ML20148G687
| 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: |
| References | |
| 50-498-88-07, 50-498-88-7, 50-499-88-07, 50-499-88-7, NUDOCS 8803290203 | |
| Download: ML20148G687 (16) | |
See also: IR 05000498/1988007
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APPENDIX B
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
NRC Inspection Report:
50-498/88-07
Operating License:
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:
Inspection Conducted:
January 25-29andFebrud 8-12, 1988
Inspectors:
a
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R. C. Stewarf,-Reactor Inspect 6f, Materials
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and Quality Programs Section, Division of
Reactor Safe y
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L. D. Gilbert, Rsactor Inspector, Materials
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and Quality Programs Section, Division of
Reactor Safety
bi YtwuenW&
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D. M. Hunnicutt, Project Engineer, Reactor
Date
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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
3b7 $$
/
I. Barnes, Chief, Materials and Quality Programs
Date
Section, Division Reactor Safety
UdO3290203 G9031e
DR
ADOCK 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
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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
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SI 2319-HF 5001
2
SI 2319-HF 5002
2
CC 9129-HS 5001
3
CC 9229-HS 5001
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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
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
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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
Regenerative Heat Exchanger No. 2R172NHX201A
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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 objective of this procedure was to varify(the integrity and
leak-tightness of the reactor coolant system RCS) 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 <br />0.0667 hours <br />3.968254e-4 weeks <br />9.132e-5 months <br /> at full flow operation with
at least 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> 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,
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dated November 5,1987, and 2-SI-P-04, "Safety Injection System
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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
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injection (LHSI) pump controls and interlocks, including
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response to SI signals and load sequencer start and stop
signals.
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(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
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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
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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
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topics and would not affect the si iv s>gr.ificance of these
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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
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Control System, Amendment 61," and Section 14.2.13.2 (47), "Control
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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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
No. 4.62, "Field Change Request (FCRs)/ Field Change
E
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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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
,
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not a violation.
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