IR 05000461/1998019
| ML20196D168 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 11/25/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20196D161 | List: |
| References | |
| 50-461-98-19, NUDOCS 9812020169 | |
| Download: ML20196D168 (46) | |
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' U.S. NUCLEAR REGULATORY COMMISSION
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REGION lll Docket No:
50-461 License No:
NPF-62 l
Report No:
50-461/98019(DRS)
Licensee:
lilinois Power Company
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Facility:
Clinton Nuclear Power Statica
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Location:
Route 54 West Clinton,IL 61727 Dates:
September 14 - October 9,1998-Inspectors:
G. Hausman, Team Leader
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J. Neisler, Team Member D. Schrum, Team Member
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T. Tella Team Member.
A. Walker, Team Member-J Approved by:
R. Gardner, Chief, Engineering Specialist Branch 2 (ESB2)
i Division of Reactor Safety
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9812020169 981125 V
PDR ADOCK 05000461
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TABLE OF CONTENTS EXEC UTIVE SU M MARY............................................... 2 E1 Conduct of Engineering E1.1 Review of Design Changes, Modifications and Technical issues.......... 4 E1.2 Safety Screenings and Evaluations.................................. 9 E1.3 Assessment of the Shutdown Service Water System.................. 10 E1.4 Adequacy and Control of Calculations.............................. 11 E1.5 Adequacy and Control of the Setpoint Program....................... 12 E3 Engineering Procedures and Documentation E3.1 Condition Reports.............................................. 13 E3.2 Post / Stamp Affixed Program...................................... 13 E3.3 Surveillance Procedures........................................ 14 E3.4 System Health Report.......................................... 15 E5 Engineering Staff Training and Qualifications E5.1 System Engineering............................................ 16 E8 Miscellaneous Engineering issues..................................... 17 M2 Material Condition of Facilities and Equipment M2.1 Plant Walkdowns............................................... 22 X1 Exit Meeting Su mmary............................................... 22 Partial List of Persons Contacted...................................... 23 Inspection Procedures Used.......................................... 24 Items Opened, Closed or Discussed.................................... 25 List of Acronyms Used............................................... 27 Partial List of Documents Reviewed.................................. 28
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EXECUTIVE SUMMARY Clinton Nuclear Power Station, Unit 1 NRC Inspection Report 50-461/98019 An announced five member team inspection was performed from September 14 through October 9,1998. The team reviewed the engineering and technical support (E&TS)
organization's effectiveness in the performance of routine and reactive site activities, including controls for the identification, resolution and prevention of technicalissues and problems that could degrade the quality of plant operations or safety. In addition, the team reviewed selected items of the Clinton Restart Action Plan for Design and Configuration Control (Case Specific Checklist Matrix Items VI.1, VI.2 and VI.3). As a result of this inspection, one violation and one non-cited violation (NCV) of Nuclear Regulatory Commission (NRC) requirements were identified.
Enaineerina Overall, the E&TS inspection team concluded that the engineering staff was effective in e
the identification of technical problems. Based on their review of selected system modifications, the team did not identify additional problems that had not been previously identified by the licensee. The team concluded that the system design and functional verification (SDFV) program reviews conducted on the residual heat removal and shutdown service water (SX) systems identified significant issues and the quality of those reviews was considered excellent. As a result of these reviews, however, a significant amount of corrective action work was identified that needed to be completed prior to restart. Through self-assessments, the licensee exhibited a pro-active trend in the attempt to disclose performance problems within the engineering organization.
The team concluded that no major problems existed with the hardware change process e
or with the selected hardware changes reviewed that had not been previously identified by the licensee. The technical quality of the selected engineering work products was generally sound and the hardware changes reviewed were adequately implemented. A violation for which enforcement discretion was exercised was identified involving the installation of a minor modification which caused the loss of suppression pool cooling.
An NCV was identified regarding the licensee's failure to take adequate and timely actions for excessive silt accumulations in the SX pump intake area. Present performance by the licensee on the SX system was excellent. However, the team noted that the licensee had not generated maintenance work requests (MWRs) or preventive maintenance tasks to assure replacement of limited life non-environmentally qualified equipment in the plant. The team considered this a weakness. (Section E1.1)
The team concluded that the licensee had an acceptable 10 CFR 50.59 program and e
that qualified personnel prepared and reviewed the 10 CFR 50.59 screenings and safety evaluations. The team also concluded that the 10 CFR 50.59 screenings and safety evaluations reviewed were adequate with the exception of some minor errors. Although no specific issues were identified by the inspectors, the number of licensee identified
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condition reports (CRs) conceming safety screenings and evaluations revealed that problems still exist. (Section E1.2)
The inspectors concluded that the SX system design and configuration controls were e
adequate. The SDFV assessment of the SX system was very comprehensive and thorough. The team concluded that actions taken as the result of the SDFV assessment
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would correct many operational problems and concerns with the SX system.
(Section E1.3)
The team concluded that the licensee's revised methodology for reviewing calculations in e
order to determine the technical adequacy of the CPS calculation program was
satisfactory. However, insufficient activities were completed by the licensee at the conclusion of this inspection to support an adequate review by the NRC. As a result, the inspection of this issue could not be completed during the E&TS inspection.
(Section E1.4)
The team concluded that the licensee's Setpoint Program Action Plan methodology was e
sound. However, insufficient activities were completed by the licensee at the conclusion l
of this inspection to allow for an adequate review by the NRC. As a result, the inspection of this issue could not be completed during the E&TS inspection. (Section E1.5)
in most instances, the corrective action process for the CRs selected for review was e
adequately implemented and resulted in acceptable corrective actions. (Section E3.1)
The team concluded that the lack of a setpoint control program and a lack of trending of
safety-related and maintenance rule-related instrument drift was a weakness. The team also concluded that lack of supporting calculations for important instrument setpoints was a weakness. (Section E3.3)
Based on their system reviews, the team concluded that the System Health Report e
provided an accurate accounting of system status with regard to the numbers of CRs, MWRs, etc. No major discrepancies were identified with the System Health Report for the systems reviewed. The automatic depressurization system status could not be
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reviewed since the system was not addressed in the System Health Report.
(Section E3.4)
The team concluded that the system engineers were generally qualified and e
experienced. However, the team identified a weakness in that detailed training was not provided to the system engineers for their assigned systems. On October 10,1998, the team was notified that system engineers would receive senior reactor operator system training for their assigned system (s). (Section E5.1)
The material condition of the walked down systems appeared to be good. The system e
engineers appeared to be knowledgeable of the systems. (Section M2.1)
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Report Details ill Enaineerina The purpose of this inspection was to review Clinton Power Station (CPS) activities in the areas of engineering and technical support (E&TS) and the status and effectiveness of licensee actions to address selected CPS Restart Action Plan Matrix Items. The inspection team focused on selected engineering design changes, modifications and technical issues related to the automatic depressurization system (ADS), direct current (DC) electrical distribution system, high pressure core spray system (HPCS), residual heat removal (RHR) systern and the shutdown
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service water (SX) system. For the selected engineering work activities, the inspection team assessed engineering staffinvolvement in root cause analysis,50.59 safety evaluations, operability assessments, trending, surveillance and post-modification testing (PMT), control and revision of drawings, the corrective action process, maintaining the licensing basis and Updated Safety Analysis Report (USAR) commitments. In addition, CPS Restart Action Plan Case Specific Checklist Matrix Items VI.1, VI.2 and VI.3 were reviewed.
E1 Conduct of Engineering E1.1 Review of Desian Chancess Modifications and Technical Issues a.
Inspection Scope (IP37550: IP37700)
The team examined electrical, mechanical, and control and instrumentation hardware changes (i.e., design changes and modifications)in various stages of implementation and other technical issues. The methods used to control hardware changes at CPS were reviewed to verify adequacy of control and compliance with regulatory requirements. The team's review included permanent and temporary hardware changa packages, condition reports (CRs), safety reviews, operability determinatioris ci d screenings, plant safety tagging, self-assessments and applicable CPS procedures. The selected hardware changes and technical issues were discussed with cognizant licensee i
personnel and the inspectors walked down selected accessible portions of the affected systems. In addition, the team reviewed selected calculations and setpoint/ scaling
change requests which supported the plant system hardware changes.
b.
Observations and Findinas At the CPS, major and minor hardware changes were identified as modifications and engineering change notices (ECNs), respectively. As part of the inspection process, the team's review included both modifications and ECNs. The team reviewed the following selected modifications and ECNs to determine whether the licensee adequately implemented the pmpused hcrdware change design, installation, and testing requirements and documented the required reviews and approvals:
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Loss of Suppression Pool Cooling A minor modification placed the plant in a condition that was outside the plant's original design basis. ECN 30211 " Install Annunciator Bypass Switch," was developed and installed in response to a 10 CFR Part 50, Appendix R " hot short" concern identified in CR 1-97-06-181," Inadequate 8 Hour Battery Powered Emergency Lighting," dated June 17,1997. The ECN and its associated procedure revisions closed the normally open RHR shutdown cooling suction valve 1E12-F006B, removed power to the F006B valve by opening the supply breaker and bypassed the valve's associated control room annunciator to prevent nuisance (i.e., open supply breaker) alarms. The licensee had the option of providing an Appendix R emergency light for the manual operation of the F006B valve. Normally, a hot short fix does not prevent a hot short from repositioning the valve, but rather it prevents the valve from being damaged due to bypass of the limit and torque switches. However, this ECN removed the power to prevent any movement of the valve during a fire.
The installation of the ECN began July 7,1997, and the RHR B loop was declared operable on July 15,1997, for an anticipated plant startup date of August 1,1997. The design change activities resulted in the inability to operate the RHR B loop in the suppression pool cooling (SPC) mode and was discovered when operators were unable to initiate SPC from the main control room during performance of CPS Procedure 3312.01, " Residual Heat Removal" on December 22,1997. The operator's inability to initiate SPC was due to the loss of control power to the interposing relay (ECN opened
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the F0068 supply breaker) which controlled the F006B valve interlocks associated with RHR suppression pool suction motor-operated gate valve F004B and test return to suppression pool valve F024B. The F006B valve was interlocked with the F004B and F024B valves to prevent an accidental reactor vessel drain down. A reactor vessel to suppression pool drain down flow path could exist if the F006B valve was open and the other two valves were not interlocked closed. However, with the interposing relay de-energized, the F004B and F0248 valves could not be opened to establish SPC.
The safety significance of an inoperable RHR B Loop was not fully identified until August 18,1998, as documented by CR 1-98-08-206," Missed Design impact May Affect RHR B Suppression Pool Cooling," Revision 0. During a design basis loss-of-coolant accident (LOCA) scenario, a complete loss of SPC would have occurred. The plant's design basis assumed a Division A diesel generator failure (single failure criterion), which would have made the RHR A Loop inoperable. The RHR C Loop was not designed for SPC and the RHR B Loop was not available as a result of ECN 30211. In addition, radiation levels during the LOCA would have prevented manual actions (i.e., opening valves F0048 and F024B (1000R/Hr) or closing the F0068 supply breaker (100-500R/HR)) to restore SPC. The failure to identify, during the modification review process, that a change to the plant's original design basis occurred constitutes a violation of 10 CFR Part 50, Appendix B, Criterion lil," Design Control" (VIO 50-461/98019-01), which requires that measures be established to assure that applicable regulatory requirements and the design basis as specified in the license application, for structures, systems, and components are correctly translated into specifications, drawings, procedures, and instructions.
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The licensee failed to identify the design problem during the modification process. In addition, it was not identified when the licensee performed a 10 CFR 50.59 safety evaluation for this modification. The licensee failed to identify this condition during PMT because only valve stroking was performed. A contributing cause was a non-standard interlock design with the interposing relay control power supplied from the supply breaker for valve F0068.
l The licensee's immediate corrective action was to invoke a mode 1,2 and 3 restraint on December 22,1997. The licensee performed a modification to remove the interlocking
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relays and replace them with hard contact limit switches on the motor-operated valve operators. In addition, the licensee's corrective actions included implementation of a
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Senior Engineering Review Group overview function to review selected engineering work products. The engineering organization was augmented with personnel experienced in the evaluation and maintenance of plant design bases. Engineering personnel were provided with training on the lessons teamed from this issue. Finally, a new engineering leadership team was established to focus on critical assessment and ongoing maintenance of the plant design bases.
The failure to identify that a modification placed the plant outside of the design basis was a violation of 10 CFR Part 50, Appendix B, Criterion Ill; however, enforcement discretion was applied based on the following: (1) significant NRC enforcement action was previously taken for design control problems; (2) additional enforcement action was not considered necessary to achieve remedial action for the violation due to CPS's commitment in its Plan For Excellence to take actions to address design control issues prior to plant restart; (3) the violation was related to problems which were present prior to the events leading to the extended shutdown; (4) the violation was not classified at a severity level higher than Severity Level ll; (5) the violation was not willful; and (6) CPS committed to provide reasonable assurance that safety-related systems, structures and
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components will perform their intended safety functions as described in the design and licensing basis prior to plant reetart as noted in the NRC Manual Chapter 0350 Restart Panel's Case Specific Checklis: Item VI.1.
Degraded Voltage Modification The inspector reviewed modification packages for the installation of Static Volt Ampere Reactive (VAR) Compensators (SVCs) and the replacement of non-regulating 480/208/120 Volt distribution transformers. In addition, the inspector observed installation activities at the work locations.
Per the CPS design basis, CPS should have two offsite power sources with adequate capacity and capability to provide reliable power to the plant under all operating conditions. The licensee determined that the offsite power sources were no longer capable of supplying reliable power under degraded grid voltage conditions when plant loads were served through the reserve auxiliary transformer (RAT) and/or the emergency reserve auxiliary transformer (ERAT). As a result of the degraded voltage issue, the RAT and ERAT were placed in a Technical Specification (TS) inoperable condition.
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The licensee's corrective action was to replace the ERAT with a new transformer that had automatic tap changing capability. Non-regulating 480/208/120 Volt distribution transformers were replaced with voltage regulating transformers. The SVCs were being installed on the low sides of both the RAT and ERAT. The SVCs being installed at each transformer had a 28.5 mega-VAR (MVAR) capacitive and 14 MVAR inductive rating.
The licensee's calculations indicated that this rating was sufficient to maintain in-plant voltages above minimum design levels.
The licensee determined that the installation of this modification could constitute an unreviewed safety question (USQ). The USQ evaluation was fonnarded to the NRC Office of Nuclear Reactor Regulation (NRR) for review. The results of the NRC's review had not been completed by the end of this inspection.
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DC Battery Charger Modifications The licensee issued ECNs 30992,30660,30699, and 30430 to provide schematic and wiring diagram changes for changing the tap settings for the divisional battery charger power supply transformers. The tap settings were based on Calculation 19-AK-13, Revision O, Volume A, which assumed the tap settings would be set at 470 Volts to ensure that minimum required voltages would be available to the battery charger during normal and degraded voltage conditions. The team verified that safety evaluation screenings had been performed, calculations had been reviewed, and applicable drawings were included with the ECN packages. In addition, engineering personnel had ensured that appropriate PMT requirements were included in the work packages.
The licensee developed ECN 30445 to change the high voltage shutdown card reset point from 132 volts to 135 volts which allowed the card to reset during charger surveillances when voltage spikes may occur. The ECN also changed the surveillance data sheet to reflect the reset change and the associated reset delay time from 30 seconds to a nominal 10-15 second time delay. The inspectors verified that the licensee's staff had performed appropriate engineering evaluations and that the revised changes had been incorporated into the applicable data sheets and specification (K2989). A detailed PMT was included in the change package.
Valve and Alarm Setpoint Modifications The team reviewed several completed hardware change packages related to ADS valve and alarm setpoint changes. The following documents were reviewed by the team:
Modification ilA-021, " Reposition Valve Limit and Torque Switches for e
Valves 11A-012A and ilA-013A,"
ECN 27675, * Replacement of Target Rock Valves ilA-044A and ilA-044B,"
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ECN 28365, " Increase of Stroke Time of Valve ilA-013B" and
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ECN 30225," Revision of ADS Instrument Air Header Low Pressure Alarm
Setpoint."
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The hardware change packages clearly described the proposed design changes and l
justification for the changes. The modification package contained an adequate 10 CFR 50.59 safety evaluation. The ECN packages contained necessary safety screenings.
The affected drawings and the required USAR changes were noted in the design packages. The PMT was performed as necessary.
Replacement of Limited Life Equipment While reviewing ECN 30225, the team noted that ADS differential pressure switches IPS-!A084 and 085 (ITT Barton type 580A) had an approved life of 20 years. These switches were not required to be included in the station's environmental qualification (EQ) program. However, the need to replace these maintenance rule-related switches
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was not specified in any preventive maintenance (PM) procedure or a maintenance work request (MWR). The lack of specific procedures for replacing limited life non-EQ plant equipment was considered a weakness in the licensee's PM program.
Excessive Silt Accumulations in the SX Pump Intake Area In August 1992, during a licensee assessment of the CPS program for Generic Letter (GL) 89-13, " Service Water System Problems Affecting Safety-Related Equipment,"
dated July 18,1989, quality assurance (QA) personnel identified the need to establish acceptance criteria for the levels of silt and other potential fouling mechanisms in the SX pump intake area. On several occasions excessive sitt had been observed by licensee personnel and removed from the screen house area. However, the criteria as to what constituted acceptable silt levels had not been established since CPS began operation.
No written response was required for the identified QA concern and no action was taken to address the issue.
During the licensee's integrated safety assessment (ISA) review of CPS GL 89-13 Program in 1997, criteria for acceptable silt levels were again questioned for the SX intake area. This concern was identified in CR 1-97-10-054, " Indeterminate Condition Because of Silt in the SX/ Unit 1 FP (Fire Pump) Bay Area of the Screen House," dated October 3,1997. In response to this concern, the maximum acceptable silt levels were calculated. The licensee determined that existing sitt levels exceeded the calculated maximum levels and that Division 1 and Division 2 SX pumps were inoperable. As a result of the inoperable pumps, Licensee Event Report (LER) 461/97-026, " Inadequate Procedure for inspection of Shutdown Service Water Pumps for Excess Silt Results in shutdown Service Water Pump inoperability," was issued on March 3,1998.
In response to the LER, licensee personnel promptly removed silt accumulation from the area to meet acceptable levels. In addition, CPS Procedure 2400.01, "Corbicula (Asiatic C!am) Control," was revised to require the inspection for and removal of silt ecumulations in excess of acceptable levels. The procedure also required a yearly
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inspection for accumulation of silt and other fouling agents and removal of the accumulations as needed.
Criterion XVI of 10 CFR Part 50 requires that conditions adverse to quality be promptly identified and corrected. The failure to take adequate and timely action to correct the SX pump intake silting problem, which was identified as early as 1992, is considered a corrective action violation (NCV 50-461/98019-02(DRS)). Since the problem was non-repetitive, identified by the licensee and necessary actions to correct and prevent
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recurrence of the violation had been completed, this violation is being treated as a NCV, consistent with Section Vll.B.1 of the NRC Enforcement Policy and no response is required.
SX Flow Balancing Without Backwash Flow Considerations During flow balance testing of the Division 1 SX system on September 11,1998, licensee l
personnel noted that, with strainer backwash in service, flow to several of the SX loads l
was below required minimums. Condition Report 1-98-09-201 was written to address the problem. During the team's review of this issue, the inspectors noted that the
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licensee had concluded in Operability Determination 1-98-09-201-0, that the reduced l
flow was not greal enough to substantially affect the short term operability of the affected
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l equipment. The teara determined that the licensee's conclusions were adequately Justified.
c.
Conclusions l
The team concluded that no major problems existed with the hardware change process l
or with the selected hardware changes reviewed that had not been previously identified by the licensee. The technical quality of the selected engineering work products was
generally sound and the hardware changes reviewed were adequately implemented. A i
violation for which enforcement discretion was exercised was identified involving the j
installation of a minor modification which caused the loss of SPC. An NCV was identified j
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regarding the licensee's failure to take adequate and timely actions for excessive silt
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accumulations in the SX pump intake area. Present performance by the licensee on the SX system was excellent. However, the team noted that the licensee had not generated MWRs or PM tasks to assure replacement of limited life non-EQ equipment in the plant.
The team considered this a weakness.
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E1.2 Safety Screeninas and Evaluations a.
Inspection Scope (IP37001)
l The team reviewed the implementation of the 10 CFR 50.59 program including
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l procedures for screening changes, tests, and experiments and preparing safety evaluations; the processes for maintaining records, updating the USAR, and reporting to the NRC; and the training and qualifications of 10 CFR 50.59 screening and safety evaluation preparers. The team's review consisted of a selected sample of 10 CFR
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50.59 screenings and safety evaluations, where emphasis was placed on the design change process.
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Observations and Findinas
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Overall, the team determined that the screenings and safety evaluations were appropriately prepared and were consistent with licensee procedures. In particular, the team determined that the preparers reviewed appropriate documents during the preparation of 10 CFR 50.59 screenings and safety evaluations; the 10 CFR 50.59 screenings and safety evaluations adequately addressed the effects of the proposed changes on plant operations, interactions with other systems and components due to the changes, any new failure modes introduced by the changes, and the effects of the
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changes on existing accident and transient analyses; and the 50.59 safety evaluations adequately addressed unreviewed safety question criteria.
However, during the 10 CFR 50.59 program review, the team noted that CR 1-98-05-081 was written for an inadequate 10 CFR 50.59 safety evaluation. The problem, as stated in the CR, read, "... SX heat exchanger minimum flows were changed without discussion in the 50.59 safety evaluation." In addition, the CR contained the statement,
"This is the third instance during the SX system design and functional verification (SDFV)
of design values being changed without discussion in the applicable 50.59." Based on this CR, it was evident that 10 CFR 50.59 screening and evaluation problems have not been completely eliminated.
c.
Conclusions Based on the inspection results, the team concluded that the licensee had an acceptable 10 CFR 50.59 program and that qualified personnel prepared and reviewed the 10 CFR 50.59 screenings and safety evaluations. The team also concluded that the 10 CFR 50.59 screenings and safety evaluations reviewed were adequate with the exception of some minor errors. Although no specific issues were identified by the inspectors, the number of licensee identified CRs conceming safety screenings and evaluations revealed that problems still exist.
E1.3 Assessment of the Shutdown Service Water System a.
Inspection Scope (IP37550)
The team reviewed the SX system to assess the system's ability to perform its intended safety function as described in the design and licensing basis (Restart Action Plan Case Specific Checklist Matrix Item VI.1). The inspectors reviewed the licensee's SDFV SX system assessment. The assessment was discussed with cognizant individuals who participated in the assessment as well as other individuals who were knowledgeable of the SX system. Selected SX related CRs and other system related documents were reviewed for appropriate actions.
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Observations and Findinas The SDFV review of the SX system was conducted in late 1997 and early 1998, primarily using independent outside contractors. The SDFV assessment resulted in the identification of a number of system related problems, sevaral of which were significant.
The final report, issued on May 13,1998, listed 50 separate SX related problems which were documented in CRs.
The team noted that many of the identified system problems had been corrected and actions for the other problems were underway or were scheduled to be completed in the near future. The problems had been evaluated and actions that affected important safety-related equipment were scheduled for completion prior to plant startup. The inspectors reviewed selected CRs to verify that the planned or completed actions were adequate. In addition, the respective design engineers were interviewed and based on these interviews, the inspectors determined that the engineers were knowledgeable of the hardware changes and the effect of the design changes on the SX system. The recent SX system hardware changes reviewed by the team consisted of minor modifications since no recent major SX system modifications had been completed, c.
Conclusions
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Based on the team's review, the inspectors concluded that the SX system design and configuration controls were adequate. The SDFV assessment of the SX system was very comprehensive and thorough. The team concluded that actions taken as the result of the SDFV assessment would correct many operational problems and concerns with the SX system.
E1.4 Adeauacy and Control of Calculations a.
Inspection Scop _e (IP37550)
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The team reviewed the adequacy and control of calculations to assess how effectively the licensee had implemented calculation and control processes (Restart Action Plan Case Specific Checklist Matrix Item VI.2).
b.
Observations and Findinas During the inspectors' initial meeting with the licensee, the licensee referenced the results of the calculation reviews associated with the SDFV and system surveillance test review (SSTR) programs to support their conc'usion that plant calculations were technically adequate. The team questioned this conclusion since the SDFV and SSTR reviews were performed to provide a functional design and licensing basis requiremen;s review and were not performed as part of a detailed calculation design review. Howeve',
the licensee had concluded based upon the quantity of the calculations reviewed in connection with these programs that plant calculations were technically adequate. The team stated that the only licensee program that performed a detailed calculation design review was the Detailed Design Review (DDR) Program, which had reviewed 40
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calculations to date.
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During followup discussions with the licensee, the licensee presented a revised approach which involved a more systematic method to support the technical adequacy of plant calculations. Although a revised methodology was adopted by the licensee, it had not yet been fully implemented. Consequently, the team could not evaluate its adequacy. Numerous reviews remained to be conducted by the licensee.
c.
Conclusions The team concluded that the licensee's revised methodology for reviewing calculations in order to determine the technical adequacy of the CPS calculation program was satisfactory. However, insufficient activities were completed by the licensee at the conclusion of this inspection to support an adequate review by the NRC. As a result, the inspection of this issue could not be completed during the E&TS inspection.
E1.5 Adeouacy and Control of the Setooint Proaram a.
Inspection Scoce (IP37550)
The team reviewed the adequacy and control of the setpoint program to assess how effectively the licensee had implemented setpoint control processes (Restart Action Plan Case Specific Checklist Matrix Item VI.3).
b.
Observations and F*ndinas The team's review of the CPS Setpoint Program was limited since the licensee's Setpoint Program Action Plan was not approved until September 30,1998. The inspectors reviewed the approved action plan, which outlined the licensee's strategy to address the setpoint program concerns using a two phase approach.
The licensee stated that Phase 1 involved establishing administrative and technical guidance to ensure adequate configuration manegement of plant instrument setpoints and to provide confidence that the plant's safety-related and accident mitigation instrumertation would conservatively satisfy its intended functions. Phase 1 activities were scheduled to be completed prior to restart.
Phase 2 consisted of implementing a long term calculation upgrade plan, which was intended to correct discrepancies identified in the licensee's design basis instrumentation setpoint program. The licensee also expected this upgrade plan to include the development of documentation to ensure that continued operations and maintenance activities are conducted within those design bases. Phase 2 activities would be implemented following restart.
The inspectors did not identify any problems with the licensee's planned approach; however, initial discussions with the licensee revealed that this plan was in its infancy and a substantial amount of work remained before it was ready to be reviewed further by the NRC.
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c.
Conclusions The team concluded that the licensee's Setpoint Program Action Plan methodology was sound. However, insufficient activities were completed by the licensee at the conclusion of this inspection to allow for an adequate review by the NRC. As a result, the inspection of this issue could not be completed during the E&TS inspection.
E3 Engineering Procedures and Documentation E3.*
Condition Reports a.
Insoection Scope (IP37550: IP37700)
The inspectors performed selected reviews of CRs to evaluate the quality of the licensee's apparent cause and extent of condition evaluations as well as the adequacy and timeliness of associated corrective actions.
b.
Observations and Findinos For most :: elected CRs, the inspectors determined that the apparent cause and extent of condition evaluations were adequate, and the associated corrective actions were timely and of sufficient scope. Some problems were identified with instrument drift trending and the resolution of a previously identified violation (50-461/95003-02) concerning inadequate evaluation of heat exchanger test results, as discussed in Sections E3.3 and E8.2, respectively.
The threshold for writing CRs was considered adequate and the problems identified in the CRs reviewed were well documented. The actions described in the CRs to resolve the problems appeared to be adequate and no problems or cencerns were noted with the selected CRs reviewed. In addition, based on their system reeiews, the inspectors did not identify any problems that had not been previously identified through the licensee's CR program.
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Conclusions in reost instances, the corrective action process for the CRs selected for review was adequately implemented and resulted in acceptable corrective actions.
E3.2 Post /Stamo Affixed Procram
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Insoection Scooe (IP37550: IP37700)
The team assessed the use of the post / stamp affixed change program and the potential impact of not using post / stamp affixed drawings on the CPS hardware changes during the SDFV and SSTR review process.
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b.
Observations and Findinas The licensee implemented a drawing change program called the post / stamp affixed
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program to identify minor equipment variations (e.g., arc strikes, grind marks, etc. on installed equipment) that were not incorporated into revised plant drawings. The minor variations were to be identified / recorded against the drawing number in the licensee's Document Status System. The intent of the program was to exclude minor variations from controlled drawings such that the minor variations would not require consideration during tag-outs or other work activities. However, the licensee determined that the program was not implemented properly in that the potential existed for the licensee to apply the program to equipment variations that were not just minor. As a result, drawing revisions might not have been initiated for more significant equipment changes. The
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licensee's corrective action plan to resolve the identified concern was described in CR 1-98-06-0186," Lack of Understanding Post / Stamp Affixed Change Documents,"
Revision 0, and the plan was approved on September 29,1998.
During the inspectors' review of this issue and discussions with the licensea, the inspectors determined that the licensee's corrective action plan would preclude treating future equipnent changes that required drawing revisions within the post / stamp affixed program. The team did not identify any new problems associated with the post / stamp affixed drawing program that had not been previously identified by the licensee.
c.
Conclusions The team concluded that not using post / stamp affixed drawings during specific engineering review processes had no major impact on the final engineering work products.
E3.3 Surveillance Procedures a.
Inspection Scoce (IP37550: IP37700)
The inspectors reviewed several recently completed surveillance procedures for adequacy of content and implementation related to the main steam and instrument air (IA) systems containing ADS instrumentation. Procedures reviewed included, but were not limited to, CPS 9430.30 "NSPS Untested Island / Calibration 1-999 Second Time Delay," dated August 22,1998; CPS 9433.03 "ECCS Reactor Water Level B21-N091 A Channel Calibration," dated August 31,1998, and; CPS 9433.10 "ECCS Drywell Pressure B21-N094A(E) Channel Calibration," dated June 26,1997.
b.
Observations and Findinas The surveillance procedures reviewed were of adequate quality and were implemented properly. However, some Indications of instrument drift were noted during the team's l
review of the completed surveillance data sheets. The as-found data for some
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safety-related instruments, specifically Drywell Pressure Cwitch B21-N094A and Reactor
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Water Level Channels B21-N691E, B21-N692A, and B21-695A were outside the specified acceptance criteria.
Although the as-left calibrations of these instruments were within acceptable tolerances, the inspectors were concerned with the observed instrument drifts and inquired whether the licensee was performing instrument drift trending. This issue was discussed with the plant control and instrumentation (C&l) engineers and the maintenance supervisor. The licensee informed the team that the plant did not have an instrument setpoint control program or an instrument performance trending program. The licensee stated that the Reliability Group had trended some instrumentation until 1995; however, the trending was discontinued when the group was disbanded. Lack of an adequate instrument setpoint control program and lack of trending of safety-related instrument drift was considered a weakness.
The licensee also stated that supporting calculations for several important instrument setpoints were either inadequate or missing. The licensee's proposed corrective actions included issuing implementing procedures for a setpoint control program and the preparation / revision of necessary setpoint calculations. Subsequent to the team's discussion concerning these issues, the licensee issued an instrument setpoint program action plan on September 30,1998, as discussed in Section E1.5.
c.
Conclusions The team concluded that the lack of a setpoint control program and a lack of trending of safety-related and maintenance rule-related instrument drift was a weakness. The team also concluded that lack of supporting calculations for important instrument setpoints was a weakness.
E3.4 System Health Reoort a.
Insoection Scope (IP37550)
The team reviewed the licensee's " System Health Report" (SHR) and discussed the report's rating criteria / system status with cognizant licensee personnel to determine if the information contained in the SHR reflected an accurate status of the monitored systems.
b.
Observations and Findinas The SHR was generated in response to the licensee's " Plan for Excellence" to provide a comprehensive representation of the material condition of the monitored systems. The SHR used a color coding scheme to communicate system status and identify problem areas. The color coding scheme was based on a rating criteria for six categories:
performance; operator work arounds and main control room deficiencies; configuration management; CRs; maintenance backlog; and physical condition. The color coding scheme was: Green - Excellent; Yellow - Acceptable; Orange - Degraded; and Red - Not Acceptable. The team evaluated the SHR status for the ADS, DC and SX systems.
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The SHR rated the DC system excellent (green) for operator work arounds and main control room deficiencies and CRs; acceptable (yellow) for configuration management and physical condition; degraded (orange) for system performance; and not acceptable (red) for maintenance backlog.. Contributors to the orange performance rating were
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potential degradation of molded case circuit breakers, lack of spare parts for battery chargers, and the lack of maintenance personnel training on battery charger
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requirements. The red performance rating was based on the large number of open
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MWRs greater than 18 months old (37% of total open).
The SHR rated the SX system excellent (green) for performance, operator work arounds and main control room deficiencies, and configuration management; degraded (orange)
for maintenance backlogs and physical condition; and not acceptable (red) for CRs.
Contributors to the orange performance rating were chronic valve leakage, excessive flow through the RHR heat exchanger bypass line, and a large number of old open MWRs. The red performance rating was based on the large number of old open CRs.
c.
Conclusions Based on their system reviews, the team concluded that the SHR provided an accurate accounting of system status with regards to the numbers of CRs, MWRs, etc. No major
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discrepancies were identi'ied with the SHR for the systems reviewed. The ADS system status could not be reviewed since the system was not addressed in the SHR.
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E.5 Engineerdng Staff Training and Qualifications E5.1 System Enaineerina a.
Inspection Scoce (IP37550)
The team interviewed selected system engineers and reviewed their qualifications and training.
b.
Observations and Findinas The inspectors interviewed two system engineers responsible for the ADS system. The
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mechanical system engineer had responsibility for the ADS system for about two years and was also responsible for 10 other small systems. This system engineer had not received any formal system training on the systems assigned to him. However, the system engineer appeared qualified and experienced. The electrical system engineer assigned to the ADS system was a contractor. Although this engineer was experienced at other plants, he had not received any training for the systems assigned to him. The licensee stated that the lack of adequate training for the system engineers was identified during self-assessments and that corrective actions had been implemented.
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Conclusions The team concluded that the system engineers were generally qualified and experienced. However, the team identified a weakness in that detailed training was not provided to the system engineers for their assigned systems. On October 10,1998, the i
team was notified in Letter Y-107159, dated October 5,1998, that system engineers
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would receive mandatory senior reactor operator system training for their assigned system (s).
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E8 Miscellaneous Engineering issues (92701 and 92702)
The inspoctors reviewed actions taken by the licensee to address issues identified in LERs and in previous NRC inspection reports. The results of this followup effort are i
discussed in this section.
E8.1
[ Closed) Unresolved item 50-461/93003-016: NRC review of licensee actions to address incorrect grid voltages as documented in CR 1-92-04-031. The licensee developed calculations 19-AJ-70,-71,-72 to establish minimum pickup and dropout voltages at motor control centers. The results of the three calculations formed the basis for a proposed license amendment to change the degraded voltage values in Technical Specification (TS), Table 3.3.8.1-1. The NRC issued license amendment No.110 documenting the acceptability of the current degraded values. This item is closed.
E8.2 (Closed) Notice of Violation 50-461/95003-02: Failure to evaluate marginal diesel generator SX heat exchanger test results. Test results for four of the five diesel generator SX heat exchangers were in the " alert" range and engineering evaluations had not been performed. The engineering evaluations were required to verify that the heat exchangers would transfer the design basis heat loads to the ultimate heat sink. In addition CPS procedures required that engineering evaluations be performed when heat exchanger test results were in the alert range. The licensee's response to the associated violation, dated July 13,1995, stated that an engineering computer analysis was performed for the respective heat exchangers and that "... this analysis determined that the heat exchangers would have been able to adequately transfer decign basis heat loads to the ultimate heat sink."
During performance of the 1997 review of the GL 89-13 program, licensee personnel questioned the adequacy of the response to this violation since the computer analysis was missing and could not be found. Condition Report 1-97-10-023," inadequate Responses to NRC NOV, EA Assessment 95E, and NAD Audit 038-94-16," dated October 1,1997, was written to address this problem. Design engineering personnel stated that it was "... not cost effective to repeat the computer analysis simply to have it on file." Other statements in the CR referred to Calculation 0-65-017-PCC-02,
" Evaluation of Diesel Generator Heat Exchanger Performance Data from 1990 to 1997,"
dated February 7,1998, which addressed the test results and uncertainty of the results for all the emergency diesel generator (EDG) heat exchanger tests performed at CPS.
The CR further stated that "... based on the calculation (0-65-017-PCC-02], the test results of one of the five tests is below the minimum acceptable design basis heat
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removal capability. If the test uncertainty band is considered, none of the five tests can be assured of exceeding design basis heat removal capability." This information was different from the information provided to the NRC in the violation response letter. Since the computer analysis documentation used to support the licensee's response to the violation was missing, the analysis could not be compared with the calculation to j
reconcile the difference.
The CR was closed on April 21,1998, indicating that the licensee did not intend to further pursue this issue. However, NRC and licensee records did not indicate that the violation j
had been closed by the NRC. Furthermore, the CR did not describe how the licensee had addressed the EDG heat exchanger problems revealed in the calculation. During inspector discussions with the licensee, the licensee agreed that CR 1-97-10-023 was closed improperly and that a supplemental response to the violation would be submitted.
A supplemental response to the violation was received on October 26,1998, and was i
considered acceptable.
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Based upon the SDFV review conducted for the SX system, licensee actions had been identified to correct the noted problems with the EDG and other SX heat exchangers.
The inspector's review of licensee records indicated that two of the EDG heat exchangers had been replaced and rework of the other three was in-progress. Licensee personnel stated that work on all five heat exchangers would be completed and the heat exchangers would be satisfactorily tested prior to plant restart. The inspector had no further concerns in this area and this item is closed.
E8.3 (Closed) Notice of Violation 50-461/97003-03: Failure to perform adequate design reviews prior to installation of Auxiliary Power (AP) Modification AP-028. The design basis for modification AP-028 was not correctly translated into design drawings.
Adequate design reviews were not performed to determine the suitability of a regulating transformer installed per modification LD-028 and an unauthorized modification to Division 2, Bus 1B1. Specifically, an uninterruptable power supply was added to the bus without any design control measures being implemented.
The licensee's CR dated January 7,1997, and subsequent detailed root cause analysis identified we'aknesses in a number of areas including the engineering staff's understanding of industry issues with microprocessors, the interface between the design change process and plant testing, and job skills.
The licensee's corrective action for this issue involved removing the regulating transformers from service and installing suitable transformers under Modifications AP-33, 34 and 35. Training was also provided to licensee engineering personnel on design l
considerations for microprocessor controlled equipment. This item is closed.
l E8.4 (Closed) Notice of Violation 50-461/97003-04: Failure to complete Clinton Procedure l
1401.01F006, " CAT 'A' Instrument Failure Checklist," following completion of the TS l
surveillance test on the Off Gas Hydrogen Analyzer Channel IN66-N012A, where two as-found data points for recorder IN66-R605 were not within the specified limits.
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Technicians did not follow the procedure step that required the initiation of the instrument failure checklist.
The licensee initiated CR 1-97-05-261 to resolve the failure to properly follow procedure 1406.01F006 and to prompt the identification of any additionalinstances of the failure to follow the procedure. The inspectors reviewed documentation for attendance at a seminar presented to instrument technicians on complying with procedures. Based on a review of procedure evaluations initiated for instrument failures, the inspectors noted an increase in evaluations since the seminar was presented. This item is closed.
E8.5 (Closed) Notice of Violation 50-461/97003-05: Failure to properly conduct a 10 CFR 50.59 safety evaluation for changes in operation of plant equipment and the installation of insulation on containment isolation system piping. The evaluation did not adequately justify that the changes did not involve a USQ.
The licensee's response to this violation was comprehensive. New safety evaluations were performed thatjustified the determination that the changes did not involve USQs.
Based on their review of these evaluations, the inspectors did not identify any discrepancies. Both parts a. and b. of this violation are closed.
E8.6 (Closed) Notice of Violation 50-461/97003-06: Failure to submit a report pursuant to 10 CFR 50.73(a)(2)(ii) within 30 days. The licensee initiated CR 1-96-10-360-0 upon identifying that 21 containment penetrations were susceptible to significant overpressure during a LOCA.
The licensee reported the potential overpressure condition pursuant to 10 CFR 50.73 on March 13,1997. Personnelinvolved in reporting decisions have been trained on the reporting requirements involving ASME piping issues. This item is closed.
E8.7 (Closed) LER 50-461/97-016-00: Failure to provide emergency lighting for safe shutdown equipment as required by the plant design basis. Emergency lighting was not provided as required by 10 CFR Part 50, Appendix R, to reposition va!ve 1E12-F024B during a hot short of the cables associated with valve 1E12-F0068. The licensee's corrective actions for the hot short concern included de-energizing valve 1E12-F006B to ensure that it would not open during the hot short condition as discussed in Section E1.1.
Additional corrective actions resulted in the emergency light no longer being required.
The inspector determined that the corrective actions were acceptable. This LER is closed.
E8.8 (Closed) LER 461/97-026-00.-01: Inadequate procedure for inspection of shutdown service water pumps for excess silt resulted in shutdown service water pump inoperability. This ! ER was written when the excessive accumulation of silt in the SX pump intake area caused the SX pumps to be declared inoperable. The licensee's actions to address this issue included determination of maximum allowed silt accumulation levels in the SX pump intake bay, removal of excessive sitt from the intake area, and establishment of a periodic silt level inspection requirement for the area as
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discussed in Section E1.1. Based on these actions, LERs 97-026-00 and 97-026-01 are closed.
E8.9 (Closed) LER 50-461/97-027-00: Misinterpretation of 10 CFR 50.59 resulted in installation of a modification to the RHR system that placed the plant in a condition that was outside of the plant's design basis. This LER was issued for a 10 CFR 50.59 evaluation that failed to evaluate the addition of relay interlocks to the RHR system during the modification process when there was a potential for an increase in the probability of equipment failure. These relays were removed during modification RH-48 as discussed in Section E1.1, so they no longer affect the probability of failure of this i
system.
i The licensee's corrective actions included a change in the safety evaluation procedure to provide additional guidance pertaining to an increase in the probability of an equipment malfunction. The inspectors verified that the corrective actions were properly implemented. This LER is closed.
I E8.10 (Closed) LER 50-461/97-035-01: Division 1 and 2 battery chargers incapable of supplying full rated voltage and current flow at degraded voltage trip setpoint in accordance with the plant's TS and design basis.
During investigation of voltage margins and a cable impedance deficiency, the licensee found that the Divisions 1 and 2 safety-related battery chargers had not been included in the degraded voltage calculations. The licensee determined that this condition had existed since initial startup.
This event was caused by a failure to include the battery charger's minimum voltage requirement in the acceptance criteria for the degraded voltage transient calculation. A design change was issued to adjust the tap settings on the Division 1 and 2 battery charger transformers to assure that the minimum voltage requirement was met. The inspectors did not identify any discrepancies based on their review of appropriate calculations, design changes and corrective action work documents.
Criterion ill of 10 CFR Part 50 requires that measures be established to assure that the design control measures provide for verifying or checking the adequacy of design. This non-repetitive, licensee-identified and corrected violation is an NCV, consistent with j
Section Vll.B.1 of the NRC Enforcement Policy (NCV 50-461/98019-03(DRS)). This LER is closed.
I E8.11 (Closed) Insoection Fe!bwuo item 461/97999-20: Possible inadequate implementation of GL 89-13 requirements for the shutdown service water system. The inspectors reviewed completed licensee assessments and evaluations of the SX system. This included the SDFV review of the SX system. The SDFV review was based on the requirements of GL 89-13 and the review appeared to be thorough and comprehensive, A number of problems were identified. Actions to resolve many of these problems had been completed and others were in-progress. The inspectors had no further concerns in I
this area. This item is closed.
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E8.12 (Closed) LER 50-461/98-001-00: Failure of Division 2 safety-related battery charger due to deficient supplier soldered connections. A low voltage alarm on the Division 2 DC bus was received in the control room. Other indications included zero ampere output from the battery charger followed by the DC bus voltage stabilizing at about 128 Volts which was the expected voltage for a loss of the battery charger. The charger appeared to stop and start itself electronically, i.e., the DC output voltage and current decreased to zero and then returned to normal.
Based on followup troubleshooting, the licensee attributed the cause of the failure to deficient supplier workmanship in soldered wire connections at transformer T1 A. The licensee confirmed this theory with a test on the plant's training battery charger. In addition, a broken wire was discovered at transformer T1 A.
The licensee's corrective action for this issue included resoldering the loose connections on the silicon controlled rectifier (SCR) firing cards, repairing the broken wire at transformer T1 A, and resoldering the wire connection at fuse F-7 in the Division 2 battery charger. Also, the other safety-related battery chargers from the same manufacturer had been inspected and necessary repairs effected. Based on a review of completed MWRs, documenting the inspection and repair activities, this LER is closed.
E8.13 (Closed) LER 50-461/98-004-00.-01: Division 2 nuclear systems protection system (NSPS) inverter not in accordance with the plant's design basis due to various deficiencies. During troubleshooting of spurious transfers of the Division 2 NSPS inverter to its 120 Volt alternate ac source, technicians discovered various deficiencies in the SCR, power diodes, a resistor, and improperly soldered connections.
The licensee, through a root cause investigation, determined that the cause was ineffective and inadequate preventive maintenance activities. Contributing factors were incomplete and incorrect calibration procedures and inadequate training of maintenance personnel on specific techniques regarding maintenance of various internal components.
The licensee replaced all defective components discovered during troubleshooting.
Connections were inspected, tightened, and tested. Calibration procedures were enhanced and equipment recalibrated to the revised procedures.
The inverter was tested under the supervision of a vendor representative and returned to service. No spurious transfers have been experienced since the inverter was returned to service. The licensee inspected the Division 1,3, and 4 inverters and the A and B NSPS solenoid inverters.
Training was provided to maintenance personnel on procedures and techniques for soldering, proper installation of power semi-conductors, and the proper application of heat transfer compound. Licensee Event Reports 98-004-00 and 98-004-01 are closed.
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M.2 Material Condition of Facilities and Equipment M2.1 Plant Walkdowns a.
Inspection Scoce (IP37700)
The inspectors walked down the IA system, as it relates to the ADS, AP and DC
systems, and the SX system to assess the material condition.
b.
Observations and Findinas The inspectors walked down the lA system and some components of the ADS system with the system engineer. No problems were identified. Based on a visualinspection of plant locations where the degraded voltage modification activities were in-progress, the l
inspectors determined that the cornponents and systems were being installed according
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to approved design drawings. The installation of the degraded voltage modification was l
not yet complete so PMT had not been accomplished. The team concluded that the degraded voltage modification was being installed according to the approved design.
Appropriate design, engineering, and management reviews were performed.
c.
Conclusions j
The material condition of the walked down systems appeared to be good. The system engineers appeared to be knowledgeable of the systems.
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V. Management Meetings XI Exit Meeting Summary l
The inspectors presented the inspection resu!is to members of licensee management at the j
conclusion of the inspection on October 9,1998. The licensee acknowledged the findings presented. The inspectors questioned the licensee as to the potential for proprietary information
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being included or retained in the inspection report as discussed at the exit. No proprietary information was identified.
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PARTIAL LIST OF PERSONS CONTACTED K. Baker, Director - Support Engineering J. Barron, Director-Plant Engineering R. Bhat, Supervisor - Fire Protection Engineering D. Busham, Supervisor - Quality W. Carsky, Director - Design Engineering V. Cwietniewicz, Manager - Maintenance L. Demick, Chairman - Senior Engineering Review Group R. Ebright, Project Manager - System Design and Functional Verification j
K. Graf, Project Manager - Nuclear Station Engineering Department (NSED)
J. Gruber, Director - Corrective Action J. Hanson, Director - Nuclear Training A. Haumann, Supewisor - Design Engineering B. Haynes, Project Manager - Setpoint Program W. Helenthal, Supervisor - Maintenance Planning (C&l)
G. Hunger, Manager - Clinton Power Station S. Lakebrink, Supervisor - Design Engineering W. MacFarland IV - Chief Nuclear Officer R. Maher, Supervisor - Plant Engineering W. Manganaro, Project Manager - Project Engineering P. Marcum, Supervisor - C&l Design M. Norris, Supervisor-Engineering Assurance E. Patel, Director - Project Engineering R. Phares, Manager - Nuclear Safety and Performance Improvement W. Romberg, Manager-NSED T. Roe, Manager - Maintenance Direct Support E. Schweitzer, Supervisor - NSSS Systems J. Sipek, Director-Licensing
' M. Stickney, Supewisor-Regional - Licensing M. Wyatt, Manager - Recovery 23 *
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t INSPECTION PROCEDURES USED IP 37001 10 CFR 50.59 Safety Evaluation Program IP 37550 Engineering IP 37700 Design Changes and Modifications IP 92701 Followup IP 92702 Followup on Corrective Actions for Violations and Deviations i
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ITEMS OPENED, CLOSED OR DISCUSSED Opened 50-461/98019-01 VIO Enforcement Discretion per Vll.B.2: Loss of Suppression
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Pool Cooling 50-461/98019-02 NCV Failure to Take Adequate and Timely Action to Correct the SX Pump Intake Silting Problem 50-461/98019-03 NCV Division 1 and 2 Battery Chargers Incapable of Supplying Full Rated Voltage and Current Flow at Degraded Voltage Trip Setpoint Closed 50-461/93003-01A URI NRC Review of Licensee Actions to Address incorrect Grid Voltages as Documented in CR 1-92-04-031 50-461/95003-02 VIO Failure to Evaluate Marginal Diesel Generator SX Heat Exchanger Test Results 50-461/97003-03 VIO Failure to perform adequate design reviews prior to installation of Auxiliary Power (AP) Modification AP-028 50-461/97003-04 VIO Failure to Complete Clinton Procedure 1401.01F006, " CAT
'A' Instrument Failure Checklist," Following Completion of Technical Specification Surveillance Test 50-461/97003-05 VIO Failure to Properly Conduct a 10 CFR 50.59 Safety Evaluation for Changes in Operation of Plant Equipment
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50-461/97003-06 VIO Faiiure to Submit a Report Pursuant to 10 CFR 50.73(a)(2)(ii)
Within 30 days 50-461/97-016-00 LER Failure to Provide Emergency Lighting for Safe Shutdown Equipment as Required by the Plant Design Basis 50-461/97-026-00,-01 LER Inadequate Procedure for Inspaction of Shutdown Service Water Pumps for Excess Silt Results in Shutdown Service Water Pump Inoperability 50-461/97-027-00 LER Misinterpretation of 10 CFR 50.59 Results in Installation of Modification of the Residual Heat Removal (RHR) That is Outside of the Plant's Design Basis
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50-461/97-035-00,-01 LER Division 1 and 2 Battery Chargers Incapable of Supplying Full Rated Voltage and Current Flow at the Degraded Voltage Trip Setpoint in Accordance with the Plant's j
Technical Specifications and Design Basis
50-461/97999-20 IFl Possible inadequate implementation of GL 89-13 Requirements for the Shutdown Service Water System 50-461/98-001-00 LER Failure of Division 2 Safety-Related Battery Charger Due to Deficient Supplier Soldered Connections
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50-461/98-004-00,-01 LER Division 2 Nuclear Systems Protection System Inverter Not in Accordance with the Plant's Design Basis Due to
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Various Deficiencies 50-461/98019-01 VIO Enforcement Discretion per Vll.B.2: Loss of Suppression
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50-461/98019-02 NCV Failure to Take Adequate and Timely Action to Correct the l
SX Pump Intake Silting Problem l
50-461/98019-03 NCV Division 1 and 2 Battery Chargers Incapable of Supplying Full Rated Voltage and Current Flow at Degraded Voltage
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Trip Setpoint Discussed i
No items identified during previous inspections were reviewed and discussed without being closed during this inspection.
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~ LIST OF ACRONYMS USED i
AP Auxiliary Power
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ADS Automatic Depressurization System C&l Control and Instrumentation CFR Code of Federal Regulations CPS Clinton Power Station CR-Condition Report DC Direct Current DDR Detailed Design Review DRP Division of Reactor Projects
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DRS Division of Reactor Safety 1 L
E&TS Engineering and Technical Support ECCS Emergency Core Cooling System ECN Engineering Change Notice EDG.
' Emergency Diesel Generator EQ Environmental Qualification.
ERAT Emergency Reserve Auxiliary Transformer GL Generic Letter HPCS High Pressure Core Spray IA Instrument Air IP inspection Procedure ISA
. Integrated Safety Assessment-LER-Licensee Event Report
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. LOCA-Loss of Coolant Accident MVAR Mega-Volt Ampere Reactive
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Maintenance Work Request NCV Non-Cited Violation
. NRC Nuclear Regulatory Commission NRR Nuclear Reactor Regulation
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NSED Nuclear Station Engineering Department PM Preventative Maintenance PMT._
. Post-Modification Testing QA Quality Assurance RAT Reserve Auxiliary Transformer
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'SCR Silicon Controlled Rectifier SDFV System Design and Functional Verification e
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System Health Report I
- SPC'
- Suppression Pool Cooling SSTR-System Surveillance Test Review SVC Static VAR Compensators SX
, Shutdown Service Water System
.. TS Technical Specifications
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USAR Updated Safety Analysis Report-USQ
.Unreviewed Safety Question
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PARTIAL LIST OF DOCUMENTS REVIEWED The following is a list of licensee documents reviewed during the inspection, including documents prepared by others for th3 licensee, inclusion on this list does not imply that NRC inspectors reviewed the documents in their entirety, l
but rather that portions or selected portions of the documents were evaluated as part of the overall inspection effort.
NRC acceptance of the documents or any portion thereof is not implied.
CPS Document Revision /
Description Number Date
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_ Calculation 011A10 Calculation
Calculation 01RH16 Calculation
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j Calculation 01RH23 Calculation
Calculation 01RH26 Calculation
l Calculation 01RH42 Calculation
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Calculation 01SX32 SX Cooling Water Requirement for OPR13A
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Calculation 0-65-017-PCC-02 Evaluation of Diesel Generator Heat Exchanger Performance Data 02/07/98 from 1990 to 1997 Calculation 19-AK-13 Calculation
Calculation 19-AJ-70 Calculation Calculation 19-AJ-71 Calculation
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Calculation 19-AJ-72 Calculation l
Calculation DC-ME-09-CP Calculation
l Calculation EMD-021930 Calculation 2-L Calculation EO-A-9 Radiation Qualification Dose for Equipment in Auxiliary Building
l Emergency Core Cooling System (ECCS) Cubicles
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Calculation EQ-A-12 Calculation
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L Calculation IP-M-0227 Calculation
Calculation IP-S-0132 Acceptance Criteria for Allowable Sediment Dept (Siltation) in the CW Screenhouse Calculation ISX-22A Calculation CR 1-95-06-036 Incomplete EDG Heat Exchanger Test Results, Evaluation / Review l
CR 1-97-06-181 Inadequate 8 Hour Battery Powered Emergency Lighting 06/17/97 i
CR 1-97-09-058 ISA - OBS #1997-0692 A Large Number of PM Tasks are Post Due 09/05/97 j.
or Late, Many Without Approved Deferral Requests CR 1-97-09-331 Division 2 Diesel Generator Heat Exchanger Overdue PMs CR 1-97-09-344 ISA: Inspection of the SX/FP Pump Intake Bays for Silt and Clams l
l CR 1-97-10-023 Inadequate Responses to NRC NOV, EA Assessment 95E, and NAD 10/01/97
Audit 038-94-16 CR 1-97-10-054 Indeterminate Condition Because of Silt in the SX/ Unit 1 FP Bay Area 10/03/97 of the Screen House
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t CPS Document Raision/
Number Description Date CR 1-97-10-131 Safety Evaluation for Modification RHF011 Did Not Fully Evaluate 10/10/97 Loss of Suppression Pool Cooling CR 1-97-11-021 Weakness in GL 89-13 (Safety Related Heat Exchangers) Program CR 1-97-11-368 GL 89-13 Heat Exchanger Test Program Deficiency 11/19/97 CR 1-97-12-333 Suppression Pool Cooling Disabled 12131/97 CR 1-98-01-165 Potential Ice Buildup at Screen House intake Due to Cold Ambie.it Temperature CR 1-98-02-457 Non-Existant Remote Shutdown Function Periodically Tested (SDFV)
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CR 1-98-03-022 Failure to Periodically Test Strainer Backwash Valves 1SX013D/E/F 03/06/98 (SDFV)
CR 1-98-03-023 SX Auto Backwash on High Dp Not Periodically Verified (SDFV)
03/06/98 CR 1-98-03-571 Inadequate Design Review 03/30/98 CR 1-98-05-081 SX Heat Exchanger Minimum Flows Were Changed Without 05/13/98 Discussion in the 50.59 Safety Evaluation CR 1-98-05-145 Inadequate Engineering Evaluations May Lead to ineffective 05/15/98 Corrective Actions CR 1-98-05-308 No Controlled Calc to Determine RCS Vol. to Calculate Reg, Boron 05/29/98 Wt. to Achieve Correct Concentration CR 1-98-05-309 No Pressure Drop Calculation Calc for SLC System 06/01/98 CR 1-98-06-036 Calculation identification and Approval 06/03/98 CR 1-98-06-132 RHR Bypass in Excess of 8000 GPM 06/16/98 CR 1-98-06-186 Lack of Understanding Post / Stamp Affixed Change Documents 06/15/98 CR 1-98-06-191 Discrepancy Between Setpoint for 1RIX-PRO 34 and Cal:ulation 06/16/98 PR-27 CR 1-98-06-302 Inadequate Control of Design Basis Calculations 06/25/98 CR 1-98-07-023 Reactor Hi Pressure Scram Setpoint Outside Design Spec Data 07/06/98 Sheet Value CR 1-98-07-288 Possible Tagging Program impacts From Lack of Understanding of 07/23/98 Post / Stamp Affixed Change Documents CR 1-98-07-303 Uncontrolled Calcs Used to Change Plant Instrument Setpoints for 07/24/98 ECCS Instrumentation CR 1-98-07-308 Adverse Trend Identified - Setpoint Calculations Not Performed or 07/24/98 Controlled in a Quality Fashion CR 1-98-08-150 Procedure inadequacy to Address Deficiencies in Instruments Used 08/14/98 to Satisfy Surveillance Requirements CR 1-98-08-206 Missed Design impact May Affect RHR-B Suppression Pool Cooling 08/18/98 CR 1-98-09-201 Division 1 SX Flow Balance Low Flows to Safety Related Components Fed by SX
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CPS Document wision/
Number Description Date CR 1-98-10-084 Failure to identify CR 1-98-08-206 as an Event Reportable under 10 10/06/98 CFR 50.73 CR 3-97-07-198 NSED Engineering Support Training Program Description (TPD)
07/18/97 Non-Compliance ECN 27675 Replacement of Target Rock Valves ilA-044A and 11A-044B 02/09/94 ECN 28365 increase of Stroke Time of Valve ilA-013B 09/16/94 ECN 30019 Utilize Spare Contacts on Remote Shutdown Panel Switch 1C61-HS510 to isolate the Electrical Return Path for the Green and Red Lights ECN 30211 Install Annunciator Bypass Switch ECN 30225 Revision of ADS Instrument Air Header Low Pressure Alarm Setpoint 07/11/97 ECN 30430 Revise Design for High Voltage Shutdown Card
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ECN 30445 Revise Setpoint for High Voltage Shutdown 1DC08E ECN 30492 Design Changes to Control Circuits of Motor Operated Valves (MOVs) Modification RH-048, Residual Hear Removal (RHR) Pump A
& B Suction Valve Interlocks ECN 30660 Transformer Tap Change on 1DC06E & 1DC07E ECN 30669 Transformer Tap Change on 1DC07E ECN 30708 Motor Replacement for the 1SX014B Valve Operator ECN 30867 Eliminate Pressure Locking on 1E12-F028A ECN 30903 Installation of a Restrictive Orifice in the SX/RHR Heat Exchanger
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Bypass Line to Reduce Bypass Flow ECN 30928 Installation Cleanout Connections on SX and WS Piping to Support Required Flow Balancing ECN 30992 Division 4 Battery Changer Tap Change EO-CLO23 Tab D Environmental Qualification of 480V Indoor Unit Substation
EQ-CL-041 (Vol 1 of 4), Tab 1 ITT Barton Qualification Test Report No. R3-580A-9 for ITT Barton 580A Series Differential Pressure Switches Gen:ric Letter 89-13 Service Water System Problems Affecting Safety-Related Equipment 07/18/89 LER 97-016-00 Failure to Provide Emergency Lighting for Safe Shutdown Equipment 07/16/97 as Required by the Plant Design Basis LER 97-025-00 Design Deficiency Results in plant Being Outside Design Basis for a 10/28/97 Fire in the Main Control Room Potentially Damaging Valves Required for Safe Shutdown of the Plant LER 97-026-00,-01 Inadequate Procedure for inspection of Shutdown Service Water 11/13/97 Pumps for Excess Silt Results in Shutdown Service Water Pump 03/03/98 Inoperability LER 97-027-00 Misinterpretation of 10 CFR 50.59 Results in Installation of 12/04/97 Modification of the Residual Heat Removal (RHR) That is Outside of the Plant's Design Basis
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CPS Document Revisioni Number Description Date LER 97-035-00,-01 Division 1 and 2 Battery Chargers Incapable of Supplying Full Rated 01/16/98
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Voltage and Current Flow at the Degraded Voltage Trip Setpoint in 03/19/98 Accordance with the Plant's Technical Specifications and Design Basis LER 98-001-00 Failure of Division 2 Safety-related Battery Charger Due to Deficient 02/17/98 Supplier Soldered Connections LER 98-004-00,-01 Division 2 Nuclear Systems Protection System Inverter Not in 02/26/98 l
Accordance with the Plant's Design Basis Due to Various 05/28/98 Deficiencies
)
l LER 98-016-00 Failure to Test Valves 1SX013D/E/F in Accordance with the in-06/11/98 service Testing Program Due to Personnel Error Lett:r (GE) from L.H. Larson to CPS, Unit 1 Setpoint Methodology Program Attachments contain 01/23/97
J.H. Greene calculations for LPCS, LPCI, ADS Bypass Timer and initiation Timer, High Drywell Pressure, RV Water Level 3, Level 1 Water Level Proprietary Information Letter (GE) from L.H. Larson to Contains applicable portions of 22AS462
J.H. Greene (Attachment)
Proprietary information Lett:r (GE) from L.H. Larson to Contains applicable portions of 22A3139AN
J.H. Greene (Attachment)
Proprietary Information Lett:r (GE) from L.H. Larson to Contains applicable portions of 22A4622
J.H. Greene (Attachment)
Proprietary Information Lett:r (GE) from L.H. Larson to Contains applicable portions of 22A4622AV
J.H. Greene (Attachment)
Proprietary Information Lett:r (GE) from L.H. Larson to Contains applicable portions of NEDC 31336 CLASS Ill GE October J.H. Greene (Attachment)
Instrument Setpoint Methodology 1986 Proprietary Information Lettsr U-602554 Proposed Amendment of Facility Operating License No. NPF-62 02/22/96 (LS-94-013)
Lett:r U-602613 Additional Info for Proposed Amendment of Facility Operating 07/24/96 License No. NPF-62(LS,-94-013)
Lett:r U-602836 USAR Submittal, Rev 7 10/20/97
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j Letter Y-107159 Mandatory SRO System Training 10/05/98 Lett:r Y-108025 Engineering Support Review of NSED Pil Surveys 04/23/98 Proprietary information Lett:r Y-108084 SDFV Project 1998 - Final Report 06/05/98 Lett:r Y-217517 ADS SSFA 03/29/96 M01-1600 Sheet 9 Drawing A
M05-1040 Sheet 7 Drawing AC M05-1052 Sheet 1 Shutdown Service Water (SX) P&lD Drawing AL M05-1052 Sheet 2 Shutdown Service Water (SX) P&lD Drawing AD M05-1052 Sheet 3 Shutdown Service Water (SX) P&lD Drawing AE
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CPS Document Revision /
Number Description Date M05-1052 Sheet 4 Shutdown Service Water (SX) P&lD Drawing P
M05-1052 Sheet 5 Shutdown Service Water (SX) P&lD Drawing Z
Modification 11A-021 Reposition Valve Limit and Torque Switches for Valves ilA-012A
and 11A-013A Modification AP-033 Replace Existing 480-208/120V/15KVA/3 Phase Non-Regulating Distribution Transformer in Division 1 Motor Modification AP-034 Replace Existing 480-208/120V/15KVA/3 Phase Non-Regulating Distribution Transformer in Division 2 Motor Modification AP-035 Replace Existing 480-208/120V/15KVA/3 Phase Non-Regulating Distribution Transformer in Division 3 Motor Modification AP-036 Install Foundations SVC's for Both RAT & ERAT 4kV Bus Source Modification AP-037 Install SVC Unit, Connected to 4kV Bus Near RAT Transformer Modification RH-047 Modification of Various MOV Circuits for Hot Shorts Modification RH-048 Residual Hear Removal (RHR) Pump A & B Suction Valve Interlocks Modification RHF-011 Interlock Installation MWR D73800 MWR D73800 (for ECN 30211)
MWR D82191 install Division 1 Portion of RH-048 Modification NSED instruction DE-12 Instruction for Marking (Redline /Greenline) Control Room Drawings
to Reflect Plant Configuration Changes NSED instruction ER-01 Engineering Product Review (EPR) Project Plan
NSED Instruction ER-2 Technical Specification Required System Surveillance Test Review
(SSTR) Plan NSED Procedure A.04 Nuclear Station Engineering Department Organization
NSED Procedure D.43 Incorporating Design Change Documents into Engineering
Documents NSED Procedure D.46 Request for Miscellaneous Drafting Services
NSED Procedure E.1 Calculations
NSED Procedure W.01 Engineering Work Request
NSED Self-Assessment NE-98-05 Control of Software-Hardware Change Process 05/29/98 NSED Self-Assessment NE-98-06 Partial Release of Modifications 06/29/98 NSED Self-Assessment NE-98-08 Review of Organizational Processes
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07/09/98 NSED Self-Assessment NE-98-13 Engineering Support Training Program 05/22/98 NSED Self-Assessment NE-98-15 Temporary Modifications 06/30/98 NSED Self-Assessment NE-98-20 Control & Instrumentation Program 07/10/98 NSED Self-Assessment NE-98-21 Drawing Changes 07/10/98
[NSED Self-Assessment NE-98-22Maintenance Rule implementation 03/18/98 NSED Self-Assessment NE-98-23 Adequacy of Closed Significant LERs, NOVs, CRs 07/10/98 NSED Self-Assessment NE-98-25 Adequacy of Open Significant CRs, LERs, NOVs 08/14/98
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CPS Document Revisioni Description Number Date NSED Self-Assessment NE-98-26 Engineering Duty Director Role 07/10/98 NSED Self-Assessment NE-98-32 Assessment of CPS Nuclear Oversight Groups 08/28/98 NSED Standard Cl-01.00 Instrument Setpoint Calculation Methodology 11/28/90 NSED Standard GS-04.00 System Design and Functional Validation (SDFV)
NTSD Training Guide LP85205 Residual Heat Removal System
NTSD Training Guide LP85239 Main Steam System (includes ADS)
NTSD Training Guide LP85263 Direct Current System
NTSD Training Guide LP85277 Shutdown Service Water System
NTSD Training Guide LP85571 Auxiliary Power System
NTSD Training Guide LP85830 High Pressure Core Spray System
Op Eval / Determination SX Division 1 Flows Lower than the Test Acceptance Criteria 1-98-09-201-OD Op Eval / Determination The Silt Levelin the Shutdown Service Water Fore Bay Area Was
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1-97-10-054-OD Indeterminate Resulting in a Degraded Condition PC-30932 Replacement Valve Operator Motors 1E12F064A Procedure 1003.01 CPS Hardware Change Program
Procedure 1005.06 Conduct of Safety Reviews
Procedure 1005.16 Self-Assessment
Procedure 1014.01 Safety Tagging
Procedure 1014.03 Temporary Modifications
Procedure 1014.06 Operability Determination
Procedure 1016.01 CPS Condition Reports
Procedure 1019.07 Leakage Reduction and Monitoring Program
Procedure 2400.01 Corbicula (Asiatic Clam) Control
Procedure 2602.01 Heat Exchanger Performance of Shutdown Service Water Coolers
Covered by NRC Generic Letter 89-13 Procedure 2800.97 RH-Gjp resting Div I Only 09/21/98 QAP-110.02 QC Quality Assurance Procedure inspection Planning
R: start Readiness Review Report ASME Repair / Replacement Program 08/19/98 Restart Readiness Review Report Control of Calculations 09/17/98 R: start Readiness Review Report Data Trending & Analysis Program (Equipment Failure)
08/14/98 R: start Readiness Review Report Environmental Qualification Program 08/13/98 R: start Readiness Review Report Erosion / Corrosion Program 08/14/98 j
R: start Readiness Review Report Fire Protection Program (Design)
08/17/98 R: start Readiness Review Report Generic Letter 89-13 Program 08/14/98 R: start Readiness Review Report Generic Letter 96-01 Program 08/13/98 R: start Readiness Review Report Infrared Thermography Program 08/14/98 R? start Readiness Review Report Maintenance Rule Program 08/12/98
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CPS Document Revisioni Number Description Date R: start Readiness Review Report Material Condition Management Team Program 08/14/98 R: start Readiness Review Report Non-Destructive Examination (NDE) Program 08/04/98 R: start Readiness Review Report NSED Procedures, Standards & Instructions Program 08/18/98 R: start Readiness Review Report Nuclear Fuel Management Program 08/19/98 R: start Readiness Review Report Predictive Maintenance - Motor Monitoring Program 08/14/98 R: start Readiness Review Report Predictive Maintenance - Oil Analysis Program 08/14/98 R: start Readiness Review Report Predictive Maintenance - Vibration Monitoring Program 08/14/98 R: start Readiness Review Report Primary Containment Leakage Rate Testing Program 08/17/98 R: start Readiness Review Report Reactor Engineering Program 07/30/98 R: start Readiness Review Report Seismic Qualification Program 08/05/98 Ristart Readiness Review Report Special Nuclear Materials Program 07/13/98 R start Readiness Review Report System Engineering Program 08/07/98 RMS Standard 4.01 Document Control
Sity Evaluation 1005.06F001 Temporary Modification 97-048,97-046, and 97-045 Saf:ty Evaluation for Log 95-058 Elimination of DG Operability for ECCS Operability During a Plant Shutdown Saf:ty Evaluation for Log 97-014 Temp Mod for DG Ventilation System (VD)
Sity Evaluation for Log 97-143 Safety Evaluation for ECN 30211
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S1ftty Evaluation Mod RH-048 08/15/98
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Srf :ty Evaluation Screening Mod IA-021, 0
V ndor Manual K2882-0116 Valcor O&M Manual for Model V520-06-14
Vendor Manual M-009-002 Section on Barton Model 580-A Indicating Switches Assessment of Maintenance Effectiveness Cycle 6 Completed PMMIAA003 (IIA 25FB) CHANGE FILTER 01/19/98 DDR Final Report (Volume 1)
09/98
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Engineering Evaluation for CR 1-98-09-201 10/01/98 E26-1003-01 A-El U
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i Maintenance Rule Expert Panel Meeting Mindes 09/10/98 NRC SET Team Report 01/02/98 NSED Monthly Performance Indicators July 1998
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Operability Determination 1-98-09-201-OD PM Evaluation Request (PMER) Sheet for EIN 1PSLSA055 04/06/98
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PMM1AA001 Completed 07/14/98
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PMMIAA001 Completed 10/29/96
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- PMMIAA001 Completed 03/14/95 l
PMMIAA002 Completed 08/18/93
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PMMIAA002 Completed 10/29/96 l
PMMIAA002 Completed 03/14/95
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CPS Document Revision /
Number Description Date PMMIAA004 (llA25FA) CHANGE FILTER Completed 01/19/98 Relief Valve Check Point Data Sheet for ilA128A 02/28/96 Relief Valve Check Point Data Sheet for ilA128B 02/28/96 Review Plan: Detailed Design Review of Selected Modifications, 08/05/98 Plant Changes and Calculations RH-048 ECN 30492 Detailed impact Assessment for Simulator RHR Reflood Analysis
SDFV Project 06/05/98 SDFV Final Report for RHR System SDFV Final Report for SX System Setpoint Program Action Plan 09/30/98 SIRG Meeting Minutes 03/05/98 SIRG Meeting Minutes 06/10/98 SSC Scoping and Performance Criteria Printout for SP, MS, ADS, IA 09/14/98 Systems from Maintenance Rule Database SX System Description
SX System Functional Evaluation Matrix
Summary Report - SIRG Assessment of SDFV Results
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System Health Report 2nd Qtr
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1998 l
Valve Safety Function Data Sheet for ilA0138
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1005.07C001 Temporary Change Checklist (93-0477)
1005.07C001 TPD Preliminary Approval Checklist (TPD No. 96-0128)
l 1005.07C001 TPD Preliminary Approval Checklist (TPD No. 96-0129)
)
1005.07C001 TPD Preliminary Approval Checklist (TPD No. 96-0130)
1005.07C001 TPD Preliminary Approval Checklist (TPD No. 96-0131)
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1005.07C001 TPD Preliminary Approval Checklist (TPD No. 96-0326)
l 1019.07D001 CPS Leakage Reduction Data Sheet
900.01C015 Division 2 - Relief Valve / Low-low Set B21-N668B (F) Relief Valve
Reactor Pressure B21-N669B (F) Channel Functional Checklist 3823.01 ISA Ring Header Isol Valve Testing
l-8801.05D001 Corrections to Instrument Calibrations Data Sheet
l 9000.01D001 Completed Surveillance 09/08/96 9030.01C001 HPCS Drywell Pressure B21-N667C(D,G,H,) Channel Functional
Checklist 9030.01C002 HPCS Reactor Vessel Water Low Level (2),High Level (8) B21-N673
C (D,G,H)
9030.01C003 NS4 Main Condenser Low Vacuum B21-N675A (B,C,D)
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CPS Document Revision /
Number Description Date 9030.01C004 NS4 Main Steam Line Low Pressure B21-N676A (B,C,D) Channel
Functional Checklist 9030.01C005 NS4 Reactor Vessel Water Level B21-N681 A (B,C,D) Channel
Functional Checklist 9030.01C006 RCIC Reactor Vessel Level 2 B21-N692A(B,E,F) Channel Functional
Checklist j
9030.01C007 RCIC Reactor Water Level 8 B21-N693A (B) Channel Functional
Checklist 9030.01C008 ECCS Drywell Pressure B21-N694A (E,B,F) Channel Functional
i Checklist 9030.01C008 Completed Surveillance 08/01/97 9030.01C009 ADS Reactor Vessel Low Level 3 B21-N695A(B) Channel Functional
Checklist 9030.01C009 Completed Surveillance 07/28/97 9030.01C010 ECCS Injection Valve Permissive - Reactor Pressure B21-N697A
(B E.F) Chan,nel Functional Checklist 9030.01C011 RPS Reactor Water Low Level (3), High Level (8) B21-N680A
(B,C,D) Channel Functional Checklist 9030.01C012 RPS Reactor Pressure High B21-N678A(B,C,D) Channel Functional
Checklist 9030.01C013 Reactor Pressure Isolation B21-N679A(B,C,D) Channel Functional
Checklist 9030.01C014 Division 1 - Relief Valve / Low-low Set B21-N668A (E) Relief Valve
Reactor Pressure B21-N669A (E)
9030.01C016 ECCS Reactor Water 1821-N691 A (E,B,F) Channel Functional
Checklist 9030.01C016 Completed Surveillance 09/09/97 9030.01C017 SCRAM Discharge Volume (SDV) High Level C11-N601 A(B,C,D)
Channel Functional Checklist 9030.01C018 SCRAM Discharge Volume (SDV) High Water Level Rod Block C11-
N602A(B) Channel Functional Checklist 9030.01C021 Rod Pattern Controller Low Power Setpoint C11-N6541 (B) Channel
Functional Checklist 9030.01C022 Rod Pattern Controller High Power Setpoint C11-N654 C (D)
Channel Functional Checklist 9030.01C023 RPS/NS4 Drywell Pressure C71-N650A(B,C,D) Channel Functional
Checklist 9030.01C024 RPS Turbine 1st Stage Pressure C71-N652A(B,C,D Channel
J Functional Checklist 9030.01C025 RHR Minimum Flow E12-N652A(B,C) Channel Functional Checklist
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CPS Document Revision /
Number Description Date 9030.01C026 ADS-RHR Pump Discharge Pressure E12-N655A(B,C) Channel
Functional Checklist 9030.01C026 Completed Surveillance 09/10/97 9030.01C027 ADS-RHR Pump Discharge Pressure E12-656A(B,C) Channel
Functional Checklist 9030.01C027 Completed Surveillance 09/22/97 9030.01C028 RHR Containment Pressure E12-N662A(B,C,D) Channel Functional
Checklist 9030.01C028 RHR Containment Pressure E12-N662A(B,C,D) Channel Functional
Checklist 9030.01C029 LPCS Pmp Dsch Press-ADS E21-652, E21-653 Channel Functional
Checklist 9030.01C029 Completed Surveillance 11/13/97 9030.01C030 LPCS Minimum Flow E21-N651 Channel Functional Checklist
9030.01C031 HPCS-RCIC Storage Tank Level E22-N654C(G) Channel Functional
Checklist
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903).01C032 HPCS-Suppression Pool Water Level E22-N655C(G) Channel
Functional Checklist 9030.01C033 HPCS Discharge Pressure, (Minimum Flow) E22-N651 (656)
Channel Functional Checklist 9030.01C034 RCIC Steam Line Flow E31-N683A(B),E31-N684A(B) Channel
Functional Checklist 9030.01C035 RCIC Main Steam Supply Pressure E31-N685A(B) Channel
Functional Checklist j
9030.01C036 NS4 MSIV Flow Isolation E31-686A(B,C,D) Channel Functional
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Checklist j
9030.01C037 NS4 MSIV Flow isolation E31-N687A(B,C.D) Functional Checklist
l 9030.01C038 NS4 MSIV Flow Isolation E31-N688A(B,C,D) Channel Functional
l Checklist l
9030.01C039 NS4 MSIV Flow isolation E31-N689A (B,CD) Channel Functional
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Checklist 9030.01C040 RCIC Storage Tank Level E51-N635A(ti) Channel Functional
Checklist
9030.01C041 RCIC Turbine Exhaust Diaphragm Pressure E51-N655A (B,E,F)
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Channel Functional Checklist 9030.01C042 RCIC Suppression Pool Level E51-N636A(E) Channel Functional
Checklist i
9030.01D001 HPCS Drywell Pressure B21-N667C(D,G,H) Channel Functional
Data Sheet 9030.01D002 HPCS Reactor Vessel Water Low Level (2), High Level (8) B21-
N673C (D,G,H) Channel Functional Data Sheet
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S CPS Document evision/
Description j
Number Date
9030.01D003 NS4 Main Condenser Low Vacuum B21-N675A(B,C,D) Channel
I Functional Data Sheet 9030.01D004 NS4 Main Steam Line Low Pressure B21-N676A (B,C,D) Channel
Functional Data Sheet 9030.01D005 NS4 Reactor Vessel Water Level B21-N681 A(B,C,D) Channel
,
Functional Data Sheet
)
9030.01D006 RCIC Reactor Vessel Level 2 B21-N692A(B,E,F) Channel Functional
Data Sheet 9030.01D007 RCIC Reactor Water Level 8 B21-N693A(B) Channel Functional Data
Sheet 9030.01D008 ECCS Drywell Pressure B21-N694A(E B,F) Channel Functional Data
Sheet 9030.01D009 ADS Reactor Vessel Low Level 3 B21-695A(B) Channel Functional
Data Sheet 9030.01D010 ECCS Injection Valve Permissive-Reactor Pressure B21-
'26 N697A(B E,F) Channel Functional Data Sheet 9030.01D011 RPS Reactor Water Low Level (3), High Level (8) B21-N680A(B,C,D)
Channel Functional Data Sheet S030.01D012 RPS Reactor Pressure High B21-NG78A (B,C,D) Channel Functional
i Data Sheet
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9030.01D013 Reactor Pressure isolation B21-N679A (B,C,D) Channel Functional
Data Sheet 9030.01D014 Division 1 - Relief Valve & Low-Low Set B21-N668A (E) & Relief
Valve Reactor Pressure B21-N669A(E) Channel Functional Data Sheet 9030.01D015 Div. 2-Relief Valve and Low-low Set B21-N668B(F) and Relief Valve
Reactor Pressure B21-N669B(F) Channel Functional Data Sheet 9030.01D016 ECCS Reactor Water Level 1 B21-N691 A (E,B,F) Channel Functional
j Data Sheet 9030.01D017 Scram Discharge Volume (SDV) High C11-N601 A (B,C,D) Channel
Functional Data 9030.01D018 Scram Discharge Volume (SDV) High Water Level Rod Block C-11-
N602A(B) Channel Functional Data 9030.01D021 Rod Pattern Controller Low Power Setpoint C11-N654A (B) Channel
Functional Data 9030.01D022 Rod Pattern Controller High Power Setpoint C11-N654C (D) Channel
Functional Data 903N1D023 RPS/NS4 Drywell Pressure C71-N650A(B,C,D) Channel Functional
Data Sheet 903[556i24 RPS Turbine 1st Stage Pressure C71-N652A (B,C,D) Channel
Functional Data 9030.01D025 RHR Minimum Flow E12-N652A(B,C) Channel Functional Data Sheet
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CPS Document Revision /
Number Description Date 9030.01D026 ADS-RHR Pump Discharge Pressure E12-N655A(8,C) Channel
Functional Data 9030.01D027 ADS-RHR Pump Discharge Pressure E12-N656A(B,C) Channel
Functional Data 9030.01D028 RHR Containment Pressure E12-N662A(B,C,D) Channel Functional
Data Sheet 9030.01D029 LPCS Pump Dsch Press-ADS E21-N652; E21-653 Channel
Functional Data Sheet 9030.01D030 LPCS Minimum Flow E21-N651 Channel Functional Data Sheet
9030.01D031 HPCS-RCIC Storage Tank Level E22-N654C(G) Channel Functional
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Data Sheet l
9030.01D032 HPCS-Suppression Pool Water Level E22-N655C(G) Channel
Functional Data Sheet 9030.01D033 HPCS Discharge Pressure (Minimum Flow) E22-N651 (N656)
Channel Functional Data 9030.01D034 RCIC Steam Line Flow E31-N683A(BJ,E31-N684A(B) Channel
Functional Data Sheet 9030.01D035 RCIC Main Steam Supply Pressure E31-N685A(B) Channel
Functional Data Sheet 9030.01D036 NS4 MSIV Flow Isolation E31-N686A(B,C,D) Channel Functional
Data Sheet 9030.01D037 NS4 MSIV Flow isolat.on E31% A(E.C D) Channel Functional
Data Sheet 9030.01D038 NS4 MSIV isolation E31-N68EM[d,D) Channel Functional Data
l Sheet 9030.01D039 NS4 MSIV Flow isolation E31-N689A(B,C,D) Channel Functional
Data Sheet 9030.01D040 RCIC Storage Tank Level E51-N635A(E) Channel Functional Data
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Sheet 9030.01D041 RCIC Turbine Exhaust Diaphragm Pressure E51-N655A (B.E.F)
Channel Functional Data Sheet 9030.01D042 RCIC Suppression Pool Level E51-N636A(E) Channel Functional
Data 9030.05 Completed Surveillance 04/20/97
9030.05 C-A12-A117 Completed Surveillance 08/02/97 i
9030.10 Analog Trip Module (ATM) Channel Functional / Calibration Check
j Instr.
9051.01 HPCS System Pump Operability
i 9051.01D001 HPCS System Pump Operability Data Sheet
I 9051.02 HPCS Valve Operability Test
9051.02D001 HPCS Valve Operability Data Sheet
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O CPS Document
"I Number Description Date 9051.04 HPCS Automatic Suction Shift Test
9051.05 HPCS Discharge Header Filled and Flow Path Verification
9056.01 Completed Surveillance 07/15/97 9056.04 Completed Surveillance 09/27/97 9061.11C022 Completed Surveillance 11/05/96 9430.30 C-A11-A111 Completed Surveillance 08/22/98 9430.30 C-A12-A111 Completed Surveillance 08/22/98 9430.30 C-A12-A117 Completed Surveillance 08/22/98 9430.30 D-A11-A111 Completed Surveillance 01/02/97 9430.30 D-A12-A111 Completed Surveillance 01/02/97 9430.30 D-A12-A117 Completed Surveillance 01/02/97 9430.30 NSPS Untested Islands / Calibration 1-999 Second Time Delay
9430.30, M001 Untested Islands / Calibration 1-999 Second Time Delay impact Matrix
> 9433.03 ECCS Reactor Vessel Water Level B21-N095A,B Channel
Calibration 9433.03 ECCS Reactor Vessel Water Level 821-N091 A
9433.03 Completed Surveillance 08/31/98 9433.030001 ECCS Reactor Vessel Water Level B21-091 A Calibration Data Sheet
9433.04 Completed Surveillance 07/10/97 9433.05 ECCS Reactor Vessel Water Level B21-091E Channel Calibration
9433.05 Completed Surveillance 09/01/98 9433.05D001 ECCS Reactor Water Level B21-N091E Channel Calibration Data
Sheet 9433.06 Completed Surveillance 12/07/96 9433.07 ECCS Reactor Vessel Water Level B21-N073(G) Channel Calibration
9433.07D001 ECCS Reactor Vessel Water Level B21-N073C Channel Calibration
Data 9433.07D002 ECCS Reactor Vessel Water Level B21-N073G Channel Calibration
Data 9433.08 ECCS Reactor Vessel Water Level B21-N037D(H) Channel
Calibration 9433.08D001 ECCS Reactor Vessel Water Level B21-NO73D Channel Calibration
Data 9433.08D002 ECCS Reactor Vessel Water Level B21-N073H Channel Calibration
Data 9433.09 Completed Surveillance 05/22/92 9433.09 Completed Surveillance 04/07/94 9433.09 Completed Surveillance 09/13/94 f9433.09 Completed Surveillance 01/23/96
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CPS Document Revision /
Number Description Date 9433.09 Completed Surveillance (95A)
09/04/97 9433.09 Completed Surveillance (95B)
09/04/97 9433.09 Completed Surveillance 01/16/98 9433.09D001 ECCS Reactor Vessel Water Level B21-M095A Channel Calibration
Data Sheet 9433.10 ECCS Drywell Pressure B21-N094A,E Channel Calibration
9433.10 Completed Surveillance 07/24/92 9433.10 Completed Surveillance 07/27/92 I
j 9433.10 Completed Surveillance 01/05/94 9433.10 Completed Surveillance 06/21/95 9433.10 Completed Surveillance (94A)
06/26/97 9433.10 Completed Surveillance (94E)
05/06/97 9433.10D001 ECCS Drywell Pressure B21-N094A Channel Calibration Data Sheet
9433.11 Completed Surveillance 06/29/94 9433.11 Completed Surveillance 01/14/97 9433.11 Completed Surveillance 04/02/97 9433.11 Completed Surveillance (94B)
04/02/97 9433.11 Completed Surveillance (94F)
01/15/97 9433.12 ECCS Drywell Pressure B21-N067C(D,G,H) Channel Calibration
j 9433.12D001 ECCS Drywell Pressure B21-N067C Channel Calibration Data Sheet
j 9433.12D002 ECCS Drywell Pressure B21-N067D Channel Calibration Data Sheet
9433.12D003 ECCS Drywell Pressure B21-N067G Channel Calibration Data Sheet
9433.12D004 ECCS Drywell Pressure B21-N067H Channel Calibration Data Sheet
9433.13 ECCS Reactor Steam Dome Pressure B21-N097A(B) Channel
Calibration
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9433.13D001 ECCS Reactor Steam Dome Pressure B21-N097A Channel
Calibration Data 9433.13D002 ECCS Reactor Steam Dome Pressure B21-N097B Channel
Calibration Data 9433.15 HPCS RCIC Storage Tank Level E22-N054C(G) Channel Calibration
9433.15D001 HPCS RCIC Storage Tank Level E22-N054C Channel Calibration
Data 9433.15D002 HPCS RCIC Storage Tank Level E22-N054G Channel Calibration
Data 9433.17 HPCS Suppression Pool Water Level E22-N055C(G) Channel
Calibration 9433.17D001 HPCS Suppression Pool Water Level E22-N055C (G) Channel
Calibration Data
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CPS Document
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Number Description Date 9433.19 ECCS RHR Pump Disch Press ADS E12-N055A, B, C and E12-
N056A,B,C Channel Calibration 9433.19 Completed Surveillance 06/28/94 9433.19 Completed Surveillance 11/14/95 9433.19 Completed Surveillance (55A)
08/12/97
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9433.19 Completed Surveillance (558)
01/14/98 9433.19 Completed Surveillance (55C)
01/14/98 9433.19 Completed Surveillance (56A)
08/12/97 9433.19 Completed Surveillance (56B)
01/15/98 9433.19 Completed Surveillance (56C)
01/14/98 9433.19D001 ECCS RHR Pump Disch Press ADS E12-N055A, B, C/ 56A, B, C
Channel Calibration Data Sheet 9433.20 ECCS LPCS Pump Dsch Press ADS E21-N052/52 Channel
Calibration 9433.20 Completed Surveillance 10/31/92 9433.20 Completed Surveillance 05/26/94 9433.20 Completed Surveillance 11/09/95 9433.20 Completed Surveillance 08/12/97 9433.20 Completed Surveillance (52)
08/12/97 9433.20 Completed Surveillance (53)
08/12/97 9#33.20D001 ECCS LPCS Pump Disch Press 1E12-N051 Channel Calibration
Data Sheet 9433.22 HPCS Flow E-22-N056 Channel Calibration
9433.22D001 HPCS Flow E-22-N056 Channel Calibration Data Sheet
9433.23 ECCS HPCS Pump Discharge Pressure E22-N051 Channel
Calti> ration 9433.23D001 ECCS HPCS Pump Discharge Pressure E22-N051 Channel
Calibration Data Sheet 9433.36 High Pressure Core Pray System Response Time Test
9433.36D001 High Pressure Core Spray System Response Time Data Sheet
9442.02 Completed Surveillance 09/03/98 9442.05 Completed Surveillance 09/03/98 t
9456.04 Completed Surveillance IA085 07/25'97 9456.04 Completed Surveillance IA084 07/26/37_
l 9843.01 ISI Category"A" Valve Leak Rate Test
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9843.01C009 Leak Rate Testing of R.CIC Head Spray (1E51F013 & 1E12F023)
9843.01D001 RCSPlV/Non-RCSPlV Leak Rate Test
9843.01D002 RCSPlV/Non-RCSPlV Leak Test Via Flowmeter
9843.01F002 ISI Category "A" Valve Leak Rate Test Results
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CPS Document Revision /
Description Number Date 9843.01F003 ISI CATEGORY "A" VALVE IDENTIFIED LEAKAGE
9843.01V001 Leak Rate Testing of LPCI"A" Injection
9843.01V002 Leak Rate Testing of LPCS Injection
9843.01V003 Leak Rate Testing of LPCI "B" Injection
9843.01V004 Leak Rate Testing of LPCI "C" Injection
9843.01V005 Leak Rate Testing of HPCG Injection
9843.01V006 Leak Rate Testing of RHR Shutdown Suction
9843.01V007 Leak Rate Testing of RHR "A" to FW & FW "A"
9843.01V008 Leak Rate Testing of RHR "B" to FW & FW "B"
l 9843.01V010 Leak Rate Testing of RPV Head Vent to Drywell Pump
9843.01V013 Leak Rate Testing of Main Steam Drain & MSIV Bypass
9843.01V014 Leak Rate Testing of RWCU From REV
9843.01V015 Leak Rate Testing of RCIC Head Spray ( 1E51F066)
9843.01V016 Leak Rate Testing of Div. I RX Level Keepfill System (1C11F376A &
i 1C11F377A)
9843.01V017 Leak Rate Testing of Div 11 RX Level Keepfill System (1C11F3768 &
1C11F3778)
9861.02 Local Leak Rate Testing Requirements And Type C (Air) Local Rate
Testing l
9861.02D001 LLRT Data Sheet For 1MC-001 - Containment Equipment Hatch Test
Connection 9861.02D002 LLRT Data Sheet For 1MC009
9861.02D003 LLRT Data Sheet For 1MC010
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9861.02D004 LLRT Data Sheet for 1MC014 (S-MC014) RHR SDC Suction
l 9861.02D005 LLRT Data Sheet for 1MC015 - RHR A LPCI Injection
9861.02D006 LLRT Data Sheet for 1MC016 - RHR B LPCI Injection
9861.02D013 LLRT Data Sheet for 1MC035 - HPCS Injection
9861.02D014 LLRT Data Sheet for 1MC036 - LPCS Injection
9861.02D016 LLRT Data Sheet for 1MC042 (S-MC042)
l 9861.02D017 LLRT Data Sheet for 1MC043 (S-MC043)
9861.02D018 LLRT Data Sheet for Retest / Pretest /Special Test
9861.02D019 LLRT Data Sheet for 1MC045 - Main Steam Line Drain
9861.02D020 LLRT Data Sheet for 1MC046 (S-MC046)
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9861.02D021 LLRT Data Sheet for 1MC047 - Containment /Drywell CC Return
9861.02D022 LLRT Data Sheet for 1MC048 - Containment SX Supply
9861.02D023 LLRT Data Sheet for 1MC049 - Containment Breathing Air
9861.02D024 LLRT Data Sheet for 1MC050 MC to Containment
9861.02D025 LLRT Data Sheet for 1MC052 FC Return from Containment
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Number Description Date 9861.02D026 LLRT Data Sheet for 1MC053 FC Supply to Containment
9861.02D027 LLRT Data Sheet for 1MC056 Fire Protection to Containment
9861.02D028 LLRT Data Sheet for 1MC057 - Containment instrument Air Supply
9861.02D029 LLRT Data Sheet for 1MC058 (S-MC058)
9861.02D030 LLRT Data Sheet for 1MC059 (S-MC059)
9861.02D031 LLRT Data Sheet for 1MC060 (S-MC060)
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9861.02D032 LLRT Data Sheet for 1MC061 (S-MC061)
9861.02D033 LLRT Data Sheet for 1MC062/1MC166 - Hydrogen Recombiner 1SB
9861.02D034 LLRT Data Sheet for 1MC063 (S-MC063)
9861.02D035 LLRT Data Sheet for 1MC064 (S-MC064)
9861.02D036 LLRT Data Sheet for 1MC065 - RWCU to Radwaste
9861.02D037 LLRT Data Sheet for 1MC067 (S-MC067)
9861.02D038 LLRT Data Sheet for 1MC068 - Post Accident Sample System
9861.02D039 LLRT Data Sheet for.1MC069 - Equipment Drain Discharge
9861.02D040 LLRT Data Sheet for 1MC070 - Floor Drain Discharge
9861.02D043 LLRT Data Sheet for 1MC078 - SX to RR Pumps
9861.02D044 LLRT Data Sheet for 1MC079 - SF Return to Suppression Pool
9861.02D045 LLRT Data Sheet for 1MC081 Fire Protection to Containment
9861.02D046 LLRT Data Sheet for 1MC082 Fire Protection to Containment
9861.02D047 LLRT Data Sheet for 1MC085 (S-MC085)
9861.02D048 LLRT Data Sheet for 1MC086 (S-MC086)
9861.02D049 LLRT Data Sheet f_or 1MC088 - SX From RR Pumps
9861.02D050 LLRT Data Sheet for 1MC101 (S-MC101 & S-MC101001)
9861.02D051 LLRT Data Sheet for 1MC10%-MC102 & S-MC102001)
9861.02D052 LLRT Data Sheet for 1MC103 IS-MC103)
9861.02D053 LLRT Data Sheet for 1MC104 (5-MC104)
9861.02D054 LLRT Data Sheet for 1MC-106 - Containment HVAC Exhaust
j 9861.02D059 LLRT Data Sheet for 1MC-113 - Containment HVAC Supply
9861.02D062 LLRT Data Sheet for 1MC-152 - lLRT Test Connections
9861.02D063 LLRT Data Sheet for 1MC-173 - H2/02 Monitor 1SB
9861.02D066 LLRT Data Sheet for 1MC-169 - Containment Supply Air Cooler
Instrument 9861.02D067 LLRT Data Sheet for 1MC-173 - H2/02 Monitor 1SA
9861.02D073 LLRT Data Sheet for 1MC204 - Containment SX Return
9861.02D074 LLRT Data Sheet for 1MC205 - Containment SX Supply
9861.02D075 LLRT Data Sheet for 1MC206 (S-MC206)
9861.02D076 LLRT Data Sheet for 1MC208 (S-MC208)
9861.02D077 LLRT Data Sheet for 1MC210 - Post Accident Sample System
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CPS Document Revision /
Number Description Date 9861.05 Water Leak Rate Testing
9861.05D001 RHR A/LPCS Water Leak Rate Test Data Sheet (S-MC021K01 & S-
MC038K04)
9861.05D002 RHR B & RHR C Water Leak Rate Test Data Sheet (S-MC025K02 &
S-MC023K03)
9861.05D003 HPCS Water Leak Test Data Sheet (S-MC037K05)
9861.05D004 RCIC Water Leak Rate Test Data Sheet (S-MC040K06)
9861.05D010 RCIC Turbine Exhaust Water Leak Rate Test Data Sheet (S-
MC039K12)
9861.05D011 SF Suction Line Water Leak Rate Test Data Sheet (S-MC034K13)
9861.05D012 Water Test Data Sheet for Retest / Pretest /Special TEST
9864.01 Excess Flow Check Valve Operability Test
9864.01D001 Low Pressure Excess Flow Check Valve Test Data Sheet for 1E22-
F332 9864.01D002 Low Pressure Excess Flow Check Valve Test Data Sheet for 1E51-
F3778 9864.01D003 Low Pressure Excess Flow Check Valve Test Data Sheet for
1CM002B 9864.01D004 Low Pressure Excess Flow Check Valve Test Data Sheet for
1SM008 9864.01D005 Low Pressure Excess Flow Check Valve Test Data Sheet for
1SM011 9864.01D006 Low Pressure Excess Flow Check Valve Test Data Sheet for
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1CM051 9864.01D007 Low Pressure Excess Flow Check Valve Test Data Sheet for
1 1CM053
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9864.01D008 Low Pressure Excess Flow Check Valve Test Data Sheet for
j 1CM066 9864.010009 Low Pressure Excess Flow Check Valve Test Data Sheet for
1CM067 9864.01D010 Low Pressure Excess Flow Check Valve Test Data Sheet for 1E22-
F330 9864.01D011 Low Pressure Excess Flow Check Valve Test Data Sheet for
1CM003A 9864.01D012 Low Pressure Excess Flow Check Valve Test Data Sheet fr/
1CM002A 9864.01D013 Low Pressure Excess Flow Check Valve Test Data Sheet for
1CM003B 9864.01D014 Low Pressure Excess Flow Check Valve Test Data Sheet for 1sm009
9864.01D015 Low Pressure Excess Flow Check Valve Test Data Sheet for
1SM010
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CPS Document Revision /
Number Description Date 9864.01D016 Low Pressure Excess Flow Check Valve Test Data Sheet for 1E51-
F377A 9864.01D017 Low Pressure Excess Flow Check Valve Test Data Sheet for
1VR016A l
9864.01D018 Low Pressure Excess Flow Check Valve Test Data Sheet for
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1VR0168 -
9864.01D019 Low Pressure Excess Flow Check Valve Test Data Sheet for
.23 1VR018A 9864.01 D020.
Low Pressure Excess Flow Check Valve Test Data Sheet for
1VG0188 9864.01D021 Low Pressure Excess Flow Check Valve Test Data for 1VG056B
9864.01D022 Low Pressure Excess Flow Check Valve Test Data Sheet for
1VG0578
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