IR 05000219/1987042
| ML20148B682 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 03/14/1988 |
| From: | Cowgill C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20148B678 | List: |
| References | |
| 50-219-87-42, NUDOCS 8803220086 | |
| Download: ML20148B682 (12) | |
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' S. REGULATORY COMMISSION
REGION I
1 Report N /87-42 Docket N License N OpR-16 Priority -- Category C Licensee: .GPU Nuclear Corporation 1 Upper pond Road Parsippany, New Jersey 07054 Facility Name: Oyster Creek Nuclear Generating Station Inspection Conducted: December 20, 1987 - February 6, 1988 Participating Inspectors: J. Wechselberger, Senior Resident Inspector
. Collins,. Resident Inspector Approved By: "dE hM Date C.Cowgill, Chief (}pactorProjectsSection1A Inspection Summary:
Areas Inspected: Routine inspections were conducted by the resident inspectors and one region-based inspector (151 hours0.00175 days <br />0.0419 hours <br />2.496693e-4 weeks <br />5.74555e-5 months <br />) of activities in progress including opera-tions, radiation control, physical security, surveillance and' maintenanc The inspectors also periodically toured the control room and other portions of the plant and reviewed periodic and special report In addition the inspectors par-ticipated in selection of fasteners for Bulletin 87-02, reviewed licensee activi-ties surrounding the freezing of instrument lines and examined feedwater flow nozzle and core thermal power calculations. The inspection period included review of operational considerations including high pressure scram surveillances (RE03),
"C" main battery, "C" EMRV acoustic monitor and safety valve thermocouples, as well as the review of two other industry problems for applicability to Oyster Creek; specifically a feedwater regulation valve failure and diesel engine exhaust mani-fold combustion. Additionally, the inspectors reviewed inspection activities associated with snubber visual inspection surveillanc Results: No Violations were identified. The high pressure scram instruments (RE03)
continue to exhibit setpoint repeatability problems. A special inspection (50- ;
219/88-02) was conducted to review the instrument line freezing event. The licen-see plans to conduct additional noisa signature recording of "C" EMRV acoustic monitor in an attempt to determine if the signaturo is representative of a steam leak or flow noise. The review of the industry diesel problem determined that this particular problem was very unlikely to occur at the facilit The licensee de-termined that the feedwater valve cyclic fatigue problem had never occurred at the facility, but elected to include this as an inspection attribute during the next outag PDR ADOCK 05000219DCD O
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L DETAILS 1.0 . Freezing Instrument Lines in Reactor Building On 1/6/88, #2 seal pressure indicator for the "B" and."E" recirculation pumps -
, and the shutdown ~ cooling pumps ' discharge pressure instrument lines showed indications of freezing. The seal pressure increased from approximately 520 psig (normal pressure indication) to about 800 psig. The operators initially suspected a seal failure, but later determined that the instrument lines'were freezing as a result of cold reactor-building temperatures. The operators measured the temperatures on the 23' elevation'in the area of-the freezing instrument lines at approximately 33 F and as.a result the operators took action to reduce the cold air induction into the reactor building by securing normal reactor building ventilation and~ initiating the standby gas treatment sy' stem (SBGTS). As a result within approximately 30 minutes the recirculation seal pressures had recovere The_ inspectors discussed the event and toured the reactor building with the licensee to conduct temperature surveys and to assess the potential impact of the freezing temperatures in the reactor building on safety system Building temperature surveys conducted with SBGTS running indicated acceptable temperature surveys throughout the building. Temperature surveys conducted-after normal building restoration, including repair of the heating coil, again, yielded acceptable result Cold building temperatures were a result of the reactor building heating coil-being out of service in need of repair. The licensee was able to repair the coil and restore normal building ventilation by late afternoon on 1/6/88.
l The inspector questioned the licensee about the repair and why it wasn't accomplished before 1/6/88 when the licensee had experienced cold building temperatures in November. The licensee explained that the heating coil work was of a lesser priority than more important work. The licensee continued to experience a problem in maintaining the ventilation heating coils as a re-suit of flow control valve malfunctions. The inspector identified and dis-l
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cussed additional concerns with the licensee which are detailed in Inspection Report 88-02.
- 2.0 Core Thermal Power Calculation
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l The licensee notified the resident inspectors that there was a concern that for some period of time the reactor may have been operated at power levels above 1930 megawatts thermal (Kdt), the licensed limit. This concern was generated by a review of plant operating parameters which prompted a calibra-tion check of the feedwater temperature instrumentatio This calibration l check indicated that all components were within the specified tolerances; however, the millivolt / current converter was slightly readjusted to increase the accuracy at the operating point. When the temperature loop was returned
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to service, it was noted that core thermal power, as indicated by the process computer, has increased from 1922 MWt to 1938 Md This indicated increase
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in reactor power level was attributed to the adjustment of the millivolt / cur-rent converter as other plant parameters had not changed. Reactor power level was lowered to below 1930 MWt indicate The resident inspectors met with licensee personnel on 1/8/88 to discuss feedwater temperature calibration. It was concluded that while the tempera-ture calibration methods included each individual component, it was also necessary to provide a icop calibration check from the source to the computer indication, with specified tolerances. The licensee stated that it was their intention to generate a procedure to accomplish thi As part of the core thermal power calculation review, detailed reviews of the feedwater flow calibration procedure 1001.30 (Reactivity Measurements -
Periods) and Technical Data Report (TOR) #842, were performed. The inspector concluded that while the methods and theoretical basis for the calibration were sound, uncertainties in actual feedwater flow could result from changes in the physical characteristics of the flow elements due to erosion or fouling during the period since installation. Also, TDR-842 indicated that the feed- 4 water flow nozzles were not calibrated prior to installation, making it im-possible to verify the nozzle discharge coefficients used in procedure 1001.3 Furthermore, the differential pressure transmitter instrument line taps were modified, but never evaluated. A similar modification at Brown's Ferry re-sulted in almost a 1% change in the nozzle discharge coefficients. All of these factors taken together serve to introduce uncertainty in the feedwater flow input to the core thermal power calculatio In addition, feedwater nozzle coefficient may change with time due to either erosion or corrosion and the nozzles are not routinely calibrated to a known standard which would identify any changes in the nozzle coefficient. The inspectors questioned the licensee if the above identified uncertainties associated with measuring feedwater flow could be bounded within the allowable uncertainty for core thermal power calculation Aaditionally, a review of several reactor plant parameters, panel steam flow, panel feedwater flow, computer steam flow, computer feedwater flow, and con-densate flow showed poor agreemen Specifically, there is a large mismatch between computer feedwater flow ar.d all other similar indications of the same parameter. This concern was expressed to the licensee who initiated mainten-ance actions to determine if any steam flow discrepancies existe The inspectors will continue to evaluate the feedwater flow conditions and its effect on core thermal powe .0 Loss of Process Computer On 1/30/88 at approximately 5:15 PM, the SIGMA process computer failed and efforts were initiated to restore the computer to operation. The operation of the process computer is significant because it uses plant parameters to calculate the core thermal power that operators use to ensure compliance with the licensed limit of 1930 Megawatts thermal (MWt).
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At approximately 9:45 PM the same day, the control room operators performed a manual heat balance using the guidelines of Station Procedure 1001.6, Core Heat Balance - Power Range, Revision 9. The calculated core thermal power using this procedure was 1964.7 MWt. Previously the inspector had questioned the licensee concerning the accuracy of procedure 1001.6 in the event that ,
.a manual heat balance would be required. Apparently, the feedwater flow in- '
dication used for this computation had drifted from its initial calibration and could not be corrected until inore flow data points were obtained. This evolution would require significant reductions in plant operating power level It was decided, after contacting the plant operations manager and the core-manager, to maintain steady state power operation using alternate plant indi-cating parameters. The parameter specified to maintain was the electrical output of the generator (660-665). These levels are commensurate with the levels observed prior to the loss of the process compute Upon change of shif t, at approximately midnight, the relieving shift performed another manual heat balance (1960 MWt) and expressed concern that the process computer was unavailable and that the manual heat balance contained error The plant operations manager was again contacted and it was agreed to reduce plant power to prcvide margin to the licen.e thermal limi plant power was reduced by approximately 20 MWt and a new manual heat balance yielded 1938.75 MWt. As a check of actual reactor power, the Group Shift Supervisor (GSS) used the feedwater calibration procedure to calculate feedwater flow and used this
- as the feedwater flow input to the ma'.ual heat balanc . Subsequently this method was incorporated as a change to the manual heat balance procedure, 1001.6. The GSS was satisfied that the reactor was below its license thermal limi i The process computer was returned to service at approximately 4
- 45 AM on 1/31/8 The inspector is concerned that the feedwater flow millivolt (MV) indicator had a known deficiency and yet this was not formally documented in the control room. The manual heat balance instruction,1001.6, clearly specifies the use of this indicator, yet the control room operators had no confidence in i Without the process computer, there was no direct indication of core thermal power available to the operator Prior to the loss of the computer the inspectors asked core engineering what operator actions would be required if the deficient 1001.6 procedure was use The core engineers stated that the requirements of the procedure would be followed even though it was considered deficient. Initially, operators choose to maintain core thermal power constant as no plant conditions had change Subsequently, though a 20 MWt power reduction was initiated af ter a shift change had occurred. The oncoming shift supervisor expressed concern at ex-ceeding the manual heat balance thermal limit and reduced power. The inspec- '
tors determined that there was no immediate safety concern as no actual ther-j mal power limit was exceeded, but were concerned with regard to the lack of
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action to identify to the operators the deficiencies and lack of direction in specifying operator actions in this situatio The licensee acknowledged the inspectors concerns and as a result planned to review the even .0 Review of NRC-Bulletin Bulletin No. 87-02, Fastener Testing to Determine Conformance with Applicable Material Specifications NRC procurement inspections in the past have included the testing of a sample of fasteners. These sample results have. indicated the need for a larger sampling basis and resulted in the current effort to request licensee's select according to usage and test fasteners to determine if specification require-ments for mechanical and chemical properties are being me The inspector observed the licensee's sampling of safety and non-safety re-lated bolts, studs or screws and nuts. A large portion of the licensee fas-teners are ASTM A193 grades B7 and B The licensee did not have usage data immediately available; therefore the selected fasteners are all ASTM A193 with one exception. The exception is indicated in the attachment to this repor All nuts selected were ASTM 194. In all cases a minimum of 2 of each item was selected to support the laboratory testing for mechanical and chemical propertie The inspector verified that the samples were properly tagged to ensure traceability of the sample to the sample data sheet. Attachment I to this report is a description of the sampling of fasteners selecte .0 Plant Operational Review 5.1 The inspector reviewed details associated with key operational events that occurred during the report period. As summary of these inspection activities follow On 1/13/88 the licensee experienced a "C" condenser vacuum transient which resulted in a vacuum drop from approximately 29 inches HG to approximately 26 inches HG in the "C" condenser. The operators recognized the problem and took actions to avert a possible plant shutdown. Initially large oscillations were experienced in the off gas line flow and operators took action in response to this indica-tion. They then recognized the potential for the steam traps for the "C" inter or af ter steam jet air ejector (SJAE) condenser to be sticking shut. When these steam traps, Y-4-001 and Y-4-007, stick shut, water is not removed from the SJAE condensers and re- .
sults in a decrease in vacuum. Subsequently the steam traps have periodically stuck resulting in a vacuum de:rease and required operator response to free the traps. The inspector questioned the licensee if a repair could be effected to prevent the steam traps i from sticking 3 hut. The licensee presently plans a power reduction to perform an unrelated surveillance and at that time performing maintenance on the steam traps.
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The resident inspector' examined the Oyster Creek emergency diesels for susceptibility to combustion in the exhaust manifold. This re-view was performed in response to a reported problem with opposed piston. diesel configurations. Apparently, oil remains on top of the upper piston after diesel engine operation and, upon shutdown, is able to seep into the cylinder area and into the exhaust manifol On subsequent engine operation, the oil in the exhaust manifold, if sufficient oxygen is present, can ignite. A review of the diesel configuration at Oyster Creek showed that the diesel configuration is of the 'V' type vice the opposed piston typ It is considered unlikely.that this phenomena, as described, could occur in this diesel configuration. Also, discussions with the licensee indicated that previous inspections of the exhaust manifold and the blower diffuser screen showed no evider :e of oil accumulation. The in-spector has no concerns in this are During the report period the licensee initiated several maintenance actions to correct discrepancies with hydraulic control accumulators (HCV). These items included level switch problems on HCU 10-11, immediate maintenance.to correct a leaking-V-111 on HCU 38-19, re-placement of valve 118 on HCU 26-47 and work on the "48" notch in-dication of the position indication probe on HCU 14-47 which was determined to a be drywell problem. The inspector will continue to follow the number of HCU problem During.this period the thermocouple for main steam relief valve NR-28H was declared inoperable as result of erratic indication, The inspector verified that in accordance with technical specifications the gain on the adjacent acoustic monitor, NR28J, was increase In additica to NR-28H, thermocouples for NR-28A and NR-28C had al-ready been declared inoperable and the gain on the adjacent acoustic monitors increase .2 Routine tour s of the control room were conducted by the inspectors during which time the following documents were reviewed:
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Control Room and Group Shift Supervisor's Logs;
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Technical Specification Log;
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Control Room and Shift Supervisor's Turnover Check Lists; I .
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Reactor Building and Turbine Building Tour Sheets;
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Equipment Control Logs; i
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Standing Orders; and,
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Operational Memos and Directives.
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5.3 Routine tours of the facility were conducted by the inspectors to make en assessment of the equipment conditions, safety, and adherence to operating procedures and regulatory requirements. The following areas are among those inspected:
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Turbine Building
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Vital Switchgear Rooms
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Cable Spreading Room
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Diesel Generator Building
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Reactor Building The following additional items were observed or verified: Fire Protection:
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Randomly selected fire extinguishers were accessible and in-spected on schedul Fire doors were unobstructed and in their proper positio Appropriate fire watches or fire patrols were stationed when equipment was out of servic Equipment Control:
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Jumper and equipment mark-ups did not conflict with Technical Specification requirement Administrative controls for the use of jumpers and equipment mark-ups were properly implemente Vital Instrumentation:
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Selected instruments appeared functional and demonstrated parameters within Technical Specification Limiting Conditions
- for Operatio Housekeeping
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Plant housekeeping and cleanliness were in accordance with approved licensee program No inspector concerns were identified.
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6.0 RE03's High Reactor Pressure Switch Ouring this report period, the inspector reviewed the failure of RE03B to trip as required during surveillance procedure 619.3.017, "Reactor High Pressure Scram Test and Calibration," on.12/17/87. The licensee determined the micro-switch to exhibit poor repeatability and placed the second microswitch in service. During the next surveillance on 1/17/88 the second microswitch in RE038 failed the surveillance test and again exhibited poor repeatabilit This repeatability problem was slightly different from the first RE03B failure in that it exhibited double trip point indication The licensee placed-a 1/2 scram in the reactor protection system and replaced the failed micro-switches.on RE03B. The instrument successfully passed the surveillance after switch replacement. The failed microswitches were sent to the licensee laboratory for analysi Earlier the RE03A microswitches failed during a surveillance and were replace The licensee conducted an analysis on the failed microswitch which indicated that carbon residue from apparent electrical arcing was found on the contact surfaces. The licensee believes that for the long period of operation the switch has been in service that the electrical arcing would occur as a result of the contact almost being made up as result of the close tolerances required by the switch setpoin The licensee is currently evaluating the switch performance and searching for a suitable-instrument replacement. The inspector will continue to follow switch performance and the licensee's effort to find a suitable replacemen .
7.0 Review of Nine Mile Point One Feedwater Transient Problem On 12/19/87, Nine Mile Point One experienced a feedwater transient that re-suited in a significant water hammer event. One of the root causes of the event was cyclic fatigue failure of the feedwater flow control valve stem and plug connection due to excessive vibration of the feedwater line The Nine Mile point flow control valve stem and disc had been fillet welded to correct early vibration problem The inspector reviewed the Nine Mile event with the licensee to determine applicability to the Oyster Creek facility. It was determined that both Nine Mile Point and Oyster Creek use a Fisher-Vulcan Valve. The licensee determined from a review of operating history that no event similar to the Nine Mile water hammer event had occurred at Oyster Creek. In addition, the licensee inspects the feedwater valve internals every outage and plans to accomplish the same activity when they shut down for the 12R outage. During the inspection, the licensee stated that they will inspect for signs of fa-tigue failur The inspector had no further questions on this matte __ __
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8.0 Snubber Visual Inspection On 1/7/88 the inspector discussed results of a visual snubber inspection con-ducted under surveillance procedure, 675.1.001, Inspection of Bergen-Paterson hydraulic Snubbers with a licensee worker. The inspector learned that the baseplate associated with a core spray system snubber NF-2-S7 potentially had oversize bolt holes. Initially the material nonconformance report (MNCR)
written on this problem was inspected by Quality Control and dispositioned as an optical illusion. The inspector discussed this with maintenance and auality-control personnel. Apparently the quality control-(Q.C.) people in-spected the snubber baseplate (75' elevation) from the floor elevation level (51') with a video camera. The video camera inspection was not adequate to reveal the oversize baseplate holes; and therefore, Q.C. incorrectly disposi-tioned the MNC The inspector concluded that better communication could have-existed between Q.C. and maintenance in resolving this nonconformance. In addition, the inspector questioned licensee personnel with regard to his per-ception of a reluctance of maintenance personnel to question Q.C.'s disposi-tion of the MNCR. The maintenance people stated that there was no reluctance but that the supervisor was not able to immediately address the issue with This was later confirmed by maintenance managemen The inspector noted that earlier inspections of the base plate had not iden-tified the oversized hole As a result ~of the inspectors concerns, the licensee accompanied the inspector in witnessing a core spray surveillance test to determine if the baseplate was moving. The inspector observed no movement in the baseplate or excessive vibration in the core-spray line. Howe /er, the licensee has recently changed the manner in which the core spray surveillance is conducted. In an effort to reduce potential water hammer problems with the core spray test line, the test valve is slowly jogged open which overall results'in less system vibra-tio In addition during bulletin 79-14 inspection conducted in 1985 the bolts were found to be loose and had to be retorque The inspector concluded that adequate attention was not paid to inspection of the baseplate during previous inspection The inspector will continue to review this area in future inspection activitie In addition there was a question regarding the procedural requirement (675.1.001, paragraph 6.6.3.1) of a minimum thread engagement of 1/2 of the available thread length of the piston rod contacting the paddle threads be-tween the piston rod and the monitoring paddle. The licensee changed the procedure to require full thread engagement instead of 1/2 thread engagemen . "C" Electromatic Relief Valve (EMRV)
The "C" EMRV acoustic monitor alarmed periodically on 1/25/88, but the down-stream thermocouple temperatures gave no indication of an open EMRV. The operators reduced power to 1900 MWt in an attempt to clear the alarm conditio The power reduction was successful in clearing the alarm and also resulted in a slight decrease in the magnitude of the vibration. The alternate acous-
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tic monitor was selected to determine if a malfunction existed particular to the primary acoustic monitor. The alternate acoustic monitor also exhibited the same behavior. In addition, the adjacent acoustic monitors exhibited the same acoustic behavior but displayed a lower magnitude. The licensee recorded the noise signature of the "C" EMRV for analysis. The inspector reviewed the-analysis with the licensee. The licensee indicated that.the noise signature frequency might be that associated with steam flow through a seal area or gasket or simply-flow noise and not that of a pilot valve disc chatter. The inspector questioned if the EMRV had to be actuated if the potential existed that the EMRV would not be able to reseat as pressure may not be able to build up in the main disc chamber area to effect reseating. The licensee determined that any potential leakage past the pilot valve seat would be of small enough magnitude to allow the main disc to reseat. The licensee discussed this with the valve vendor and currently plans to have an outside contractor perform and analyze the noise signature of the "C" EMRV. The inspector will continue to follow the licensee efforts in this are .0 (0 pen) Unresolved Item (219/87-33-01): "C" Main Station Battery Disconnected The inspector conducted further review of this event concerning the alarm capability'for the "C" aain station battery. Annunciator "Bus C Input BRKRS (9xF-5-d)" open, alarms when either of the following occur, both "C" battery charger's output DC breakers are open or the "C" battery output breaker is open. The inspector reviewed Station Procedure 634.2.001, "Main Station Bat-tery Discharge and Low Voltage Relay Annunciator Test," and discussed battery breaker lineup for the' performance of the equalizing charge and test discharge i of the 125 VDC battery with the licensee. The "C" battery breaker, in ac-cordance with procedure, was open for the entire period, approximately 60 days,
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and thus the alarm window should have been lighted. The licensee conducted maintenance to determine if this annunciator was functioning properly during i the event as the operators were not aware of the annunciator ever being in
! an alarm condition during the 60-day period the "C" battery breaker was open.
l The licensee lifted a lead to the annunciator and received an alarm light in the control room. Presently the licensee suspects the contact switch in the
"C" battery breaker malfunctioned and will test this at the first available opportunit The inspector will review this effort.
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11.0 Radiation Protection During entry to and exit from the RCA, the inspectors verified that proper warning signs were posted, personnel entering were wearing proper dosimetry, personnel and materials leaving were properly monitored for radioactive con-tamination, and monitoring instruments were functional and in calibratio Posted extended Radiation Work permits (RWPs) and survey status boards were reviewed to verify that they were current and accurat The inspector ob-served activities in the RCA to verify that personnel complied with the re-quirements of applicable RWPs and that workers were aware of the radiological conditions in the area.
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12.0 Observation of Physical Security
.During daily tours, the_ inspectors verified-that access controls were in accordance with the Security Plan, security posts were properly manned, pro-tected area gates were locked or guarded and that isolation zones were free of obstructions. The inspectors examined vital area access points to verify that they were properly locked or guarded and that access control was in accordance with the security pla .0 Review of Periodic and Special Reports Upon receipt, periodic and special reports submitted by the licensee pursuant to Technical Specification requirements were examined by the inspectors. This review included the following considerations: the report includes the infor--
mation required to be reported to the NRC; planned corrective actions are adequate for resolution of identified problems; and the reported information is vali The following reports were reviewed:
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Monthly Operating Report for December 198 Special Report 87-07 dated 11/19/87 concerning a non-functional fire barrier. A fire watch was established in accordance with the technical specification requirements until repairs were mad Safeguards events log quarterly submittal dated January 20, 198 .0 Instrumentation and Control Technician Concerns The inspector reviewed the licensee's preliminary findings and recommendations regarding their review of some concerns raised by an instrumentation and con-trol (I&C) technician. The report made recommendations for implementation by licensee management and concluded there were no nuclear safety concerns identified. The licensee presently plans to interview the remaining I&C technicians and to incorporate any additional findings in a final report to be issued on their review of this area. The inspector will review this report when it is completed by the license .0 Exit Interview A summary of the results of the inspection activities performed during this report period were made at meetings with senior licensee management at the end of this inspectio The licensee stated that, of the subjects discussed at the exit interview, no proprietary information was include ~
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ATTACHMENT 1 FASTENER SURVEY'
ID# SIZE MAT'l SPE ~0ESCRIPTION 001 1/4-20VNC X 1.50- ASTM A193 B8M SCREW HEX CAP 002- 5/16-18UNCX 2.0 , ASTM A193.08M SCREW HEX CAP 003 3/8-16UNC X 2.25 ASTM A193 B7 SCREW HEX CAP
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004 7/16-14UNCX 1.00 ' ASTM A193 B7 SCREW HEX CAP 005 1/2-13UNC X 2.00 . ASTM A193 87 SCREW,HVY HEX 006' 9/16-12VNCX 1.00 ASTM A193 D8M SCREW,HVY HEX 007 -5/8-11UNC X 6.00 ASTM A193 D8M . SCREW HVY HEX 008 3/4-10VNC X 2.00 ASTM A193'B7 SCREW HVY HEX 009 7/8-9UNC'X'3.00 . ASTM A193 B7 SCREW HVY HEX 010 1-8 UNC X ;2.50 ASTM 193 08M- SCREW HVY HEX SAFETY RELATED NUTS 011 1/4-20 UNC ASTM A194 8M NUT HVY HEX 012 5/16-18 UN ASTM A194 8M NUT HVY HEX 013 3/8-16 UNC- ASTM A194 2H NUT HVY HEX 014 7/16-14 UNC- AS!!! A194 2H NUT HVY HEX 015 1/2-13 UNC ASTM A194 2H NUT HVY HEX 016 9/16-12 UNC- ASTM A194 8M . NUT HVY HEX 017 5/8-11 UNC ASTM A194 8M NUT HVY HEX 018 3/4-10 UNC ASTM A194 2H NUT HVY HEX 019 7/8-9 UNC ASTM A194 2H NUT HVY HEX 020 1-8 UNC ASTM A194 8M NUT HVY HEX NON-SAFETY RELATED BOLTS, STUDS, OR SCREWS 021 1/2-13 X 2.50 ASTM A193 B8 SCREW, HEX CAP 022 5/16-18 X 1.50 ASTM A193 B7 SCREW, HEX CAP 023 1/2-13 X 1.25 ASTM A193 B8 SCREW, HEX CAP 024 5/8-11 X 4.00 ASTM A193 B7 STUD 025 1 1/8-7 X 6.50 ASTM A193 B7 STUD 026 3/4-10 X 4.50 ASTM A325 BOLT 027 7/8-9 X 4.00 ASTM-A193 B7 STUD 028 3/8-16 X 1.75 ASTM-A193 87 STUD 029 1 1/4-8 X 1 ASTM A193 B7 CAPSCREW 030 1/1/2-8 X 9.25 ASTM A193 B7 STUD
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