IR 05000461/1986018

Safety Insp Rept 50-461/86-18 on 860224-28 & 0303-07. Deviation Noted:Lack of Periodic Testing of Instrument Air Quality to Verify Cleanliness Requirements

ML20140G045
Person / Time
Site:	Clinton
Issue date:	03/25/1986
From:	Knopp R, Scheibelhut C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20140F979	List: IR 05000461/1986018 ML20140F973 ML20140F996
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-1.D.1, TASK-2.K.3.16, TASK-2.K.3.18, TASK-2.K.3.21, TASK-2.K.3.25, TASK-2.K.3.28, TASK-TM 50-461-86-18, NUDOCS 8604010247
Download: ML20140G045 (10)

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U. S. NUCLEAR REGULATORY COMMISSION

REGION III

Report No. 50-461/86018(DRP)

Docket No. 50-461 License No. CPPR-137 Licensee: Illinois Power Company 500 South 27th Street Decatur, IL 62525 Facility Name: Clinton Power Station Inspection At: Clinton Site, Clinton, IL Inspection Conduc ed- Februa y 24-28 and March 3-7, 1986 Inspector: . cheibelhu

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26 8b a l Date Approved By: R. C. Knopp, Chief )sIl Yd6 Reactor Projects Section 1C Date Inspection Summary Inspection on February 24-28 and March 3-7, 1986 (Report N /86018(DRP))

Areas Inspected: Routinc safety inspection by a Regional Inspector of applicant actions on previous inspection findings, evaluation of applicant action with regard to Three Mile Island action plan requirements, and independent inspection activities. This inspection involved a total of 75 inspector-hours onsite by one NRC inspector and includes 0 inspector-hours during off-shift Results: The deviation was related to a lack of periodic testing of instrument air quality, contaninated instrument air could result in common mode failures of air-operated safety-related instruments and component PDR ADOCK 0 % g1 Q

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DETAILS 1. Personnel Contacted Illinois Power Company (IP)

• G. W. Bell, Special Assistant to the Manager, Scheduling

• J. A. Brownell, Licensing Specialist

• J. G. Cook, Assistant Plant Manager

• E. J. Corrigan, Director, Quality Engineering and Verification

• W. C. Cerstner, Executive Vice President

• D. P. Hall, Vice President, Nuclear

• D. L. Holepinge, Director, Testing

• D. Holtzscher, Director, Nuclear Safety

• J. S. Perry, Manager, Nuclear Program Coordination

• W. S. Rives, Supervisor, Training

• J. M. Skov, Supervisor, Commitments Baldwin Associates (BA)

• D. Schlatka, Project Manager

• Denotes those attending the exit meetin The inspector also contacted others of the construction project and operations staf . Safety Evaluation Report Follow-Up Inspection Items (92701)

The Office of Nuclear Reactor Regulation (NRR) requested that Region III confirm that the applicant has acceptably implemented certain confirmatory issues as described in the Safety Evaluation Report (SER)

for the Clinton Power Station (NUREG-0853 with Supplements 1 through 4). In Inspection Reports 50-461/85005 and 50-461/f,5015, these items were listed as Open Items and entered into the Region III tracking system for future inspection. Reported below are the results of the inspection of one of the, item (Closed) Open Item (461/85005-10): " Verify ADS logic modification installed prior to fuel load (TMI Item II.K.3.18) (SER Para. 7.3.3.4)".

In NUREG-0737, " Clarification of TMI Action Plan Requireuente", Item II.K.3.18 required that a ftasibility and risk assessment study be nade to determine an optimum approach to modify Automatic Depressurization System (ADS) logic to eliminate the need for manual actuation under certain conditions to assure adequate core cooling. Through the Boiling Water Reactor Owner's Group (BWROG), the applicant submitted the results of the required study. The study called for the modification of the ADS logic that involves adding a timer that automatically bypasses the need for a high drywell pressure signal to initiate ADS. In Supplement 4 to the SER, Paragraph 7'3. . accepted the results of the study and called for verification of installation of the modificatio ,

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Preoperation Test Procedure (PTP)-NB-05 tested the ADS logic as well as other safety systems logic. The testing was-successfully completed November 5, 198 The inspector reviewed the completed PTP-NB-05 and found in Sections and 7.8 that'the six minute tiners for the logic channel's drywell pressure bypass circuitry performed as required and actuated at 6 i minutes. This testing verified the successful modification of the ADS logi No violations or deviations were identifie . Evaluation of Applicant Action with Regard to Three Mile Island (TMI)

Action Plan Requirements (25401)

The NRC Office of Inspection and Enforcement issued Temporary Instruction (TI) 2514/01, Revision 2, dated December 15, 1980, to supplement the Inspection and Enforcement Manual. The TI provides TMI-related inspection requirements for operating license applicants during the phase between prelicensing and licensing for full power operation. It is divided into two parts. Part I lists requirements that must be closed prior to fuel load. Part 2 lists requirements that must be closed prior to full power operation. Parts 1 and 2 of the TI were used as the basis for inspection of the following TMI items found in NUREG-0737,

" Clarification of TMI Action Plan Requirements". (Closed) Item II.K.3.16: " Reduction of challenges and failures of relief valves". In NUREG-0737, Item II.K.3.16 required that an investigation of the feasibility and contraindications of reducing challenges to the relief valves be conducted. Through the BWROG, the applicant submitted the results of the required study. The study showed that " single failure proofing" of the ADS initiation logic would provide the most effective solution. That is, ADS logies A and E must have concurrent initiation signals before the affected Safety Relief Valves (SRVs) would open. Similarly with ADS logics B and F. The NRC concurred with this conclusion since it is the same logic employed by the General Electric Standard Safety Analysis Report (CESSAR) that has been approve PTP-NB-05 tested this logic change as well as other safety-related systems logic. The testing was completed November 5,.198 The inspector reviewed the completed PTP-NB-05 and found in Sections 7.5 and 7.6 that concurrent initiation signals nust be present in logics A and E before the designated SRVs would open and concurrent initiation signals must be present in logics B and F before their designated SRVs would open. This testing verified the successful modification of the logic circuitr (Closed) Item.II.K.3.18: " Modification of automatic depressurization system logic for increased diversity for some event sequences". The requirements for this item are identical to SER Open Item 461/85005-10. Since the open item was closed in paragraph 2, above, this item is also close .

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c. (Closed) Item II.K.3.21: " Restart of core spray and low pressure coolant-injection systems". In NUREG-0737, Item II.K.3.21 required an analysis to determine if the Low Pressure Core Spray (LPCS) and Low Pressure Coolant Injection (LPCI) systems should not be modified to provide automatic restart. The analysis showed that this restart feature would not be necessary if the High Pressure Core Spray (HPCS) logic were changed to provide autonatic restart of HPCS on low vessel level following manual termination of the HPCS syste However, the Licensing Review Group, Phase 2 (LRG-II) documented a position that the restart capability indicated above was not necessary. In Supplement 4 to the SER, Section 7.3.3.4, the NRC accepted this position and no changes to the emergency core cooling system's automatic restart capabilities were necessar In PTP-HP-01, the applicant tested the control logic for the HPCS and found that it functioned as designed. The inspector reviewed the completed PTP and found that the testing showed that if the HPCS system tripped on high level (level 8) it would automatically restart on low level (level 2) if there were no operator intervention. If the HPCS system tripped on level 8 and an~ operator manually shut down the system, it would not automatically restart on level 2. This is in accordance with the system desig d. (Closed) Item II.K.3.25: "Effect of loss of alternating-current power on pump seals". In NUREG-0737, Item II.K.3.25 required that applicants demonstrate by analysis or experiment that the main recirculating pump seals be,able to withstand a complete loss of alternating-current power for at least two hours without excessive leakage. Through the BWROG,.the applicant demonstrated that the seals could withstand the stated conditions. In Section 15.2 of the SER, the NRC accepted the submittal. In addition, the applicant has installed an auxiliary seal pump (the normal source of seal water is the control rod drive hydraulic pumps) that can be powered from Division 2 of the emergency diesel generator The inspector determined by direct observation that the auxiliary pump was installed and connected to the normal seal piping, that the pump would be powered from the Disision 2 Class lE bus, and that it could be manually started from the control roo e. (Closed) Item II.K.3.28: " Verify qualification of accumulators on automatic depressarization system valves". In Supplement 5 to the SER, the NRC required verification of the following items to assure compliance with Item II.K.3.28: A daily verification that the pressure in the backup air-storage bottles was acceptabl . Operability testing of the SRV accumulator check valves was in accordance with Section XI, Article IWV of the ASME Boiler and Pressure Vessel Cod . Surveillance testing of the SRV pneumatic operators every refueling shutdown to quantify pneumatic operator leakage rate __

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. Environmental qualification of equipmen . Existence of maintenance procedures for the SRV actuator The inspector reviewed the following documents to determine if the above items were being complied with:

CPS No. ISP 9000.02S, " Unit Attendant's Surveillance Log". A unit attendant, as close to the beginning of the shift as possible, must fill out this log. Section 8.6.1 of the procedu're require 4 the recording of the pressure in each of the 16 backup air-supply bottles for the AD3. Minimum acceptable pressure is stated on the form. This complies with item 1, abov CPS No. 1887.00, " Inservice Inspection Program Manual". ~Part B,

" Valves" of the manual provides the requirements for testing Class 1, 2, and 3 valves as delineated by Article IWV of Section X Appendix C of the manual lists the pertinent check valves by valve number, Class (3), Category-(C), and test. method. This complies with item 2, abov Centralized Commitment Tracking (CCT) Item 015372. This item commits the applicant to develop a surveillance p ocedure to determine the actual leakage rates of the SRV pneumatic operators every refueling cycle. The CCT schedule requires the procedure completion before the first refueling outage. Since the CCT system is a quality controlled system, it is reasonable to expect that it will be done. This complies alth item 3, abov Environmental' Qualification Documents: MEQ-83 covers the environmental qualification of the check valves; MEQ-73 covers the SRVs including the actuators; and EQ-65 covers the air solenoid valves. This complies with item'4, abov CPS No. 8216.01, " Safety / Relief Valve Maintenance / Repair". This is a maintenance department procedure that covers the maintenance and/or repair of the SRVs, the pneumatic actuators and solenoid valves that are part of the SRVs. This complies with item 5, abov The inspector reviewed the above documents and found that for item 1, the surveillance log requirerents assure that the backup air-storage supply would always he adequate. For item 2, the inservice inspection program will assure periodic testing of the check valves in accordance with the Code. For item 3, a surveillance procedure was scheduled to be written and implemented prior to the first refueling outage. For item 4, the qualification records show that the equipment is qualified to survive the drywell post-LOCA environment. For item 5, an adequate maintenance / repair procedure exists for the pneumatic operators that includes comprehensive post maintenance / repair testing requirement (Closed) Item I.D.1: " Control-room design reviews". In Supplement I to NUREG-0737, Item I.D.1 required that a detailed control-room design review be conducted to identify and correct design

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deficiencies in the area of human factors and instrumentation. As a first step, a Preliminary Design Assessment (PDA) of the control room and remote shutdown panel was performed in 1981. The'PDA identified a total of 146 Human Engineering Design discrepancies (HEDs). Correction of.these HEDs was accomplished concurrent with the plant construction. The required Detailed Control Room Design Review (DCRDR) was submitted to the NRC for preimplementation review in September, 1985. In Supplement 5 to the SER, Section 8.3, the NRC concluded that the DCRDR satisfied the requirements of Supplement 1 to NUREC-0737 with a number of unresolved items that were made a licensing condition. The DCRDR listed a total of 202 HEDs. These were categorized as follows: Cateory A, High potential for error, or high safety importance or could violate Technical Specifications. This category contained 29 HEDs that must be corrected prior to fuel load. Category B, High potential for error or low safety importance or could result in an unsafe condition in conjunction with other errors. This category contained 34 HEDs that must be corrected within 180 days after fuel load. Category C, Some potential for error or could reduce plant reliability / availability or exhibit a potential for significant financial loss or low safety importance. This category contained 42 HEDs that were scheduled for completion before restart from the first refueling outage. Category D, Low potential for error or no safety implication or no significant financial loss or not interactive or cumulative with other HEDs. This category contained 97 HEDs. No schedule exists for their correction at this tim At the time of the inspection, the applicant stated that all PDA HEDs and Category A HEDs had been correcte To verify that the PDA HEDs and the DCRDR Category A HEDs (a total of 175) were corrected, the inspector utilized the guidelines of MIL Std 105-D to determine the sample size. A sample size of 50 would permit one " reject" to remain at the 95% confidence level that 95%

of the deficiencies were corrected. A random selection of 41 PDA HEDs and 9 DCRDR HEDs was made for detailed review. The inspector determined by direct observation in the control room and at the remote shutdown panel that all of the selected HEDs had been satisfactorily corrected. The inspector also determined that the-Category B HEDs were being tracked as CCT-040777 with completion required'180 days after fuel load. The Category C HEDs are being tracked as CCT-040814 with completion required prior to startup after the first refueling. The inspector concluded that the requirements of Item I.D.1 have been fulfilled as a condition for loading fue No violations or deviations were identifie . Independent Inspection Effort Because of difficulties at other nuclear plants that may have generic significance, the inspector chose to look into a few of these areas at the Clinton Power Statio l

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. Emergency Diesel Generator Control Logic (92701) Some plants have two switches controlling the preferred source tie breaker feeding a Class 1E bus; one in the control room and one in the diesel generator room. Under some conditions, one switch can be in the breaker "open" condition while the other is in the breaker " closed" condition with the breaker close Under_these conditions and with a loss of offsite power, the diesel would start.but the diesel generator tie breaker to the bus would not close and emergency power would not be available. The inspector determined by direct observation that this situation could not occur at CPS because only one preferred source tie breaker switch (in the control room)

exist . Some plants have voltage regulator switches mounted on the diesel generator control panel in the diesel generator roo These switches can be placed in the off, manual, or auto position. In the off position, the generator cannot develop any power. In the manual position, the generator can put out power, but the voltage may be too low. The switch must be in the auto position for proper diesel generator operatio The inspector reviewed the diesel generator electrical drawings with the applicant's personnel and found that in Divisions 1 and 2, if the switch is not in the auto position,- the fact is annunciated in the control room as a " diesel not in auto condition". However, in Division 3, the drawings showed that if the switch is not in the auto position, no annunciation is provided. While administrative controls exist to assure that the switch is.in the auto position, it was determined that a strict interpretatien of Regulatory Guide 1.47, " Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Systems", would require annunciation in this instanc Accordingly, the applicant wrote Condition Report (CR)-1-86-03-037 to provide annunciation in the control room when the Division 3 voltage regulator selector switch is not in the auto position. This is considered to be an open item pending inspection to verify that the condition has been corrected (461/86018-01). Instrument Air Quality (92701)

The instrument air (IA) system at Clinton Power Station is classified as a nonsafety-related, nonquality-related system. This classification is common in the nuclear industry and was accepted for CPS in the SER. IA is used to actuate safety-related components in several safety-related systems, including the following:

(1) Automatic Depressurization System safety-relief valves (2) Other reactor safety-relief valves (3) Main steam isolation valves

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(4) Control rod drive scram valves Industry experience with IA systems indicates that contaminated IA can be the cause of common mode failures. This experience was related to the nuclear industry in IE Information Notice 81-38,

"Potentially Significant Equipment Failures P.esulting From Contamination of Air-Operated Systems", and in IE Circular 81-14,

" Main Steam Isolation Valve Failures to Close".

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The NRC also issued Regulatory Guide (Reg Guide) 1.80, "Preoperational Testing of Instrument Air Systems", in partial response to the concern regarding common mode failures caused by IA systems. Illinois Power committed to the guidance contained in Regulatory Guide 1.80 in the CPS Final Safety Analysis Report (FSAR).

The following provides requirements and recommendations relevant to IA quality at CPS:

(1) Reg Guide 1.80, Section C, " Regulatory Position", states in part, "As part of the initial preoperational testing program, and also after major modifications or repairs to a facility, the instrument air system should be tested as described below

... 5. Verify by test that the system will meet cleanliness requirenents with respect to oil, water, and particulate matter contained in the product air ...".

(2) The CPS SER, paragraph 9.3.1, " Compressed Air Systems", stated in part that "The instrument air system takes filtered, dried, and oil-free air, according to the guidelines of ANSI MC 11.1-1976 (ISA S7.3), " Quality Standard for Instrument Air", fror, the service air system and distributes it to air-operated valves and instrumentation throughout the plant. The air for instrument systems from the service system will be tested at the filter discharge at least yearly for dewpoint and particulate contamination. Acceptable air quality will be in accordance with ANSI MC 11.1-1976. On failure to meet acceptable air quality, branch lines will be tested to determine the extent of problems and corrective action needed".

(3) ANSI MC 11.1-1976/ISA S7.3 states in part that "The maxinum particle size in the air strean at the instrument shall be three (3) micrometres", and "A regular periodic check should be made to assure high quality instrument air".

(4) General Electric (CE) Company, the Nuclear Steam Supply System vendor for CPS, recommends a maximum particle size of 50 microns at the instrument in the IA air strea The requirements and recommendations documented above reflect standard industry practice to address the concern for common mode failures caused by contaminated IA system The inspector reviewed CPS No. ITP2603.01N, Instrument Air Quality, revision 0, dated March 26, 1983. The purpose of this procedure was

"... to provide a means of monitoring and maintaining

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Instrument / Service Air quality". The discussion section of the procedure recognized that potentially significant equipment failures can result from contamination of.the IA system, including contamination by oil, water, dessicant, and rust or other corrosion products. The procedure referenced IE Information Notice 81-38, SER Section 9.3.1, and ANSI MC 11.1-1976/ISA S However, the procedure did not address testing the product air for particulates on any periodic frequency or after major modifications or repair The procedure did not provide acceptance criteria for particulate size in the instrument air stream and did not address the air quality at the instrument. Recent industry experience with IA systems similar in design to the system at CPS indicates that excess soldering flux inside the copper piping system can become entrained in the instrument air stream and potentially prevent proper operation of IA actuated components. In addition, a brief review of IA system maintenance and modification procedures did not identify provisions for tasting the IA product air for particulates after major repairs or modification The failure of Illinois Power Company to address particulate contamination of the IA system product air and to provide for testing of IA product air quality after major modifications or repairs to the facility is a deviation from Reg Guide 1.80, and an apparent deviation from ANSI MC 11.1-1976/ISA S7.3, the SER infornation reflected as a reference to CSP No. ITP2603.01N, and GE recommended practice (461/86018-02).

T.a addition, the applicant classified CPS No. ITP2603.0lN as class code NWWN (no review and approval required other than tLe individual supervisor responsible for preparation of the procedure). Since the procedure apparently was intended to fulfill PSAR commitmants and relates to the safety of operation of the facility, the classification of the procedure was considered to be incorrect. The procedure review and approval process should have identified and corrected the above deviation. This matter is considered to be another example of unresolved item 461/85012-0 No violations were identified. One deviation and a further example of a previous unresolved item were identified in paragraph 4.b abov . Open Items Open items are matters which have been discussed with the applicant, which will be reviewed further by the inspector, and which will involve some action on the part of the NRC or applicant or both. One open item disclosed during the inspection is discussed in Paragraph . Exit Interview The inspector met with the resident inspector and applicant representatives (denoted in paragraph 1) at the conclusion of the inspection on March 7, 1986. The resident inspector summarized the scope and findings of the inspection. The applicant acknowledged the inspector's findings. The applicant did not indicate that any of the

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information disclosed during the inspection could be considered proprietary in natur