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{{Adams | |||
| number = ML20155B973 | |||
| issue date = 06/03/1988 | |||
| title = Insp Repts 50-499/88-24 on 880405-0502.Violations Noted. Major Areas Inspected:Tagging Status of Equipment,Cable Protection,Followup on Previous Insp Findings & Followup on Licensee Reported Significant Const Deficiencies | |||
| author name = Constable G, Garrison D, Hunnicutt D | |||
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) | |||
| addressee name = | |||
| addressee affiliation = | |||
| docket = 05000499 | |||
| license number = | |||
| contact person = | |||
| document report number = 50-499-88-24, NUDOCS 8806140038 | |||
| package number = ML20155B942 | |||
| document type = INSPECTION REPORT, NRC-GENERATED, INSPECTION REPORT, UTILITY, TEXT-INSPECTION & AUDIT & I&E CIRCULARS | |||
| page count = 13 | |||
}} | |||
See also: [[see also::IR 05000499/1988024]] | |||
=Text= | |||
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APPENDIX B, | |||
, | |||
. ;, . U.S. NUCLEAR REGULATORY COMMISSION- . | |||
> ' REGION IV- | |||
NRC Inspection. Report: 50-499/88-24 Construction Permit: CPPR-129- | |||
- | |||
Docket: 50-499 CP Expiration Date: December 1989 , | |||
Licensee: Houston Lightir.g & Power Company (HL&P) ' | |||
P.O. Box 1700 | |||
Houston, Texas 77001 | |||
" | |||
Facility Name: South Texas Project, Unit 2 (STP) | |||
Inspection 'At: STP, Ma,tagorda County, Texas | |||
Inspection Conducted: April 5 <throuch May 2,1988 | |||
, | |||
r | |||
Inspectors: / h www | |||
D. L. Garrison, Resident Inspector, Reactor | |||
~ | |||
dA | |||
Date / | |||
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Project Section D, Division;of Reactor | |||
Projects | |||
hY/ Yu-> af | |||
D. M. Hunnicott, Senior Reactor Inspector | |||
.d3b? | |||
Date / | |||
Reactor Project Section D, Division of | |||
Reactor Projects | |||
Approved: h orAnc.' | |||
G. L. Constable, Chief, Reactor Project | |||
d/3/87 | |||
Date ' | |||
Section D, Division of Reactor Projects | |||
8806140038 880607 | |||
{DR ADOCK 05000499 | |||
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Inspection Summary | |||
. | |||
Inspection Conducted April'5 through May 2,1988 (Report 50'-499/88-24) | |||
Areas Inspected: Routine, unannounced inspection including tagging. status of | |||
' | |||
equipment, cable protection, followup on previous inspection findings, followup 4 | |||
on licensee reported significant construction. deficiencies, review of 10 CFR | |||
.Part 21 reporting program, reactor vessel and internal work observation, | |||
.t | |||
. | |||
reactor vessel and internal .QA review, standby diesel generator. expansion seals | |||
replacement, and safety-related piping installation observations,- | |||
4 s- | |||
Results: .Within the seven areas inspected, three apparent violations were | |||
widentified (improper tagging of equipment, paragraph 2; inadequate cab'e | |||
protection, paragraph 3; and inadequate housekeeping, paragraph 10). | |||
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DETAILS- | |||
0 | |||
1. Persons Contacted | |||
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HL&P | |||
*J. S. Phelps, Supervising Engineer, Project Compliance- | |||
*S. M. Head, Support Licensing Engineering | |||
*W. G. Wellborn, Supervising Project Engineer | |||
*S. D. Phillips, P Z Engineer' | |||
*X. M. 0'Gara, Project Compliance | |||
*T. J. Jordan, Project Quality Assurance Manager | |||
*G. Ondriska, Startup Engineer | |||
*A. R. Mikus, General Supervisor | |||
*T. Quirk, General Supervisor, Records Management System | |||
Bechtel | |||
*R. D. Bryan, Field Construction Manager | |||
' | |||
*R. H. Medina, Quality Assurance Supervisor | |||
, Ebasco | |||
*R. Abe'i, Quality Program Site Manager | |||
*R. C. Sisson, Site Resident Engineer | |||
In addition to the above, the NRC inspectors also held discussions with | |||
various licensee, architect engineer (AE), constructor and other contractor | |||
personnel.during this inspection. | |||
* Denotes those individuals attending the exit interview conducted on | |||
April 22, 1988. , | |||
2. Tagging Status of Equipment - Unit 2 (50071) | |||
Numerous procedures on the site required tagging to reflect the correct | |||
status of the items tagged. During the inspection in the "B" Isolation | |||
Valve Cubicles (IVC), five instances were observed by the NRC inspectors | |||
where the tagging on valves and components did not reflect the current | |||
status of the tagged equipment. The observed incorrect tagging is listed | |||
.'' below: | |||
* One tag indicated a valve to be temporarily installed. The valve was | |||
permanently installed. , | |||
* Two tags indicated parts were removed or to be removed and had been | |||
on the equipment for three years. The required action was found to | |||
be abandoned. | |||
4 | |||
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,0ne. gate valve was found with only the body intact in the line. No | |||
status or tagging was on or near the valve. | |||
One secondary system hydrostatic test tag was observed which should | |||
have been removed before the primary hydrostatic test was perfonned. | |||
During the review of the procedures, the NRC inspector concluded that | |||
clarification in some instances was needed for removal and change of | |||
status. This problem could be a plant generic problem. The licensee's | |||
failure to maintain current status tags on equipment (tags that were | |||
invalid or did not correctly identify the status of components) is an | |||
apparent violation (499/8824-01) of NRC requirements and licensee | |||
commitments. , | |||
, | |||
3. Cable protection - Unit 2 (51065 and 51063) | |||
During the inspection of the "B" Isolation Valve Cubicle on April 20 | |||
and 21, 1988, and previous inspections of the electrical switchgear rooms | |||
in the electrical auxiliary buildin | |||
there were numerous (more than instances 25) g, the of | |||
NRC inspectors | |||
failure observed | |||
to install cable that | |||
softeners over the sharp edges on some of the cable trays. The sharp edges | |||
on cable trays were observed in the cable trays where electrical cables | |||
were routed from one cable tray to another cable tray. Licensee procedures | |||
are clear concerning this item. The NRC inspection results indicated that | |||
failure to install cable softeners could be a generic plant problem. The | |||
licensee's failure to provide cable softeners, where required, to preclude | |||
damage -to electrical cableijacket material is an apparent violation | |||
(499/8824-02) of NRC requirements and licensee coninitments. | |||
4. Followup on Licensee Reported Significant Construction Deficiencies | |||
(10 CFR 50.55(e)) (36100 and 92/00) | |||
(Closed) Incident Review Committee (IRC) No. 398 | |||
This item involved a finding that if the standby diesel generator (SDG) | |||
was being operated in the testing mode and a loss of offsite power (LOOP) | |||
occurred, the electrical supply breakers to the 480 VAC load centers would | |||
be tripped. Then, since the SDG would be running at design speed, the | |||
breakers would be signaled to close shortly thereaf ter. The signal to | |||
close prior to full spring recharge would cause the antipump feature on | |||
the electrical supply breakers to the 480 VAC loads to lock out the | |||
breakers. The breaker lockout would prevent energizing of all 480 VAC | |||
loads connected to the affected train. | |||
Configuration control package (CCP) 2-E-ST-833, Revision 0, dated | |||
August 20, 1987, was issued to add an Agastat relay (6-second setting) in | |||
the LOOP control circuit f rom the sequencer to ensure that the DG breaker | |||
does not close until the 480 VAC circuit breaker closing springs are | |||
, | |||
cha rged. Site electrical discipline personnel had been reinstructed in | |||
the proper design methods assnciated with conditions to prevent | |||
recurrence. This item is considered closed. | |||
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(Closed) IRC No. 400 , | |||
This item involved concerns related to inadequate cooling of the Unit 2 | |||
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high voltage cubicle panels for the SDG. HL&P completed and installed a | |||
design, modification to provide additional cooling for the Unit ~2 SDG " | |||
panels. A 100-hour test run of the SDG was completed successfully. No | |||
recurrence of overheating conditions were observed. This item is- | |||
considered closed. | |||
(Closed) IRC No. 402 | |||
This item involved a finding that some motor shaft-to-pinion gear keys | |||
sheared (failed) in Limitorque Model SMB-0-25 operators. The shearing of | |||
the keys was apparently due to'the keys being machined from incorrect or | |||
defective material. Twelve motor shaft-to-pinion gear keys in valve , | |||
operators in Unit 2 (Deficiency Evaluation Report 87-046) were replaced | |||
with keys manufactured from AISI 1018 steel. According to the material | |||
specifications, AISI 1018 steel is the correct material. This item is | |||
. | |||
considered closed. | |||
(Closed) IRC No. 403 | |||
This item involved Class 1E electrical cable splices which utilize Raychem | |||
^ insulation material and which had been identified as nonconforming after | |||
final QA-acceptance. Three different problem areas were identified by the | |||
licensee. The licensee reviewed the completed cables and terminations; ' | |||
reviewed a computer resort of cables and terminations to validate previous | |||
data; reviewed data on electrical penetration termination cards to assure | |||
that electrical splices to electrical penetrations had been reinspected; | |||
reviewed the transfer of data from master splice lists to the reinspection | |||
data lists to identify any omissions; and reviewed the qualifications and | |||
. work performed by off-project Bechtel personnel to assure that the | |||
methodology and reviews had been performed correctly. This item is | |||
considered closed. | |||
4 | |||
(Closed) IRC No. 408 | |||
This item involved the failure of tubes in the component cooling | |||
water (CCW)heatexchangers. The tube failures resulted from shell side | |||
flow induced vibrations. Design modifications to the heat exchangers were | |||
o required to prevent further damage. In each CCW, 30 tubes were removed in | |||
the areas above and below the impingment plate. Two hundred sixty-four | |||
tubes were rodded with steel rods inserted to dampen vibration ar.d to | |||
decrease the vibration level in the adjacent inboard tubes. The vacated | |||
tube sheet holes and the rodded tubes were plugged with plugs manufactured | |||
. | |||
from aluminum bronze material (similar to and compatible with the | |||
tubesheet cladding and essential cooling water (ECW) piping materials). | |||
1he ECW flow (design flow rate was 15,000 gpm) and pressure drop through | |||
the heat exchanger tubes have been recalculated for the modified condition | |||
of the heat exchangers. The seismic qualification and the weight increase | |||
due to the modification of the heat exchangers have been reevaluated. | |||
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Other safety-related heat exchangers have been evaluated to determine if | |||
any of these heat exchangers would be susceptible to the problems observed | |||
in the CCW heat exchangers. None of the heat exchangers evaluated were | |||
' | |||
found to be susceptible to these identified problems. The changes that | |||
have been made to the CCW heat exchangers to resolve the vibration induced | |||
problems do not alter previous FSAR commitments. This item is considered | |||
closed. | |||
(Closed) IRC No. 410 | |||
This item involved an 8-inch pneumatic operated butterfly-valve that was | |||
found to "fall closed" instead of "fail open" during system testing | |||
(system flush) in Unit 2. Nonconformance Report (NCR) SN-03566 was | |||
dispositioned to change couplings on the pneumatic operator. The | |||
. replacement of those couplings changed the failure position from "close" | |||
to "open." The licensee checked pneumatic operated butterfly valves of | |||
similar design to assure that correct couplings had been installed. No | |||
other valves have failed in an incorrect position. This item is | |||
considered closed. | |||
(Closed) IRC No. 411 | |||
This , item involved loose valve-shaft-to-actuator keys on motor operated | |||
valves (M0Vs). During performance of preventive maintenance on MOVs in | |||
the reactor containment building, the electrical maintenance division | |||
found that keys used to key the motor actuator to the valve stem on | |||
butterfly valves were loose or out of place on some MOVs. The licensee | |||
identified the manufacturer of these 12- and 16-inch diameter butterfly | |||
- | |||
valves. The licensee identified a total of 42 M0Vs to be inspected for | |||
icose or missing keys in the keyways. The licensee replaced the loose - | |||
keys with "snug tight" keys in the respective keyways. The licensee also | |||
. | |||
followed up on an NRC issued Information Notice (IEN) 85-67, | |||
"Valve-Shaft-to-Actuator Key May Fall Out of Place When Mounted Below | |||
' Horizontal Axis" and an NRC Circular (IEC), "Valve-Shaft-to-Actuator Key | |||
May Fall Out of Place When Mounted Below Horizontal Axis." This item is | |||
considered closed. | |||
No violations or deviations were identified. | |||
5. 10 CFR Part 21 Reporting (36100) | |||
An inspection and review of site documents was initiated by the NRC | |||
inspector to verify that the requirements included in 10 CFR Part 21 were | |||
being adhered to and that licensee documentation and implementation of the | |||
10 CFR Part 21 process functioned as required by NRC regulations and | |||
licensee commitments. The licensee controls were adequate to assure that | |||
the reporting, disposition, evaluation, and records management met NRC | |||
requirements and licensee commitments. | |||
The NRC inspector randomly selected 12 licensee 10 CFR Part 21 packages | |||
for review. The packages selected are listed below. (NOTE: Those 10 CFR | |||
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Part 21 packages identified with an "*" befora the 10 LFR Part 21 | |||
identification number had previously been reviewed by an NRC inspector.) | |||
ID Number Title * | |||
*8604642 Yokogawa AB 40 Voltmeters and Ammeters | |||
8400551 Circuit Breaker Failures During IEEE-323 Testin,g , | |||
(IEN 83-72) | |||
P21-87-28 Improper Seating of Agastat GP Series Relays | |||
, | |||
*P21-87-29 Inadequate Instructions to-Maintain Torque Switch | |||
Balance (IEB'85-03) | |||
*P21-87-31 Haughto-#620 Lubricant Attacks and Degrades Aluminum | |||
in Valves | |||
1 | |||
P21-87-51 Erratic Behavior of Static "0" Ring Differential , | |||
Pressure Switches" (IEN 86-47) | |||
. | |||
*P21-87-53 Failure o'f Neodyn Pressure Switches Used in Valve | |||
Operators for PORVs (main steam power operated relief | |||
- valve actuator-hydraulic pressure switches) | |||
P21-86-02 Pipe Support Tolerance and Installation Procedures | |||
Improper Electrical Manhole Duct Seal Design | |||
P2}-87-o0 | |||
P21-86-03 Defective Emergency Head Lever Supplied for Auxiliary | |||
Feed Pump | |||
P2!-87-16 Damaged Insulation on Valve Operator DC Motor Caused | |||
Motor Failure (IEN 87-08) | |||
* | |||
P21-87-19 Design Defect in Valve Operators Manufactured Prior | |||
to 1975 | |||
No violations or deviations were identified. | |||
6. Reactor Vessel and Internals Work Observation - Unit 2 (50053) | |||
a. Observations and Evaluations | |||
The NRC inspectors performed direct ot,servations and independent | |||
evaluations of the licensee's work performance, work in progress, and | |||
completed work. These observations determined that activities | |||
related to the reactor vessel, internals, and reactor vessel head had | |||
, been accomplished in accordance with NRC requirements and facility | |||
SAR commitments. The licensee was in the process of completing | |||
electrical wire installation on the reactor vessel closure head guide | |||
_ _ _ _ . _ , _ _ _ . . _ - ___. _ _- _ _ , _ _ . . . . . _ _ _ . | |||
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tubes and related components. The wiring in progress was being | |||
.. perfonned in accordance with procedures and general construction | |||
. | |||
. practices. | |||
' | |||
b. Installed Reactor Vessel Protection | |||
The installed reactor vessel was being protected in accordance with | |||
approved pr_ocedures. A protective covering over the top of the | |||
reactor vessel assured that foreign objects and debris would not fall | |||
into the reactor vessel. Personnel access into the covered reactor | |||
vessel was controlled. No personnel entryway into the reactor vessel | |||
was provided through to the covering over the reactor vessel into the | |||
' | |||
reactor vessel. | |||
c. Installation of Reactor Vessel Closure Head Components | |||
The NRC inspectors verified that the reactor vessel closure head | |||
assembly disconnect devices and the instnament port columns were | |||
installed in accordance with approved procedures and the following | |||
drawings: | |||
* CE E 12173-101-003, "Closure Head Assembly," Revision 3, dated | |||
December 1973. | |||
Westinghouse 6123E44, "Instr Port Column, Sef., Loading and | |||
Hydrostatic Test Assembly," Revision 3, dated March 1, 1978. | |||
- Westinghouse 1455E38 "(TG) South Texas Report No.1 Reactor | |||
General Assy." Sheet 2 of 3, Revision 0, dated September 27, | |||
h 1978. (NOTE: Drawing for Unit 1 is used for Unit 2.) | |||
Westinghouse 1209E92, "Interface Feature Critical 4 XLR Reactor | |||
Internals," Sheet 1 of 6, Revision 3, dated February 13, 1978. | |||
d. Weld Inspection of Reactor Vent Piping | |||
The NRC inspector performed an inspection of the 1-inch piping from | |||
the near center of the reactor vessel head including two manual | |||
relief gate valves and the balance of piping, including the four | |||
Target Rock solenoid operated relief valves. The Target Rock valves | |||
were identified by their respective serial numbers: 103, 105, 108, | |||
and 110. The inspection determined that the weld quality and the | |||
installed piping and valve configurations were in accordance with | |||
design and construction requirements. No indications of damage or | |||
inadequate installation were identified. Visual inspections of | |||
weldin'; and general installation indicated that the welding, | |||
components, and reactor vessel; closure head were in satisfactory | |||
condition. | |||
. | |||
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e. Review of Data Packages | |||
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s The NRC inspectors reviewed the data packages for the two manually | |||
operated globe valves and for each of the four solenoid operated | |||
globe valves. The solenoid valves were manufactured by Target Rock. | |||
These six valves and the associated piping were designed to meet the | |||
prinary system requirements, including design operating pressure of | |||
2485 psig and 650 F. The results of the review of the-data packages | |||
for the four Target Rock manufactured solenoid valves.are as follows: | |||
Component Material Spec. No. Material | |||
Body ASME SA 182 SS316L | |||
Bonnet ASME CA 479 SS316 | |||
Disc ASME SA 564 SS17-4ph | |||
Indicator ASME SA 479 SS316 | |||
e tube | |||
The data packages for these'six valves were complete and contained | |||
, | |||
the information required by ASME, Westinghouse, and the licensee to | |||
fully document the material, testing, welding., and inspections | |||
performed on each of these six valves. The data packages also met | |||
the requirements stated in Regulatory Guide (RG) 1.88, "Collection, | |||
Storage and Maintenance of Nuclear Power Plant Quality Assurance | |||
Records." | |||
f. Procedure and SAR Review | |||
' | |||
The NRC inspectors determined that the inspection, installation, and | |||
wiring in progress were in accordance with licensee approved | |||
procedures and the SAR (Chapters 1, 3, 4, 5, and 17, including | |||
appropriate codes and standards referenced in these SAR Chapters). | |||
g. _ Containment Vessel Housekeeping | |||
The licensee was adhering to the recommendations stated in RG 1.39, | |||
"Housekeeping Requirements for Water-Cooled Nuclear Power Plants" in | |||
the general area of the reactor vessel internals and reactor vessel | |||
head laydown areas. | |||
No violations or deviations were identified. | |||
7. Reactor Vessel and Internals QA Review - Unit 2 (50051) | |||
: | |||
The NRC inspectors determined that the licensee's onsite QA | |||
responsibilities relative to the reactor vessel and internals installation | |||
had been established and implemented. The licensee had' established an | |||
audit program (including plans, procedures, and schedule) covering | |||
safety-related work and control functions related to the reactor vessel, | |||
reactor vessel closure head, and internals. The licensee and each | |||
contractor had an established program to assure that craft, eramination, | |||
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and inspection personnel associated with the reactor vessel, reactor | |||
vessel closure head, and internals had been trained and qualified to | |||
perform the designated tasks. The licensee had developed appropriate and, '- | |||
adequate procedures. These procedures assured that activities associated | |||
with the reactor vessel, reactor vessel closure head, and internals were | |||
" | |||
controlled and performed in accordance with NRC requirements and SAR | |||
commitments-(SAR Chapters 1, 3, 4, 5, and 17 and the codes and standirds | |||
referenced in these SAR Chap;ers). | |||
These procedures included precautions to preclude damage or mishandling of | |||
equipment and components. The installation of the reactor vessel | |||
internals and the reactor vessel closure head were properly addressed. | |||
No violations or deviations,were identified. | |||
8. Standby Diesel Generator Expansion Seals (Bellows) Replacement - Unit 2 | |||
(50073) | |||
During the licensee's prerequisite testing (initial startup) of the "A" | |||
SDG, leaks were discovered in cylinder liner expansion seals (bellows) on | |||
cylinders 10R, SR, 2L, 3L, and 4L. Each of the three Unit 2 SDGs is a | |||
V-type, 20 cylinder, Model KSV, manufactured by Cooper-Bessemer. | |||
These bellows are designed to allow the cylinder liners to expand and | |||
contract. The bellows are a cooling water jacket pressure boundary | |||
secondary seal between the cylinder liners and the cylinder block. The | |||
secondary seal prevents jacket water cooling system fluid from entering | |||
. the diesel lube oil sump. The primary seal is the metal to metal contact | |||
l of the liner to the engine block. | |||
L The NRC inspectors performed 19 separate inspections during the expansion | |||
seals (bellows) removal and replacement activities on the "A" SDG. The | |||
NRC inspectors visually inspected all 20 of the bellows that were removed | |||
for replacement by contractnr personnel. Visual inspection by the NRC | |||
l | |||
inspectors verified that the bellows removed from cylinder 10R- had a | |||
l through wall hole about 1/4 inch in diameter and that the bellows removed | |||
I from cylinders SR, 2L, 3L, and 4L had through wall pin holes. Replacement | |||
l of all 20 bellows on "A" SDG and related work activities was completed by | |||
the contractor on April 22, 1988. | |||
, | |||
The licensee shipped the two bellows that were removed from cylinders 10R i | |||
! and 10L to the Bechtel Material Laboratory for analysis. The analysis | |||
reported that both bellows exhibited evidence of Microbiological Induced | |||
' | |||
Corrosion (MIC) and trar,sgranular stress corrosion cracking. | |||
, | |||
The licensee prepared a detailed report for submittal to the NRC. The | |||
licensee's report was submitted to the NRC on May 11, 1988. This report | |||
discussed the discovery of MIC in the jacket water cooling systems, piping | |||
l replaced, bellows replacement, an evaluation of storage history, Unit 1 | |||
l | |||
t | |||
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- | |||
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storage history and inspection activities, and overall conclusions of | |||
, | |||
Units 1 and 2 SDG inspection, repair, replacement, and analysis | |||
activities. | |||
No violations or deviations were identified. . | |||
9. Safety-Related Piping Observations - Unit 2 (49063 and 49065) | |||
' | |||
The NRC inspector performed an inspection of safety related piping to | |||
assess the licensee's program for welding off-site fabricated spool pieces | |||
into completed systems and to determine the adequacy of inspection to the | |||
specifications, | |||
a. Work Observation , | |||
The NRC inspector selected the main steam and feedwater systems for | |||
the "B" train in the "B" bay of the Isolation Valve Cubicles (IVC). | |||
The following portions of the completed systems were inspected: | |||
' | |||
Main steam line from the north wall to containment penetration | |||
M-3 which included the following: | |||
' | |||
, | |||
(1) Main steam isolation valve - FSV 7424 | |||
, | |||
(2) Main steam isolation valve bypass - FV 7422 | |||
(3) Five safety valves (PSV 7420 through PSV 7420D) | |||
(4) Manual steam dump valve - MS-0038 , | |||
, | |||
(5) Remote operated steam dump valve - PV-7421 | |||
(6) Five main steam line hangers (Numbers HL-5010 through | |||
HL-5014) | |||
(7) Main steam manifeld and balance of piping - Feedwater line | |||
f rom the north wall to the containment penetration | |||
including: | |||
a) Feedwater Check Valve - FW-0C65 | |||
b) Feedwater bypass and Valve - FV-7147 | |||
c) Feedwater Control Valve - FV-7142 | |||
d) Balance of uninsulated piping | |||
The following isometric drawings were reviewed and compared to | |||
installed components during the inspection of the above: | |||
2G369 PMS 646, Sheet 7, Revision 6 | |||
' | |||
SG369 PMS 646, Sheet 2, Revision 7 | |||
* 2G3818 FW 1030, AA2, Sheet 1, Revision 2 . | |||
2G3618 FW 1053, AA2, Sheet 1, Revision 2 | |||
< | |||
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Five main steam line hangers (Numbers HL-5010 through HL-5014) were | |||
; found to oe installed in accordance with the following drawings: | |||
MS-9002-HL5010, Sheet 1, Revision 7 | |||
* | |||
MS-9002-HL5011, Sheet 1, Revision 6 | |||
MS-9002-HL5012, Sheet 1, Revision.6 | |||
MS-9002-HL5013, Sheet 1, Revision 4 | |||
MS-9002-HL5014, Sheet 1, Revision 2 | |||
' | |||
The NRC inspectors visually examined the pi)ing systems for correct | |||
length, size, and configuration. The pipe langers were inspected for | |||
, location, configuration, welding, and hanger type. The valves were " | |||
examined for data plate information, identification, visual damage, | |||
installation assured flow in the proper direction, and general | |||
condition. | |||
' | |||
The NRC inspectors determined that the two piping systems inspected | |||
had been installed in accordance with the procedures and drawings, | |||
b. Record Review | |||
The NRC inspectors reviewed the code data packages for the spool | |||
pieces and valves in the safety-related portions of two systems (main | |||
steam and feedwater) inspected. The following records were found to | |||
be properly stored, retrievable, and representative of this | |||
particular segment of the installation. | |||
Valves | |||
30 inch Atwood and Morrill - Main steam isolation valve - FSV-7424 | |||
4 inch WMK - Main steam isolation bypass - FV-7422 | |||
6 inch Dresser - Main steam safety - BT02199 - PSV-7420 C | |||
8 inch Anchor Darling - Main steam dump - MS-0038 ' | |||
8 inch WMK - Main steam dump - PV-7421 | |||
18 inch Anchor Darling - Feedwater control - FV-7142 | |||
18 inch Anchor Darling - Feedwater check valve - FW-0066 | |||
2 inch Valtek - Feedwater bypass - FV-7147A | |||
Piping | |||
. | |||
30 inch MS-1002-GAZ | |||
3 inch FW-1053-AAZ | |||
No violations or deviations were identified during the inspection of | |||
the piping systems. However, the inspection was expanded to include | |||
construction activities in the "B" IVC, and the results of the | |||
inspection are documented in paragraph 10 below. | |||
10. Housekeeping - Unit 2 (50073 and 49063) | |||
During an inpsection of the ICV's, the NRC inspectors observed excessive | |||
amounts of construction debris, abandoned tools, and various supplies | |||
_ | |||
- _ - - -. .- _ - - , | |||
' | |||
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' | |||
. | |||
4 | |||
- 13 ' | |||
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scattered about. Also, the inspector noted dirt in cable trays, under | |||
grates and on beams and flanges. Tnis is an apparent violation of the ' | |||
licer3ee's Construction Site Procedure (CSP) .12. "General Instruction for | |||
Houseneeping During Construction," RG 1.39, and American National | |||
Standards Institute ' ANSI) N45.2.3. The licensee is committed to RG 1.39 | |||
and ANSI N45.2.3. The NRC inspectors determined that the lack of | |||
appropriate housekeeping ac | |||
- work being performed in an ,tivities in | |||
inefficient this area | |||
manner that could | |||
could result | |||
producein low | |||
routine | |||
quality results that would not be in conformance with NRC requirements and | |||
licensee commitments. This unsatisfactory level of housekeeping was | |||
evident.,in the IVC area only. The licensee's failure to maintain minimum | |||
housekeeping) standards | |||
(499/8824-03 of the in the IVC area | |||
above requirements is an apparent violation | |||
and commitments. | |||
11. Exit Interview (30703) , | |||
The NRC inspector met with licensee representatives (denoted in | |||
'',. | |||
paragraph 1) on April 22, 1988, and summarized the scope and findings of | |||
the inspection. Other meetings between NRC inspectors and licensee | |||
management were held periodically during the inspection to discuss | |||
identified concerns. The licensee did not identify as proprietary any of | |||
the information provided to or reviewed by the inspectors during this | |||
inspection. | |||
, | |||
& | |||
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9 | |||
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}} |
Latest revision as of 10:04, 21 December 2021
ML20155B973 | |
Person / Time | |
---|---|
Site: | South Texas |
Issue date: | 06/03/1988 |
From: | Constable G, Garrison D, Hunnicutt D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
To: | |
Shared Package | |
ML20155B942 | List: |
References | |
50-499-88-24, NUDOCS 8806140038 | |
Download: ML20155B973 (13) | |
See also: IR 05000499/1988024
Text
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APPENDIX B,
,
. ;, . U.S. NUCLEAR REGULATORY COMMISSION- .
> ' REGION IV-
NRC Inspection. Report: 50-499/88-24 Construction Permit: CPPR-129-
-
Docket: 50-499 CP Expiration Date: December 1989 ,
Licensee: Houston Lightir.g & Power Company (HL&P) '
P.O. Box 1700
Houston, Texas 77001
"
Facility Name: South Texas Project, Unit 2 (STP)
Inspection 'At: STP, Ma,tagorda County, Texas
Inspection Conducted: April 5 <throuch May 2,1988
,
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Inspectors: / h www
D. L. Garrison, Resident Inspector, Reactor
~
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Date /
70
Project Section D, Division;of Reactor
Projects
hY/ Yu-> af
D. M. Hunnicott, Senior Reactor Inspector
.d3b?
Date /
Reactor Project Section D, Division of
Reactor Projects
Approved: h orAnc.'
G. L. Constable, Chief, Reactor Project
d/3/87
Date '
Section D, Division of Reactor Projects
8806140038 880607
{DR ADOCK 05000499
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Inspection Summary
.
Inspection Conducted April'5 through May 2,1988 (Report 50'-499/88-24)
Areas Inspected: Routine, unannounced inspection including tagging. status of
'
equipment, cable protection, followup on previous inspection findings, followup 4
on licensee reported significant construction. deficiencies, review of 10 CFR
.Part 21 reporting program, reactor vessel and internal work observation,
.t
.
reactor vessel and internal .QA review, standby diesel generator. expansion seals
replacement, and safety-related piping installation observations,-
4 s-
Results: .Within the seven areas inspected, three apparent violations were
widentified (improper tagging of equipment, paragraph 2; inadequate cab'e
protection, paragraph 3; and inadequate housekeeping, paragraph 10).
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DETAILS-
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1. Persons Contacted
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HL&P
- J. S. Phelps, Supervising Engineer, Project Compliance-
- S. M. Head, Support Licensing Engineering
- W. G. Wellborn, Supervising Project Engineer
- S. D. Phillips, P Z Engineer'
- X. M. 0'Gara, Project Compliance
- T. J. Jordan, Project Quality Assurance Manager
- G. Ondriska, Startup Engineer
- A. R. Mikus, General Supervisor
- T. Quirk, General Supervisor, Records Management System
Bechtel
- R. D. Bryan, Field Construction Manager
'
- R. H. Medina, Quality Assurance Supervisor
, Ebasco
- R. Abe'i, Quality Program Site Manager
- R. C. Sisson, Site Resident Engineer
In addition to the above, the NRC inspectors also held discussions with
various licensee, architect engineer (AE), constructor and other contractor
personnel.during this inspection.
- Denotes those individuals attending the exit interview conducted on
April 22, 1988. ,
2. Tagging Status of Equipment - Unit 2 (50071)
Numerous procedures on the site required tagging to reflect the correct
status of the items tagged. During the inspection in the "B" Isolation
Valve Cubicles (IVC), five instances were observed by the NRC inspectors
where the tagging on valves and components did not reflect the current
status of the tagged equipment. The observed incorrect tagging is listed
. below:
- One tag indicated a valve to be temporarily installed. The valve was
permanently installed. ,
- Two tags indicated parts were removed or to be removed and had been
on the equipment for three years. The required action was found to
be abandoned.
4
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,0ne. gate valve was found with only the body intact in the line. No
status or tagging was on or near the valve.
One secondary system hydrostatic test tag was observed which should
have been removed before the primary hydrostatic test was perfonned.
During the review of the procedures, the NRC inspector concluded that
clarification in some instances was needed for removal and change of
status. This problem could be a plant generic problem. The licensee's
failure to maintain current status tags on equipment (tags that were
invalid or did not correctly identify the status of components) is an
apparent violation (499/8824-01) of NRC requirements and licensee
commitments. ,
,
3. Cable protection - Unit 2 (51065 and 51063)
During the inspection of the "B" Isolation Valve Cubicle on April 20
and 21, 1988, and previous inspections of the electrical switchgear rooms
in the electrical auxiliary buildin
there were numerous (more than instances 25) g, the of
NRC inspectors
failure observed
to install cable that
softeners over the sharp edges on some of the cable trays. The sharp edges
on cable trays were observed in the cable trays where electrical cables
were routed from one cable tray to another cable tray. Licensee procedures
are clear concerning this item. The NRC inspection results indicated that
failure to install cable softeners could be a generic plant problem. The
licensee's failure to provide cable softeners, where required, to preclude
damage -to electrical cableijacket material is an apparent violation
(499/8824-02) of NRC requirements and licensee coninitments.
4. Followup on Licensee Reported Significant Construction Deficiencies
(10 CFR 50.55(e)) (36100 and 92/00)
(Closed) Incident Review Committee (IRC) No. 398
This item involved a finding that if the standby diesel generator (SDG)
was being operated in the testing mode and a loss of offsite power (LOOP)
occurred, the electrical supply breakers to the 480 VAC load centers would
be tripped. Then, since the SDG would be running at design speed, the
breakers would be signaled to close shortly thereaf ter. The signal to
close prior to full spring recharge would cause the antipump feature on
the electrical supply breakers to the 480 VAC loads to lock out the
breakers. The breaker lockout would prevent energizing of all 480 VAC
loads connected to the affected train.
Configuration control package (CCP) 2-E-ST-833, Revision 0, dated
August 20, 1987, was issued to add an Agastat relay (6-second setting) in
the LOOP control circuit f rom the sequencer to ensure that the DG breaker
does not close until the 480 VAC circuit breaker closing springs are
,
cha rged. Site electrical discipline personnel had been reinstructed in
the proper design methods assnciated with conditions to prevent
recurrence. This item is considered closed.
.
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(Closed) IRC No. 400 ,
This item involved concerns related to inadequate cooling of the Unit 2
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high voltage cubicle panels for the SDG. HL&P completed and installed a
design, modification to provide additional cooling for the Unit ~2 SDG "
panels. A 100-hour test run of the SDG was completed successfully. No
recurrence of overheating conditions were observed. This item is-
considered closed.
(Closed) IRC No. 402
This item involved a finding that some motor shaft-to-pinion gear keys
sheared (failed) in Limitorque Model SMB-0-25 operators. The shearing of
the keys was apparently due to'the keys being machined from incorrect or
defective material. Twelve motor shaft-to-pinion gear keys in valve ,
operators in Unit 2 (Deficiency Evaluation Report 87-046) were replaced
with keys manufactured from AISI 1018 steel. According to the material
specifications, AISI 1018 steel is the correct material. This item is
.
considered closed.
(Closed) IRC No. 403
This item involved Class 1E electrical cable splices which utilize Raychem
^ insulation material and which had been identified as nonconforming after
final QA-acceptance. Three different problem areas were identified by the
licensee. The licensee reviewed the completed cables and terminations; '
reviewed a computer resort of cables and terminations to validate previous
data; reviewed data on electrical penetration termination cards to assure
that electrical splices to electrical penetrations had been reinspected;
reviewed the transfer of data from master splice lists to the reinspection
data lists to identify any omissions; and reviewed the qualifications and
. work performed by off-project Bechtel personnel to assure that the
methodology and reviews had been performed correctly. This item is
considered closed.
4
(Closed) IRC No. 408
This item involved the failure of tubes in the component cooling
water (CCW)heatexchangers. The tube failures resulted from shell side
flow induced vibrations. Design modifications to the heat exchangers were
o required to prevent further damage. In each CCW, 30 tubes were removed in
the areas above and below the impingment plate. Two hundred sixty-four
tubes were rodded with steel rods inserted to dampen vibration ar.d to
decrease the vibration level in the adjacent inboard tubes. The vacated
tube sheet holes and the rodded tubes were plugged with plugs manufactured
.
from aluminum bronze material (similar to and compatible with the
tubesheet cladding and essential cooling water (ECW) piping materials).
1he ECW flow (design flow rate was 15,000 gpm) and pressure drop through
the heat exchanger tubes have been recalculated for the modified condition
of the heat exchangers. The seismic qualification and the weight increase
due to the modification of the heat exchangers have been reevaluated.
4
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Other safety-related heat exchangers have been evaluated to determine if
any of these heat exchangers would be susceptible to the problems observed
in the CCW heat exchangers. None of the heat exchangers evaluated were
'
found to be susceptible to these identified problems. The changes that
have been made to the CCW heat exchangers to resolve the vibration induced
problems do not alter previous FSAR commitments. This item is considered
closed.
(Closed) IRC No. 410
This item involved an 8-inch pneumatic operated butterfly-valve that was
found to "fall closed" instead of "fail open" during system testing
(system flush) in Unit 2. Nonconformance Report (NCR) SN-03566 was
dispositioned to change couplings on the pneumatic operator. The
. replacement of those couplings changed the failure position from "close"
to "open." The licensee checked pneumatic operated butterfly valves of
similar design to assure that correct couplings had been installed. No
other valves have failed in an incorrect position. This item is
considered closed.
(Closed) IRC No. 411
This , item involved loose valve-shaft-to-actuator keys on motor operated
valves (M0Vs). During performance of preventive maintenance on MOVs in
the reactor containment building, the electrical maintenance division
found that keys used to key the motor actuator to the valve stem on
butterfly valves were loose or out of place on some MOVs. The licensee
identified the manufacturer of these 12- and 16-inch diameter butterfly
-
valves. The licensee identified a total of 42 M0Vs to be inspected for
icose or missing keys in the keyways. The licensee replaced the loose -
keys with "snug tight" keys in the respective keyways. The licensee also
.
followed up on an NRC issued Information Notice (IEN) 85-67,
"Valve-Shaft-to-Actuator Key May Fall Out of Place When Mounted Below
' Horizontal Axis" and an NRC Circular (IEC), "Valve-Shaft-to-Actuator Key
May Fall Out of Place When Mounted Below Horizontal Axis." This item is
considered closed.
No violations or deviations were identified.
5. 10 CFR Part 21 Reporting (36100)
An inspection and review of site documents was initiated by the NRC
inspector to verify that the requirements included in 10 CFR Part 21 were
being adhered to and that licensee documentation and implementation of the
10 CFR Part 21 process functioned as required by NRC regulations and
licensee commitments. The licensee controls were adequate to assure that
the reporting, disposition, evaluation, and records management met NRC
requirements and licensee commitments.
The NRC inspector randomly selected 12 licensee 10 CFR Part 21 packages
for review. The packages selected are listed below. (NOTE: Those 10 CFR
,
. . _ - _ __ . - _ - _ ._. . _ _ _ _
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Part 21 packages identified with an "*" befora the 10 LFR Part 21
identification number had previously been reviewed by an NRC inspector.)
ID Number Title *
- 8604642 Yokogawa AB 40 Voltmeters and Ammeters
8400551 Circuit Breaker Failures During IEEE-323 Testin,g ,
(IEN 83-72)
P21-87-28 Improper Seating of Agastat GP Series Relays
,
- P21-87-29 Inadequate Instructions to-Maintain Torque Switch
Balance (IEB'85-03)
- P21-87-31 Haughto-#620 Lubricant Attacks and Degrades Aluminum
in Valves
1
P21-87-51 Erratic Behavior of Static "0" Ring Differential ,
Pressure Switches" (IEN 86-47)
.
- P21-87-53 Failure o'f Neodyn Pressure Switches Used in Valve
Operators for PORVs (main steam power operated relief
- valve actuator-hydraulic pressure switches)
P21-86-02 Pipe Support Tolerance and Installation Procedures
Improper Electrical Manhole Duct Seal Design
P2}-87-o0
P21-86-03 Defective Emergency Head Lever Supplied for Auxiliary
Feed Pump
P2!-87-16 Damaged Insulation on Valve Operator DC Motor Caused
Motor Failure (IEN 87-08)
P21-87-19 Design Defect in Valve Operators Manufactured Prior
to 1975
No violations or deviations were identified.
6. Reactor Vessel and Internals Work Observation - Unit 2 (50053)
a. Observations and Evaluations
The NRC inspectors performed direct ot,servations and independent
evaluations of the licensee's work performance, work in progress, and
completed work. These observations determined that activities
related to the reactor vessel, internals, and reactor vessel head had
, been accomplished in accordance with NRC requirements and facility
SAR commitments. The licensee was in the process of completing
electrical wire installation on the reactor vessel closure head guide
_ _ _ _ . _ , _ _ _ . . _ - ___. _ _- _ _ , _ _ . . . . . _ _ _ .
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tubes and related components. The wiring in progress was being
.. perfonned in accordance with procedures and general construction
.
. practices.
'
b. Installed Reactor Vessel Protection
The installed reactor vessel was being protected in accordance with
approved pr_ocedures. A protective covering over the top of the
reactor vessel assured that foreign objects and debris would not fall
into the reactor vessel. Personnel access into the covered reactor
vessel was controlled. No personnel entryway into the reactor vessel
was provided through to the covering over the reactor vessel into the
'
reactor vessel.
c. Installation of Reactor Vessel Closure Head Components
The NRC inspectors verified that the reactor vessel closure head
assembly disconnect devices and the instnament port columns were
installed in accordance with approved procedures and the following
drawings:
- CE E 12173-101-003, "Closure Head Assembly," Revision 3, dated
December 1973.
Westinghouse 6123E44, "Instr Port Column, Sef., Loading and
Hydrostatic Test Assembly," Revision 3, dated March 1, 1978.
- Westinghouse 1455E38 "(TG) South Texas Report No.1 Reactor
General Assy." Sheet 2 of 3, Revision 0, dated September 27,
h 1978. (NOTE: Drawing for Unit 1 is used for Unit 2.)
Westinghouse 1209E92, "Interface Feature Critical 4 XLR Reactor
Internals," Sheet 1 of 6, Revision 3, dated February 13, 1978.
d. Weld Inspection of Reactor Vent Piping
The NRC inspector performed an inspection of the 1-inch piping from
the near center of the reactor vessel head including two manual
relief gate valves and the balance of piping, including the four
Target Rock solenoid operated relief valves. The Target Rock valves
were identified by their respective serial numbers: 103, 105, 108,
and 110. The inspection determined that the weld quality and the
installed piping and valve configurations were in accordance with
design and construction requirements. No indications of damage or
inadequate installation were identified. Visual inspections of
weldin'; and general installation indicated that the welding,
components, and reactor vessel; closure head were in satisfactory
condition.
.
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e. Review of Data Packages
'
s The NRC inspectors reviewed the data packages for the two manually
operated globe valves and for each of the four solenoid operated
globe valves. The solenoid valves were manufactured by Target Rock.
These six valves and the associated piping were designed to meet the
prinary system requirements, including design operating pressure of
2485 psig and 650 F. The results of the review of the-data packages
for the four Target Rock manufactured solenoid valves.are as follows:
Component Material Spec. No. Material
e tube
The data packages for these'six valves were complete and contained
,
the information required by ASME, Westinghouse, and the licensee to
fully document the material, testing, welding., and inspections
performed on each of these six valves. The data packages also met
the requirements stated in Regulatory Guide (RG) 1.88, "Collection,
Storage and Maintenance of Nuclear Power Plant Quality Assurance
Records."
f. Procedure and SAR Review
'
The NRC inspectors determined that the inspection, installation, and
wiring in progress were in accordance with licensee approved
procedures and the SAR (Chapters 1, 3, 4, 5, and 17, including
appropriate codes and standards referenced in these SAR Chapters).
g. _ Containment Vessel Housekeeping
The licensee was adhering to the recommendations stated in RG 1.39,
"Housekeeping Requirements for Water-Cooled Nuclear Power Plants" in
the general area of the reactor vessel internals and reactor vessel
head laydown areas.
No violations or deviations were identified.
7. Reactor Vessel and Internals QA Review - Unit 2 (50051)
The NRC inspectors determined that the licensee's onsite QA
responsibilities relative to the reactor vessel and internals installation
had been established and implemented. The licensee had' established an
audit program (including plans, procedures, and schedule) covering
safety-related work and control functions related to the reactor vessel,
reactor vessel closure head, and internals. The licensee and each
contractor had an established program to assure that craft, eramination,
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and inspection personnel associated with the reactor vessel, reactor
vessel closure head, and internals had been trained and qualified to
perform the designated tasks. The licensee had developed appropriate and, '-
adequate procedures. These procedures assured that activities associated
with the reactor vessel, reactor vessel closure head, and internals were
"
controlled and performed in accordance with NRC requirements and SAR
commitments-(SAR Chapters 1, 3, 4, 5, and 17 and the codes and standirds
referenced in these SAR Chap;ers).
These procedures included precautions to preclude damage or mishandling of
equipment and components. The installation of the reactor vessel
internals and the reactor vessel closure head were properly addressed.
No violations or deviations,were identified.
8. Standby Diesel Generator Expansion Seals (Bellows) Replacement - Unit 2
(50073)
During the licensee's prerequisite testing (initial startup) of the "A"
SDG, leaks were discovered in cylinder liner expansion seals (bellows) on
cylinders 10R, SR, 2L, 3L, and 4L. Each of the three Unit 2 SDGs is a
V-type, 20 cylinder, Model KSV, manufactured by Cooper-Bessemer.
These bellows are designed to allow the cylinder liners to expand and
contract. The bellows are a cooling water jacket pressure boundary
secondary seal between the cylinder liners and the cylinder block. The
secondary seal prevents jacket water cooling system fluid from entering
. the diesel lube oil sump. The primary seal is the metal to metal contact
l of the liner to the engine block.
L The NRC inspectors performed 19 separate inspections during the expansion
seals (bellows) removal and replacement activities on the "A" SDG. The
NRC inspectors visually inspected all 20 of the bellows that were removed
for replacement by contractnr personnel. Visual inspection by the NRC
l
inspectors verified that the bellows removed from cylinder 10R- had a
l through wall hole about 1/4 inch in diameter and that the bellows removed
I from cylinders SR, 2L, 3L, and 4L had through wall pin holes. Replacement
l of all 20 bellows on "A" SDG and related work activities was completed by
the contractor on April 22, 1988.
,
The licensee shipped the two bellows that were removed from cylinders 10R i
! and 10L to the Bechtel Material Laboratory for analysis. The analysis
reported that both bellows exhibited evidence of Microbiological Induced
'
Corrosion (MIC) and trar,sgranular stress corrosion cracking.
,
The licensee prepared a detailed report for submittal to the NRC. The
licensee's report was submitted to the NRC on May 11, 1988. This report
discussed the discovery of MIC in the jacket water cooling systems, piping
l replaced, bellows replacement, an evaluation of storage history, Unit 1
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storage history and inspection activities, and overall conclusions of
,
Units 1 and 2 SDG inspection, repair, replacement, and analysis
activities.
No violations or deviations were identified. .
9. Safety-Related Piping Observations - Unit 2 (49063 and 49065)
'
The NRC inspector performed an inspection of safety related piping to
assess the licensee's program for welding off-site fabricated spool pieces
into completed systems and to determine the adequacy of inspection to the
specifications,
a. Work Observation ,
The NRC inspector selected the main steam and feedwater systems for
the "B" train in the "B" bay of the Isolation Valve Cubicles (IVC).
The following portions of the completed systems were inspected:
'
Main steam line from the north wall to containment penetration
M-3 which included the following:
'
,
(1) Main steam isolation valve - FSV 7424
,
(2) Main steam isolation valve bypass - FV 7422
(3) Five safety valves (PSV 7420 through PSV 7420D)
(4) Manual steam dump valve - MS-0038 ,
,
(5) Remote operated steam dump valve - PV-7421
(6) Five main steam line hangers (Numbers HL-5010 through
HL-5014)
(7) Main steam manifeld and balance of piping - Feedwater line
f rom the north wall to the containment penetration
including:
a) Feedwater Check Valve - FW-0C65
b) Feedwater bypass and Valve - FV-7147
c) Feedwater Control Valve - FV-7142
d) Balance of uninsulated piping
The following isometric drawings were reviewed and compared to
installed components during the inspection of the above:
2G369 PMS 646, Sheet 7, Revision 6
'
SG369 PMS 646, Sheet 2, Revision 7
- 2G3818 FW 1030, AA2, Sheet 1, Revision 2 .
2G3618 FW 1053, AA2, Sheet 1, Revision 2
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Five main steam line hangers (Numbers HL-5010 through HL-5014) were
- found to oe installed in accordance with the following drawings
MS-9002-HL5010, Sheet 1, Revision 7
MS-9002-HL5011, Sheet 1, Revision 6
MS-9002-HL5012, Sheet 1, Revision.6
MS-9002-HL5013, Sheet 1, Revision 4
MS-9002-HL5014, Sheet 1, Revision 2
'
The NRC inspectors visually examined the pi)ing systems for correct
length, size, and configuration. The pipe langers were inspected for
, location, configuration, welding, and hanger type. The valves were "
examined for data plate information, identification, visual damage,
installation assured flow in the proper direction, and general
condition.
'
The NRC inspectors determined that the two piping systems inspected
had been installed in accordance with the procedures and drawings,
b. Record Review
The NRC inspectors reviewed the code data packages for the spool
pieces and valves in the safety-related portions of two systems (main
steam and feedwater) inspected. The following records were found to
be properly stored, retrievable, and representative of this
particular segment of the installation.
Valves
30 inch Atwood and Morrill - Main steam isolation valve - FSV-7424
4 inch WMK - Main steam isolation bypass - FV-7422
6 inch Dresser - Main steam safety - BT02199 - PSV-7420 C
8 inch Anchor Darling - Main steam dump - MS-0038 '
8 inch WMK - Main steam dump - PV-7421
18 inch Anchor Darling - Feedwater control - FV-7142
18 inch Anchor Darling - Feedwater check valve - FW-0066
2 inch Valtek - Feedwater bypass - FV-7147A
Piping
.
30 inch MS-1002-GAZ
3 inch FW-1053-AAZ
No violations or deviations were identified during the inspection of
the piping systems. However, the inspection was expanded to include
construction activities in the "B" IVC, and the results of the
inspection are documented in paragraph 10 below.
10. Housekeeping - Unit 2 (50073 and 49063)
During an inpsection of the ICV's, the NRC inspectors observed excessive
amounts of construction debris, abandoned tools, and various supplies
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scattered about. Also, the inspector noted dirt in cable trays, under
grates and on beams and flanges. Tnis is an apparent violation of the '
licer3ee's Construction Site Procedure (CSP) .12. "General Instruction for
Houseneeping During Construction," RG 1.39, and American National
Standards Institute ' ANSI) N45.2.3. The licensee is committed to RG 1.39
and ANSI N45.2.3. The NRC inspectors determined that the lack of
appropriate housekeeping ac
- work being performed in an ,tivities in
inefficient this area
manner that could
could result
producein low
routine
quality results that would not be in conformance with NRC requirements and
licensee commitments. This unsatisfactory level of housekeeping was
evident.,in the IVC area only. The licensee's failure to maintain minimum
housekeeping) standards
(499/8824-03 of the in the IVC area
above requirements is an apparent violation
and commitments.
11. Exit Interview (30703) ,
The NRC inspector met with licensee representatives (denoted in
,.
paragraph 1) on April 22, 1988, and summarized the scope and findings of
the inspection. Other meetings between NRC inspectors and licensee
management were held periodically during the inspection to discuss
identified concerns. The licensee did not identify as proprietary any of
the information provided to or reviewed by the inspectors during this
inspection.
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