Azua, Reactor Inspector

C, E. Johnson, Senior Resident Inspector (Construction)

D. R. Carpenter, Senior Resident Inspector (Operations)

D. L. Garrison, Resident Inspector

A. B. Beach, Deputy Director, Division of Reactor Projects

J. P. Jaudon, Deputy Director, Division of Reactor Safety

The NRC inspectors also interviewed other licensee and contractor

employees during the inspection.

Denotes those attending exit interview on January 29, 1988.

+ Denotes those attending exit interview on February 12, 1988

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-2. Licensee Action on Previously Identified Inspection Findings (92702)

(Closed) Deviation (498/8726-06): Inadequate Control of_ Records. This

deviation identified the failure to properly controi quality

assurance (QA) records while in temporary working files and the failure to

comply with transmittal requirements when transferring QA records. This

deviation, as reported in NRC Inspection Report 50-498/87-26, dated

June 25, 1987, was discovered prior to issuance of the operating license

and the technical specifications (TSs).

In the corrective action portion of the licensee's response letter, dated

July 27, 1987, to the above noted deviation, the licensee committed to the

following action:

Each division that maintains QA records will assign a designated

record custodian who will be responsible for filing records,

maintaining accountability, controlling access, indexing stored

record, and transmitting those records to the record retention area.

Promulgation of the procedural requirement to use transmittal form.

In the response letter, the licensee also reported that some divisions had

determined that QA records will no longer be maintained cutside of the

Operations Document Control Center (0DCC) storage facility; therefore, QA

records will be transmitted to 0DCC in an expeditious manner. The

licensee stated "STP is in full compliance at this time" in the response

letter.

During this inspection, the NRC inspector reviewed approximately 30 recent

QA record transmittals to the ODCC. All these transm1ttals were

accompanied by the required transmittal forms. The ODCC supervisor stated

that ODCC personnel will not accept QA records without a transmittal form.

While inspecting reactor operations (RO) division for compliance with the

corrective action, the NRC inspector learned that no QA records are being

stored in division QA record files. Present policy requires that all QA

records generated by R0 division be transmitted to 0DCC upon completion

with copies of selected records being retained in the division file.

The chemistry department maintains QA records in their division records

file for 60 days. A records custodian has been assigned with

responsibility for maintaining control of QA records while in the division

files. Access to the files is limited to the custodian and selected

supervisors with any additional access being controlled by a checkout card

system. The NRC inspector verified that recent transmittals of chemistry

department QA records to ODCC were completed with transmittal forms.

During this inspection, the NRC inspector noted that the engineering

department also maintains QA records. Selected surveillance tests require

a trending review by system engineering upon completion of the test.

Those completed surveillance tests, that are awaiting engineering review,

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makeup the QA records in the divisional file reviewed by the NRC

' inspector. A record custodian had not been assigned for this divisional

QA records file. Access to the records is not limited. Also, no system

exists for maintaining accountability, indexing, or ensuring that the

maximum retention time is not exceeded for records being maintained in the

division file.

10 CFR Part 50, Appendix B, Criterion XVII, "Quality Assurance Records,"

states, in part, ". . . . Consistent with applicable regulatory

requirements, the applicant shall establish requirements concerning record

retention, such as duration, location and assigned responsibility."

Section 6.10, "Record Retention," of the TSs requires "Records of

surveillance activities, inspections, and calibrations required by these

TSs" be retained for at least 5 years. Station Procedure

No. OPGP03-ZA-0042, "Operations Quality Records," dated December 31, 1987,

describes the requirements and responsibilities for the control of QA

reccrds. Paragraph 4.3.4 of this procedure states, "Retention time in

Division QA Record files shall not exceed 90 days", while paragraph 4.3.5

states, "A Record Custodian, who will be responsible for filing records,

maintaining accountability, controlling access, index stored record and

transmitting those records to the Record Retention Area, should be

designated in writing."

This deviation is considered closed based on the overall licensee

compliance with comitted corrective actions. The licensee's failure to

properly maintain a divisional QA record file in the engineering

department is contrary to the requirements established in Station

Procedure OPGP-03-ZA-0042, paragraphs 4.3.4 and 4.3.5 and constitutes an

apparentviolation(498/8807-01).

3. Followup Inspection of Welding of Structural Steel (55100)

During the period January 25-29 and February 8-12, 1988, the NRC inspector

conducted a followup inspection to determine through direct observations

and records review, whether the structural welding activities performed at

the site are performed in accordance with specifications, procedures, and

Safety Analysis Report (SAR) commitments to the American Welding Society

D1.1 Code,

a. Observation of Work

The NRC inspector made a random selection of 118, Category B, field

welds for direct visual examination. The field welds selected were

composed of electrical raceway hanger welds (25), pipe support

welds (56), and HVAC welds (37). In conducting the visual

examinations, the NRC inspector utilized the 11 attributes prescribed

by the AWS D1.1-85 Code, "Visual Weld Acceptance Criteria" and

licensee Procedure SSP-16, "General Structural Welding Requirements "

Revision 3. The criteria are applicable to structural systems

subject to static loading (seismic loads included) for which fracture

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resistance and fatigue resistance are not governing design

considerations.

No violations or deviations were identified,

b. Review of Weld Records

In conjunction with the visual examination of field welds, the NRC

inspector selected the corresponding 118 weld records for the welds

examined and an additional 50 weld records for review.

The records consisted of 29 travelers and drawing and weld maps for

the selected welds. In addition to the AWS Code requirements, the

NRC inspector utilized the licensee's Procedures SSP-11

"Fabrication, Erection, and Bolt-up of Structural and Miscellaneous

Steel," Revision 2 and SSP-16 "General Structural Welding

Requirements," Revision 3, which delineate requirements for recording

and documenting field welding activities.

The NRC inspector observed that the entries on the weld records were

consistent with AWS D1.1 Code and Procedure SSP-11 requirements,

which include welder identification, weld process used, electrode

traceability, preheat temperature, weld identification traceabP to

specific component, QC inspector signoffs, and approvals.

No violations or deviations were identified,

c. Records Review - Welders and Weld Inspectors Qualifications

The NRC inspector selected eight welders and five QC weld inspectors

training and qualification records for the period March 11, 1986,

through September 23, 1987. The NRC inspector observed that welder

records reflect that all welders were qualified in accordance with

established licensee Procedure SSP-31, "Welder Qualification," and in

accordance with Section 5 of AWS D1.1 Code requirements. In

addition, the licensee maintains a continuous computer data record

system establishing the qualification status of all welders.

During the review of QC weld inspector records, the NRC inspector

observed that individual inspector training and certification records

were well documented in the specific training and certifications

received, including ASME and AWS Code requirements. In addition,

each inspector's records indicated specific training in the visual

acceptance criteria of SSP-16 for structural welds.

No violations or deviations were identified.

4. Nuclear Welding (55050)

As a folicwup inspection to NRC Inspection Report 50-499/88-02, the NRC

inspector reviewed the records associated with the postweld heat treatment

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of welds SB1101-FW0009, SB1201-FW0009, SB1301-FW0009, and SB1401-FW0010

for Unit 1, and SB2101-FW0009, SB2201-FW0009, SB2301-FW0009, and

SB2401-FW0010 for Unit 2. These welds are the penetration assembly to

penetration sleeve welds in the steam generator blowdown piping system.

The penetration assembly and sleeve material specifications are SA182

Grade F22 and SA333 Grade 6, respectively. Bechtel Specification

SA010PS002, Revision 13, specifies that the penetration assembly to

penetration sleeve weld shall be in accordance with ASME III, Division 1,

1974 Edition through Winter 1975 Addenda, Subsection NE or Subsection NC

when criteria for welding postweld heat treatment, or material is not

provided in Subsection NE. The penetration assembly material is

classified as a P-Number 5 material in ASME Section IX. P-Number 5

materials are included in Subsection NC, but are not included in

Subsection NE; therefore, the requirements of Subsection NC are applicable

to the above penetration weld. Paragraph NC-4600 of Subsection NC

specifies that P-Number 5 materials shall be postweld heat treated at 1250

to 1400 F. Paragraph NC-2400 of Subsection NC specifies testing of all

welding material used in construction and that the test coupons shall be

I postweld heat treated to the specified temperature indicated in the

welding procedure specification. The welding procedures, WP-129 and

WP-69, specify postweld heat treatment temperatures of 1300 to 1400 F and

1325 to 1375 F, respectively. The welding materials used for making these

welds were tested using coupons postweld heat treated at temperatures of

1100 to 1200 F. The failure to test welding materials in accordance with

the postweld heat treatment requirments of the applicable welding

procedure specif" stions is an apparent violation. (498/8807-02;

499/8807-01)

5. Unit 2 Safety-Related Pipe Support and Restraint Systems (50090)

a. Observation of Work and Work activities

The NRC inspector observed the following small bore pipe supports:

Support Drawing Class

CV 9141-HS 5003 1

RC 9419-HS 5001 1

RC 9419-HS 5006 1

CV 2142-HF 5039 2

CV 2142-HF 5041 2

CV 2142-HF 5044 2

SI 2306-HF 5006 2

SI 2306-HF 5005 2

, SI 2306-HF 5004 2

SI 2319-HF 5001 2

SI 2319-HF 5002 2

CC 9129-HS 5001 3

CC 9229-HS 5001 3

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The NRC inspector also observed the crossbracing weldments for the

2A, 28, and 2C residual heat removal pump supports.

In the areas inspected, the supports and weldments were consistent

with the requirements of the drawings. The supports were inspected

for type, location, dimensions, orientation, clamps, bolting, and

clearances. The weldments were inspected for size and appearance.

b. Records

The NRC inspector reviewed the quality control records for the pipe

and component supports identified above.

In the areas reviewed, the records were complete, accurate, and

retrievable.

No violations or devit.tions were identified.

6. Unit 2 Safety-Related Components (50073 and 50075)

An inspection was conducted of activities related to selected

safety-related components other than reactor pressure vessel and piping.

This inspection was performed to determine whether specific activities

associated with the reviewed components were being controlled and

performed according to NRC requirements, FSAR commitments, and licensee

procedures.

a. Work Observations

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The NRC inspector examined the following equipment for which work had

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been completed or was in progress to determine conformance with the

applicable procedural requirements:

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Pressurizer Power Operated Relief Valve (PORV) Nos. 2RC-PVC-655A

and -656A

Pressurizer Spray Valve Nos. 2RC-PVC-655B and -655C

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Pressurizer Safety and Relief Valve Discharge Header Serial

No. 39813

Steam Generator PORV No. A2MS-PV-7411

i Regenerative Heat Exchanger No. 2R172NHX201A

Three-Way Power Operated Valve No. B2CU-FV-3123

All attributes that could be visually inspected were examined fo-

adequacy of design and completeness of construction. Particular

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areas examined included welding, bolting identification, switches,

restraints, locking devices, flow direction, and cleanliness. No

problems were identified by the NRC inspector in this area of

inspection.

b. Records Review

The NRC inspector reviewed the work packages associated with the

installation of the previously listed equipment. Documents reviewed

included installation drawings, mechanical equipment installation

travelers, valve checklists, nondestructive examination reports,-

nonconformance reports (NCRs), material list, and process data

checklists. The manufacturer's ASME Code Data Reports were reviewed

for the steam generator PORV and the 3-way power operated valve.

Also included in the records review was the examination of the

fabrication package for the pressurizer safety and relief valve

piping and support assembly. The package contained data associated

with the actual fabrication of the assembly and the mater ial tests

ar d certification reports.

The records were reviewed for attributes required by the codes or

specification from which they were fabricated to and also for

retrievability, completeness, and legibility. No problems or

discrepancies were identified by the NRC inspector during review of

the installation and fabrication records.

No violations or deviations were identified.

7. Review of Unit 2 Preoperational Test Procedures

The NRC inspector reviewed the following preoperational test procedures:

a. Procedure 2-RC-P-01, "Reactor Coolant System Cold Hydrostatic Test,"

Revision 0, dated October 23, 1987. (70362)

The

leak-tightness objective of thisreactor

of the procedure coolant wassystem

to varify(the

RCS) and integrity and

the associated

systems that form the RCS boundary. This scheduled hydrostatic test

of the primary system is performed to meet the requirements of the

FSAR, Section 14.2.12.2(73), and the ASME Boiler and Pressure Vessel

Code,Section III, Division 1, Class 1 requirements. This primary

system hydrostatic test will be performed at a test pressure of

3107 (+20, -0) psig and a system temperature greater than 150 F, but

less than 250 F.

The NRC inspectors identified several potential problem areas (e.g.,

resolve relief valve pressure relief setting of 3185 (+65, -0) psig

versus lower weld on tube sheet to tube maximum pressure of

3121 psig; establish provisions for completing hydrostatic test if

1 of the 2 calibrated gauges should fail; and appropriate protection

for the suction side of the positive displacement pump from

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overpressure) and discussed these areas with cognizant Unit 2

licensee personnel. These identified areas and other areas of lesser

concern will be reviewed and evaluated by the licensee's technical

staff prior to initiation of the scheduled primary system hydrostatic

test. The NRC inspectors will "re-review" the procedure, if

revisions are made prior to performance of the scheduled cold

hydrostatic test of the RCS. These potential problem areas would not

have resulted in a reduction in the safe operation capabilities of

the plant or the scheduled performance of this test, if these items

had not been identified.

b. Procedure 2-SP-P-01, "Solid State Protection System Reactor

Protection Logic Test," Revision 0, dated November 13, 1987. (70305)

The objective of this test is to verify that the reactor protection

logic functions as designed and that the solid state protection

system (SSPS) internal logic, excluding inputs and outputs, is in the

correct configuration using the installed test equipment. This test

will partially satisfy the requirements stated in Chapter 14,

Section 14.2.12.2 (42.b.2) of the FSAR by verifying the combinational

logic internal to the SSPS logic trains functions, as designed, and

by verifying that the combinational logic associated with the SSPS

inputs is correct. The proper SSPS internal logic configurations

must demonstrate functions as designed, using the installed test

equipment to meet acceptance criteria.

The NRC inspector did not identify any technical, safety, or

operational problems in this test procedure.

c. Procedure 2-SP-P-02, "Solid State Protection System," Revision 0,

dated September 25, 1987. (70305)

The purpose of this SSPS test is to verify the master relay - output

relay configuration is installed and will function as designed. This

test is to verify continuity from the master relays to the respective

output relays, using the fnstalled test system and the approved

procedure. This te.tt will partially satisfy the requirements stated

in Chapter 14, Sect; n 14.2.12.2 (42.b.1) of the FSAR. The test is

designed to verify that the output relays test panel functions as

designed and that the setpoints associated with the SSPS inputs are

within the specified tolerances. The test must demonstrate proper

relationship between master relay and the output relays by verifying

continuity from the mater relays to the appropriate output relays,

using the installed test equipment to meet acceptance criteria.

The NRC inspector did not identify any technical, safety, or

operational problems in this test precedure,

d. Procedure 2-RC-P-02, "Hot Functional Test," Revision 0, dated

November 24, 1987.

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The objectives of this procedure are:

(1) Provide a guideline for the sequence of event and testing that

is to be performed during the initial primary system heatup,

testing at normal primary coolant system temperatures, and plant

cooldown at. completion of the testing.

(2) Provide a record that the required reactor coolant pump (RCP)

operating time is at least 240 hours0.00278 days 0.0667 hours 3.968254e-4 weeks 9.132e-5 months at full flow operation with

at least 120 hours0.00139 days 0.0333 hours 1.984127e-4 weeks 4.566e-5 months of this operating time at a temperature of at

least 525 F. Two hundred forty hours of RCP running time is

adequate to provide the required hours at temperature on the

reactor vessel internal components.

(3) Verification of plant operating procedures during the scheduled

evolutions involved during the hot functional testing (HFT),

including:

(a) heatup of RCS to normal ope.iting temperature,

(b) solid plant control.

(c) cocidown of RCS to ambient conditions,

(d) degassing of the RCS,

(e) RCS temperature control,

(f) RCP operation,

(g) pressurizer steam bubble formation and collapse of the

steam bubble (plant going into solid RCS status),

(h) charging and letdown operation, and

(i) cooldown from hot no load from outside the control room.

Acceptance criteria included satisfactory performance of a solid

plant pressure control test, opening of the PORVs in two seconds or

less with the RCS at hot no-load conditions, and the specific

acceptance criteria stated in the various procedures that are

scheduled to be performed during the HFT.

The tests scheduled to be performed during the HFT should demonstrate.

that the HFT will meet the requirements stated in

Section 14.2.12.2 (98), "Reactur Coolant System Hot Functional

Preoperational Test Summary," of the FSAR.

The NRC inspector did not identify any technical, safety, or

operational problems in procedure,

e. Procedure 2-SI-P-02, "Safety Injection Accumulators," Revision 0,

dated November 18, 1987.

The objective of this test procedure was to verify that the safety

injection (SI) accumulator system discharge and operational

prformances in the cold unpressurized RCS meet design requirements.

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The NRC inspector discussed some specific comments that could improve

the overall comprehension of this procedure. The comments, if

implemented by the licensee, would not change the safety

considerations of this procedure,

f. Procedures 2-SI-P-01, "Safety Injection System Train A," Revision 0,

- dated November 5,1987, and 2-SI-P-04, "Safety Injection System

Train B," Revision 0, dated November 5, 1987. (70301)

}

The objectives of these two test procedures were to demonstrate

proper operation of:

(1) High head safety injection (HHSI) and low head safety

[ injection (LHSI) pump controls and interlocks, including

g response to SI signals and load sequencer start and stop

a signals.

(2) SI system valves to an SI signal and to a containment isolation

phase A signal.

(3) Containment sump isolation valves interlock w'en reactor water ,

storage tank outlet valves.

The tests scheduled to be performed in accordance with each of these

two test procedures should demonstrate that each train of the SI

system functions as required by design and will meet the requirements

stated in Section 6.3, "Emergency Core Cooling System," and

_ Section 14.2.12.2 (76) of the FSAR.

The NRC inspector requested clarif'.0 br on three groups of wording

on these two procedures. The cit.: P.atic.,s were related to general '

.

topics and would not affect the si iv s>gr.ificance of these

procedures.

g. Procedure 2-RS-P-01, "Rod Control System," Revision 0, dated

December 17, 1987. (70332)

The objectives of this preoperational test were to demonstrate that

the control rod drive system (CRDS) functions in automatic and manual

(programmed / individual bank) modes, under no-load (control rod drive

mechanisms (CRDMs) not connected) conditions and also to demonstrate

that the CRDS protective and control functions operate correctly,

providing rod stop capabilities. The scheduled performance of this

test of the CRDS should meet the requirements stated in

Section 7.7.1.1, "Control Systems Not Required for Safety; Rod

Control System, Amendment 61," and Section 14.2.13.2 (47), "Control

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Rod Drive System Preoperational Test Sumary, Amendment 61," of the

FSAR.

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The acceptance criteria were established to assure that: (1) the

CRDS functions as designed in the various modes; and (2) the CRDS

protective, control and alarm functions operate as designed to verify

the various CRDS functions during testing of stop rod motion on a

control rod stop or plant interlock.

The NRC inspector did not identify any technical, safety, or

operational problems in this test procedure.

No violations or deviations were identified.

8. Inspection of the Location of the Manual Reactor Trip Circuit in the

South Texas Project, Units 1 & 2, 55P5 (92703)

The inspection of the STP (Units 1 & 2) SSPS was prompted by

IE Bullei.in G5-18.

The objective of this inspection was to determine whether the licensee's

controlled drawings of the SSPS correctly depict the actual location of

the manual trip circuit and confirm that the manual trip circuits are

located downstream of the output transistors, Q3 and Q4, in the

undervoltage (UV) output circuit.

The NRC inspector reviewed the controlled drawings of the STP,

Units 1 & 2, SSPS. In addition, the NRC inspector examined the installed

card containing the manual reactor trip circuit for Unit 2.

The NRC inspector had no further questions concerning the drawings and the

subsequent hardware.

9. Onsite Design Activities (37055)

An inspection was conducted to determine whether onsite design activities, t

including controls for engineering and construction initiated field

changes, are being conducted in compliance with the technical and QA

requirements described in the FSAR. The licensee has the overall

responsibility for the design and engineering of the facility. The

programmatic description for the implementation of this responsibility and

the requirements imposed on contractors authorized to performed design

work is provided in Section 3.0, "Design Control," of HL&P's Quality

Assurance Program Description (QAPD). The NRC inspector reviewed

Revision 19 of the QAPD and found it to satiFfy the design control

criteria of 10 CFR Part 50, Appendix B.

The licensee has assigned the authority to Bechtel and Westinghouse to

perform the design, engineering, and design verification at STP.

Presently both of these organizations perform desian control functions

with onsite personnel. In 1987, Bechtel expanded the engineering effort

located onsite to include the project engineering personnel formerly

assigned to the Houston field office. This newly formed project

engineering team now performs virtually all the Bechtel-related design

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14

work. Section 3.0, "Design Control," of Bechtel's Project Quality Program

Manual contains the policies and requirements for Bechtel design work.

Westinghouse has an onsite support engineering team (SET) which performs

limited design work. The SET is a multidisciplined group of engineers

operating onsite as an extension of the design engineering functions in

Pittsburgh. The STP Interface Document for Westinghouse Support

Engineering Team provides the scope and responsibilities for onsite

design-related work. The NRC inspector found both of these documents to

be in compliance with the design control requirements imposed by the

licensee in the QAPD.

The following implementing procedures utilized in design-related work at

STP were reviewed by the NRC inspector:

a. Bechtel Engineering Department Procedures (EDPs).

ho. 2.13, ' Project Engineering Team Organization and

Responsibilities," Revision 5

No. 4.1, "Design Criteria and Project Q-List," Revision 6

No. 4.26, "Interdisciplinary Design Review," Revision 0

No. 4.27, "Design Verification," Revision 3

No. 4.37, "Design Calculations," Revision 6

No. 4.46, "Project Drawings," Revision 10

No. 4.47, "Drawing Change Notice," Revision 6

No. 4.61, "Nonconformance Reports," Revision 2

E No. 4.62, "Field Change Request (FCRs)/ Field Change

Notices (FCNs)," Revision 8

No. 4.72, "Configuration Control Package," Revision 6

b. Bechtel Engineering Directive (PED) No. 041, "Design

Checklist (DCL)," Revision 2

c. Bechtel Work Plan Procedure (WPP) No. 201, "Field Change Notice,"

Revision 8

d. STP Standard Site Procedures (SSPs)

No. 8, "Nonconformance Reporting," Revision 4

No. 37, "Configuration Control Package," Revision 3

No. 49, "Field Change Requests," Revision 2

No. 68, "Change Approval Request (CAR)," Revision 0

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The reviewed procedures provide control of engineering design work to

ensure technical and regulatory requirements are met. Also included

in these procedures is the assignment of responsibility and listing

of requirements to assure existing design criteria are not affected

by design-related changes and if so, that design requirements are

properly reviewed. One observation the NRC inspector had concerned

the incomplete incorporation of Bechtel's onsite project engineering

team and associated responsibilities into applicable EDPs. Several

EDPs still have the distinction of a Houston based project

engineering group and a site engineering office. The current

engineering structure should be properly depicted in EDPs to ensure

assigned responsibilities concerning design work is fulfilled by the

engineering staff.

During the present stage of construction at STP, the majority of

design-related activities involve changes to the existing design of

the facility (design changes). The NRC inspector reviewed the

following engineering initiated documents to determine if established

design controls were being observed.

Westinghouse FCN Nos. THXM-10633 (Addition of Limit Switches on

the Refueling Machine) and THXM-10630 (Modification to Reactor

Coolant Water Purity Panel)

Bechtel Configuration Control Package (CCP) Nos. 2-M-FST-244

l (Replace Concentrates Transfer Pump 1B and Add Flow Orifice),

2-M-ST-0235 (Relocate Valve WS 051), and 2-M-FST-0220

l

'

(Modification of Diesel Generator Air Filter Basket Lifting

Device)

FCR Nos. SM-00519, EM-00603, EM-00605, and EM-00084.

The NRC inspector discussed with the Westinghouse cognizant engineer

the design-related details of the two FCNs and observed the completed

work on the Unit I refueling machine. The engineer was very

knowledgeable of the existing design criteria governing these two

jobs and the applicable design control procedures. Proper design

review and approval was apparent during the preparation of these

documents.

The NRC inspector also discussed with Bechtel engineers the design

considerations associated with the three CCPs. While familiar with

procedural design control and the application of these controls to

the CCPs, the Bechtel engineers were not as knowledgeable in the

overall content of the jobs as the Westinghouse engineer. The Design

Checklist (DCL) is used by Bechtel project engineering to document

considerations made during the preparation of documents which change

or imoact other project documentation. The DCL is an effective

methou v/ ensuring responsible engineers make conscious decisions

'

concerning the proposed change and how existing plant design may be

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16

affected. During review of the CCP that modifies the lifting device

on the diesel generator air filter basket, the NRC inspector noted

the calculation for sizing the lifting . lug attachment weld was not in

the work package. The Bechtel engineers informed the NRC inspector

that an informal calculation concerning weld size was performed

during CCP preparation. Considering the addition of lifting lugs was

a new design for the filter basket, this type documentation should be

included in the job package.

The reviewed FCRs received the proper engineering review of the

proposed change. The affected drawings, while not ctrrently revised

to reflect the FCR change, were posted with the appropriate FCR. The

NRC inspector did note that one of the affected drawings exceeded the

maximum number of amendments that are allowed to be posted for a

drawing. EDP No. 4.46, "Project Drawings," allows an aggregate total

of ten amendments, before a drawing must be updated to incorporate

the changes. Greater attention is required to updating drawings to

prevent excessive number of amendments from overly complicating

drawings used in project construction and modification.

The design-related activities at STP effectively control changes that

can impact design criteria and requirements. STP procedures provide

the required controls to ensure design consideration are addressed.

While the NRC inspector noted some minor concerns during the review

of design change documents, the engineering organizations with design

responsibility do appear to be properly controlling design

requirements.

No violations or deviations were identified.

10. Exit Interview

The NRC inspectors met with the licensee representatives denoted in

paragraph 1 on January 29 and February 12, 1988, respectively, and

summarized the inspection scope and findings. The licensee commented that

the testing of welding material at different postweld heat treatment

,

temperatures should be considered an ASME Code interpretation problem and

I not a violation.

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ML20148G687

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