IR 05000424/1986076
| ML20215E772 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 09/23/1986 |
| From: | Blake J, Hallstrom G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20215E709 | List: |
| References | |
| 50-424-86-76, 50-425-86-37, NUDOCS 8610150503 | |
| Download: ML20215E772 (9) | |
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c, e at?u UNITED STATES
'o NUCLEAR REGULATORY COMMISSION y"
REGION 11
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101 MARlETTA STREET, N.W.
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ATLANTA, GEORGI A 30323
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Report Nos.:
50-424/86-76 and 50-425/86-37 Licensee: Georgia Power Company P. O. Box 4545
. Atlanta, GA 30302 Docket Nos.:
50-424 and 50-425 License Nos.: CPPR-108 and CPPR-109 Facility Name:
Vogtle 1 and 2 Inspection Conducted: August 18-22, 1986 9[#
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Inspector:
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litr Dat6 Signed 9 g3
Approved by.
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J J B ake, Section Chief
'Date' Signed a rials & Processes Section Di ision of Reactor Safety SUMMARY Scope: This routine, unannounced inspection was conducted on-site in the areas of licensee action on previous enforcement matters (Units 1 & 2), housekeeping (Units 1 & 2), materials control (Units 1 & 2), Train A Encapsulation Vessel Bellows Repair (Unit 1) and licensee identified items (50.55(e)).
Results: One violation was identified Failure to complete adequate corrective action to ensure material control from instrumentation satellite warehouse.
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REPORT DETAILS l
1.
Persons Contacted Licensee Employees R. H. Pinson, Vice President, Construction
- D. O. Foster, Vice President and Project Support Manager
- W. T. Nickerson, Assistant to Project Director P. D. Rice, Vice President, Engineering
- C. W. Hayes, Project Quality Assurance (QA) Manager
- E. D. Groover, QA Site Manager, Construction
- G. A. McCarley, Project Compliance Coordinator
- B. C. Harbin, Manager of Quality Control (QC), Construction
- R. W. McManus, Assistant Readiness Review (RR) Project Manager A. N. Lankford, Civil QC Support Supervisor
- C L. Cross, Senior Regulatory Specialist W. C. Gabbard, Regulatory Specialist
- G. E. Spell, Jr., QA Engineering Support Supervisor W. L. Burmeister, Operations Supervisor J. A. Edwards, Senior Nuclear Specialist, Operations
- G. R. Frederick, Senior QA Engineer
- J. L. Blocker, Mechanical Section Supervisor, Construction R. M. Bellamy, Manager of Tests and Outage, Operations Other licensee employees contacted included construction craftsmen, engineers, technicians, operators, mechanics, and office personnel.
Other Organizations
- D. A. Bartol, Westinghouse, Project Engineering Manager D. R. Fraser, Westinghouse - Vogtle Structural Analysis Mobile Unit (V-SAMU), Manager of Vogtle Final Design Verification S. T. Cloyd, Stress Analysis Engineer, V-SAMU A. Tredham, Stress Analysis Engineer, V-SAMU D. L. Kinnsch, Bechtel Power Corporation (BPC), Project Field Engineer (PFE)
D. Niehoff, BPC, Deputy Engineering Group Supervisor L. Case, BPC, Codes and Standards
- P. R. Thomas, BPC, RR Construction Team Leader D. L. Carlson, BPC, Materials and Quality Services (M & QS) Welding Engineer H. Freddy, BPC, PFE Staff Supervisor
- J. W. Carson, BPC, QA Supervisor
- B. L. Edwards, Pullman Power Products (PPP), Resident Construction Manager
- J. E. Miller, PPP, QA Manager
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_ _ _. NRC Resident Inspectors'
- H. Livermore, Senior Resident Inspector (Construction)
- J. Rogge, Senior Resident Inspector (Operations)
- Attended exit interview 2.
Exit Interview The inspection scope and findings were summarized on August 22, 1986, with
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those persons indicated in paragraph 1 above.
The inspector described the areas inspected. No dissenting comments were received from the licensee.
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(0 pen) Violation 424/86-76-01, 425/86-37-01, Failure to accomplish adequate corrective actions to ensure material control from instrumentation satellite l
warehouse, paragraph 5.
(0 pen) Unresolved Item 424/86-76-02, Weld Repair of Train A Encapsulation
l Vessel Bellows Expansion Joint, paragrapi 6.
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The licensee did identify as proprietary some of the materials provideJ to
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and reviewed by tne inspector during this inspection; however, details from
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those materials are not included in this report.
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Licensee Action on Previous Enforcement Matters
a.
(0 pen) Unresolved Item (424/86-03-03, 425/86-02-03) High Strength
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Bolted Connections This item concerns ASTM A325 and A490 high strength steel bolts installed at plant Vogtle.
Discussions with QC inspectors and other cognizant licensee personnel had identified potential overtensioning of bolts on mainplate girder #5 at elevation 240' in the Control Butiding.
The girder was installed in August 1982, and potential overtorquing was observed due to lack of conformance to " turn-of nut" installation requirements within construction procedure CD-T-16. Bolts in girder #5 were replaced.
However, followup discussions with cognizant licensee personnel had
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established additional potential for overtensioning due to the following:
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A common philosophy that bolt overtensioning would not present a
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problem as long as the bolts did not break during installation. A technically supported justification of this philosophy was requested.
- An apparent inability of the QC inspection program (past or present) to identify overtensioning (overtorquing) to near failure limits. Quality Control inspection personnel uniformly stated that 29-T-16 specified only that the minimum required torque be checked; i.e., no check for potential overtorque is required or conducted and QA surveillance of " snug tightening" or application or reference match marks is not required.
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- Installation of high strength bolts by craftsmen who had not received training on the " turn-of-nut" method.
The turn-of-nut method was initiated on Revision 3, dated July 23, 1982, of CD-T-16.
Initial training of construction craftsman was P
completed on September 8, 1982.
- Statements by cognizant licensee personnel that the minimum time frame during which installation by turn-of-nut method could be i
suspect and overtorque a potential problem is from July 23 to September 8, 1982.
Further GPC response on this issue was transmitted by letters to Region II dated April 17, 1986, (Log:
GN-864) and May 12, 1986
(Log: GN-906).
The adequacy of the engineering justification was questioned on several points as detailed within Inspection Report
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i Nos. 50-424/86-39 and 50-425/86-19.
J The inspector completed additional discussions with cognizant licensee
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personnel during this inspection regarding further NRC concern on this issue. Additional licensee response is anticipated by September 15, 1986. This item remains open.
b.
(Closed) Violation (50-424/85-40-04):
Failure to Protect Permanent Plant Equipment GPC's letter of response dated November 1, 1985, has been reviewed and determined to be acceptable by Region II. The inspector examined the corrective actions as stated in the letter of response and further i
reported in paragraph 3.c below regarding similar violation
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424/86-07-02. The inspector concluded that this item was closed based j
on acceptable corrective actions regarding the more recent similar violation.
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(Closed) Violation (424/86-07-02):
Failure to Protect Permanent Plant Equipment)
I-GPC's letter of response dated June 13, 1986, has been reviewed and determined to be acceptable by Region II. The inspector examined the i
corrective actions as stated in the letter of response. The examina-
tion included the following:
Verification of programmatic controls to identify the de-energiza-
tion of required motor heaters and accomplish appropriate corrective actions (startup manual procedure SUM-23).
- Verification of understanding and conformance to SUM-23 by
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operations electrical maintenance personnel.
Field verification of completion of permanent power supplies to
i Train A Nuclear Service Cooling Water (NSCW) Pump Motors and Motor
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Operated Discharge Valves (MOVs).
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Verification of design changes for removal or disconnection of space heaters from NSCV MOVs (support information for 50.55(e) ~
item 424, 425/85-87:
Limitorque Valve Motor Operators).
The inspector concluded that GPC had determined the full extent of the subject violation, performed the necessary survey and followup actions to correct the subject conditions, and developed the necessary corrective actions intended to preclude the recurrence of similar circumstances.
The corrective actions identified in the letter of response were implemented.
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Unresolved Items Unresolved items are matters about which more information is required to determine whether they are acceptable or may involve violations or deviations. One new unresolved item identified during this inspection is discussed in paragraph 6.
5.
Independent Inspection Effort Housekeeping (54834B), Material Identification and Control (429028), and Material Control (429408)
The inspector conducted a general inspection on Units 1 and 2 containments, the control building and the reactor auxiliary building to observe activities such as housekeeping, material identification and control; material control, and storage.
During the above inspection, the inspector noted deficiencies in the physi-cal control of access from the laydown area of the Pullman Power Products (PPP) satellite warehouse for instrumentation sub-storage located near the
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I turbine building (Troy No. 4). The deficiencies included a top-to-bottom
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gap in the security fencing at the rear of the area and the placement of a stairway over an adjoining fuel tank spillage retaining wall which permits access by stepping over the security fence onto the ASME code class storage racks adjacent to the fence. Cognizant PPP supervisory personnel indicated that no action had been taken to repair the fence even though the gap had been reported by warehouse personnel three months previous to the date of the inspection.
PPP personnel further informed the inspector of instances when locks securing the gate to the area had been cut as well as purported instances when ASME code materials had been released without conformance to the material control requirements of PPP procedure IX-40 " Instrumentation installation, inspection and testing." During subsequent review of security reports and discussions with PPP and GPC operations personnel, the inspector verified that locks had been cut and replaced on at least four separate occasions since mid May 1986, and the date of this inspection.
Two instances had occurred due to need for entry by authorized operations
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personnel to enable fuel delivery from tanker trucks. Two instances had been due to unauthorized craft entry during times between shifts when authorized warehouse and QC inspection personnel were absent. Disciplinary action had been taken in one instance but the responsible parties were not identified in the second instance.
The inspector was informed that no followup communication to craft personnel had occurred as a result of the unauthorized lock cutting incidents. The inspector was also informed that an authorized QC inspector with keys was present on-site between shifts to accommodate instrumentation crews working between end of second and beginning of first shift.
The inspector noted that this information was not posted in the immediate area and that need for more efficient communication to the craftsmen as well as between construction and operations personnel was indicated.
-The inspector was unable to verify any lack of conformance to PPP procedure IX-40, but stressed need for adherence to its requirements for material control to both warehouse /QC personnel and other cognizant licensee management personnel.
The inspector further noted that lack of access control to the instrumenta-
l tion warehouse had been the cause of previous violation 85-08-02 and that j
GPC's May 10, 1985 response to the violation had included actions intended to maintain programmatic control of ASME code class instrumentation materials. Therefore, this lack of control of access was considered a lack of conformance to 10 CFR 50, Appendix B, Criterion XVI and will be
identified as Violation 424/86-76-01; 425/86-37-01:
Failure to complete adequate corrective actions to ensure material control from instrumentation satellite warehouse.
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Train A Encapsulation Vessel Bellows Repair (Unit 1) (551888)
The inspector examined welding activities for repair of the bellows expansion joint of Unit 1 Train A isolation valve encapsulation vessel 1-1206-V4-002 to determine whether applicable code and procedure require-ments were met. The applicable code for the isolation valve encapsulation vessel' assemblies (including expansion bellows) isSection III, Subsection NE of the 1974 ASME Code, through summer 1976 addenda.
Weld repair of bellows elements is permitted under code case N-315 dated February 14, 1983.
The inspector examined quality records for the repair as follows:
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Operations Deficiency Report (00R) T-1-86-2797 dated August 8, 1986
PPP welding procedure 27-111/1-8-08-12 l
PPP Supporting Procedure Qualification Records Nos. 110, 132 and 133 The inspector completed field examination of the repair weld and noted that the repair was limited to the outer face of the initial convolution.
Mowever, contrary to the sketch in ODR T-1-36-2797, the weld heat affected zone extends into the highly stressed radius zone of the convolution at the top and bottom of the repair.
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During a subsequent meeting with cognizant licensee personnel, additional engineering justification of the repair (BPC letter dated August 21, 1986,~
to Mr. P. D. Rice and supporting attachments - BPC Log No. PFE-12579) was
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discussed.
The inspector noted that the Vogtle FSAR section 3.6.2.1.4 included the commitment that the encapsulation vessels and their bellows expansion joints conform to ASME class MC requirements.
Section 3.8.2.2 indicates that the requirements of Regulatory Guide (RG) 1.57 also apply and Section 3.8.2.4.4
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commits to design and analysis in accordance with article NE-3000 of the ASME code. The inspector also noted that the Vogtle FSAR Section 1.9.84
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commits to RG 1.84 for the use of ASME Code Cases and that RG 1.84 requires NRC approval prior to use of Code Case N-315.
Cognizant licensee personnel responded that the bellows repair was not conducted to the requirements of Code Case N-315, but had been accomplished in accordance with " guidelines" of ASME Section III and Section XI.
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Cognizant licensee personnel also indicated that the August 21, 1986, letter was intended as the Certification of Compliance that the repair had been
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completed in accordance with guideleines of ASME Section XI and will meet its intended design function.
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i The inspector questioned whether conservatism equivalent to Code Case N-315 had been maintained, since N-315 requires the use of a full scale facsimile i
bellows to qualify the repair procedure by subsequent fatigue testing and proof testing of the facsimile. Cognizant licensee personnel contended that
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adequate design margin existed.
The report from a recent joint testing i
program between Pathway Bellows and a Region II licensee was referenced as i
justification that this type of repair would require accepting a design
penalty of loss of about 60% of average cyclic life on a worst case basis.
Calculations by BPC with the 60% penalty indicated that design requirements i
of NE-3365.2(g) were satisfied. The inspector noted that the referenced
report stressed in its' conclusion that prototype tests should continue to be required for repairs to existing bellows and anticipated that those repairs would conform to Code Case N-315 requirements.
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The inspector further noted that the statistical stress factors K used in ss
meeting NE-3365.2 Design Requirements was influenced by the number of
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replicate tests.
Code Case N-315 requires that the number of replicate tests be taken as zero when calculating K The referenced design j
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calculations completed by Johnson Controls for the initial bellows used a
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value of ten for the replicative number of tests and were no longer valid.
Cognizant licensee personnel requested additional time for further BPC engineering evaluation on this issue.
The inspector informed cognizant licensee personnel that NRC concern regarding ASME Code Conformance for this
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repair and the confirmed capability of the bellows to meet its intended
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design function would be identified as Unresolved Item 424/86-76-02, Weld i
l Repair of Train A Encapsulation Vessel Expansion Joint.
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7.
Licensee Identified Item (92700)
Prior to the inspection, the licensee identified the following items under 10 CFR 50.55(e)
a.
(Closed) Item 424, 425/CDR 81-19:
Design Calculations Concerning Primary Loop Crossover Leg Pipe Restraints On July 6,1981, the licensee notified Region II of a 50.55(e) item concerning an inappropriate design assumption for the primary loop crossover leg restraints. The initial load calculations assumed a free standing crossover leg restraint fixed at the basemat and neglected the presence of the surrounding 2' 9" thick fill slab. Other unrelated design deficiencies were also identified due to the lack of considera-tion of bending stress on shear lugs for steam generator lower lateral support embeds and reactor coolant pump tie rod embeds. A reanalysis was completed and corrective actions required installation of stiffener plates in the restraints and embeds involved.
The final report was submitted on November 11, 1981. The final report has been reviewed and determined to be acceptable by Region II. The inspector held discussions with responsible licensee representatives and reviewed supporting documentations to verify that corrective actions identified in the report have been completed. The item is considered closed.
b.
(Closed) Item 424, 425/CDR 84-73:
NSCW Piping System Design Temperatures i
On December 18, 1984, the licensee notified Region II of a 50.55(e)
item concerning NSCW piping system design temperatures.
The analysis of some of the NSCW piping system was completed based on a design temperature of 150'F. However, ultimate heat sink analyses indicated that a design temperature of 250*F should be used for a portion of the piping to accommodate increased temperature during a design basis accident. Corrective actions required the reanalysis of the associated piping stress calculations for a 280 F design temperature and attendant redesign of piping supports.
The final report was submitted on January 17, 1985. During review of the stress calculations associated with the 280*F design temperature the inspector noted apparent inconsistencies with the final report due to inclusion of NSCW piping with redesigned supports which remained at a 150'F design temperature.
Clarification was obtained that the support redesign for these lines was due to transfer of loads from lines nearby which were analysed at 260*F.
Copy of a letter of clarification to Region II dated August 22, 1986, was provided to the inspector as additional information. The inspector reviewed additional supporting documentation to verify that corrective action has been completed. This item is considered close _
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(Closed) Item 424, 425/CDR 85-81: Turbine Pedestal Seisaic Calcula-tions On June 24, 1985, the licensee notified Region II of a potential 50.55(e) item concerning the seismic design calculations for the turbine building and turbine pedestal.
The specific concern was
associated with the turbine building seismic analysis involved in the determination of the gap between the turbine building and the turbine generator pedestal and the gap between the structural frames.
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The final report was submitted on November 11, 1985.
The licensee concluded that this item was not reportable since all design criteria were met with the existing gap conditions when the turbine analyses were subjected to loads associated with the postulated design basis seismic event. The final report has been reviewed and determined to be
acceptable by Region II.
The inspector held discussions with responsible licensee representatives and reviewed supporting documentation to verify the conclusions reached on the final report.
This item is considered closed.
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d.
(Closed) Item 424, 425/CDR 85-90:
Anchor / Darling Main Feedwater Isolation Valves
On December 4, 1985, the licensee notified Region II of a 50.55(e) item
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concerning the main feedwater isolations valves (MFIVs) furnished by j
Anchor / Darling Valve Company. The deficient condition was associated with the air check valves of the MFIV hydraulic actuator.
It was
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determined that the air check valves do not seal reliably under a l
gradual loss of air supply. The resultant is the prevention of proper i
positioning of the hydraulic actuator 4-way control valves which adversely impacts MFIV closure time.
Corrective actions include replacement of all air check valves with a modified check valve designed to seal reliably under a gradual loss of air supply.
The final report was submitted on January 2,1986. The final report has been reviewed and determined to be acceptable by Region II. During review of supporting documentation for Unit 1 (Nuclear Plant Mainte-nance Work Orders Nos. 18513020, 18513021, 18513019, and 18513022), the inspector noted a lack of clarification regarding replacement with
" modified" check valves.
Field verification of replacement with
" modified" check valves was completed by an NRC resident inspector and is documented in Inspection Report No. 50-424/86-74.
Replacement of air check valves for Unit 2 is not complete. However, this item is closed based on completion of corrective action for Unit 1 and existence of programmatic controls for completion of corrective action on Unit 2.
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