Text

LER 93-015-00:on 930630,HPCI Sys Declared Inoperable Due to Indicated Flow During Surveillance.Caused by Foreign Matl That Plugged Portion of Restricting Orifice.Restricting Orifice Removed,Cleaned & reinstalled.W/930728 Ltr
	ML20046C313
Person / Time
Site:	Pilgrim
Issue date:	07/28/1993
From:	Boulette E, Ellis D; BOSTON EDISON CO.
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	BECO-LTR-93-96, LER-93-015, LER-93-15, NUDOCS 9308100126
	Download: ML20046C313 (8)
	v • d • e

. .

. !$f 10 CFR 50.73 BOSTON EDISON Pilgrim Nuclear Power Station Rocky Hill Road Plymouth, Massachusetts o2360 July 28 ,1993 BEco Ltr. 93-96 E. T. Boulette, PhD Senior Vice President-Nuclear U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555 Docket No. 50-293 License No. DPR-35 The enclosed Licensee Event Report (LER) 93-015-00, "High Pressure Coolant Injection

' System Made Inoperable Due to Indicated Flow During Surveillance", is submitted in accordance with 10 CFR Part 50.73.

Please do not hesitate to contact me if there are any questions regarding this report.

bY LLLL ,

E. T. Boulette, PhD DWE/bal

Enclosure:

LER 93-015-00 cc: Mr. Thomas T. Martin Regional Administrator, Region I ,

U.S. Nuclear Regulatory Commission ,

475 Allendale Rd.

King of Prussia, PA 19406 Mr. R. B. Eaton Div. of Reactor Projects I/II Office of NRR - USNRC-One White Flint North - Mail-Stop 14D1 11555 Rockville Pike Rockville, MD 20852 Sr. NRC Resident Inspector - Pilgrim Station Standard BECo LER Distribution n o n n e ,-

9308100126 930729 PDR S

ADOCK 05000293 PDR

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NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104

< s-m EXPIRES 5/31/95 ESTIMATED DURDEN PER RESPONSE TO COMPLY WITH THIS LICENSEE EVENT REPORT (LER) 7"OSnE&"#f0n#3%SOEsi?"

AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20S554X)01, AND TO THE PAPERWORK REDUCTION PROJECT 13150-010AJ, OFFICE

@ee reverse for number v dig'ts!charactets for each blocN OF MAh.AGEMEN1 AND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) PAGE (3)

PILGRIM NUCLEAR POWER STATION 05000 - 293 1 of 7 VIYLE (4)

High Pressure Coolant injection System Made inoperable Due to Indicated Flow During Surveillance EVENT DATE (5) LER NUMBER (61 - REPORT DATE (7) OTHER FACILITIES INVOLVED (8)

SEQUENTIAL REVISION f ACluTY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER NUMBER MONTH DAY YEAR N/A 05000 F ACiUTY NAME DOCKET NUMBER 06 30 93 93 015 00 07 28 93 N/A 05000 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REOutREMENTS OF 10 CFR 6: (Check one or rnore)(11)

DE (9) N 20.402(b) 20.405(c) 50.73(a)(2)(iv) 73.71(b) 20.405(a)(1)(i) 50.36(c)(1) X 50.73(a)(2)(v) (D) 73 71(c)

LE E 0) 100 20.405(a)(1)(ii) 50.36(c)(2) 50.73(a)(2)(vii) OTHER 20.405(a)(1)(iii) 50.73(a)(2)(i) 50.73(a)(2)(viii)(A) ,

20 405(a)(1)(iv) 50.73(a)(2)(ii) 50.73(a)(2)(viii)(B) grg 20 405(a)(1)(v) 50.73(a)(2)(iii) 50.73(a)(2)(x) Form 366A) -

LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER pnclu:le Avea Codel Douglas W. Ellis - Senior Compliance Engineer (508) 747-8160 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE SYSTEM COMPONENT MANUFACTURER RD CAUSE SYSTEM COMPONENT MANUrACTURER f S E BJ P B580 Y f SUPPLEMENTAL REPORT EXPECTED (14) " * " # #

EXPECTED YES NO SUBMISSION mes, compete EXPECTED sueutssioN DATE) y DATE (15)

ABSTRACT (Limit to 1400 spaces, i e., approximately 15 single-spaced typewritten lines)(16)

On June 30,1993, at 0001 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , the High Pressure Coolant Injection (HPCI) System was declared inoperable and a seven day Technical Specification 3.5.C.2 Limiting Condition for Operation (LCO) was entered. The system was declared inoperable because the pump flowrate of 4000 gpm was less than 4250 gpm during the monthly surveillance test. The system was i later made inoperable for purposes of investigation.

The cause of the low flowrate was some foreign material that plugged a portion of the restricting orifice located in the system's full flow test pipeline. The source of the foreign material was most probably the Condensate Storage Tanks (CSTs). Corrective action taken consisted of the removal, cleaning, and re-installation of the restricting orifice.

After the restricting orifice was cleaned and re-installed, the system was tested with satisfactory results and the LC0 was terminated at 2001 hours0.0232 days 0.556 hours 0.00331 weeks 7.613805e-4 months on July 1, 1993. During the period the system was declared and made inoperable, the systems specified by Technical Specification 3.5.D.2 were verified operable. Additional corrective action being planned ,

includes visual inspection of at least one of the CSTs for foreign material.

The event occurred while at 100 percent reactor power. The reactor mode selector switch was in the RUN position. The Reactor Vessel (RV) pressure was 1021 psig with the RV water temperature at 533 degrees Fahrenheit. This report is submitted in accordance with 10 CFR 50.73(a)(2)(v)(D). The event posed no threat to the public health and safety.

RC FORM SU'sA (5-92)

MRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-o104

. o.m . EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WfTH THIS LICENSEE EVENT REPORT (LER) = ^#"o's20 4" ERTUO L Ja*T2 "

TEXT CONTINUATION $EC8Y CC SS N OWDCWO TO THE PAPERWORK REDUCTION PROJECT (3150@04), orFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

FACIUTY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)

YEAR M 2 'of 7 I PILGRIM NUCLEAR POWER STATION 05000 293 93 --015- 00 TEXT pf more space is required, use additional copies of NRC Form 366A)(17)

BACKGROUND The High Pressure Coolant Injection (HPCI) System is designed to pump water into the Reactor Vessel (RV) for a wide range of RV pressures. The Reactor Core Isolation Cooling (RCIC) System is also designed to pump water into the RV, similar to the HPCI System. Two sources of water for the HPCI/RCIC System are available, the Condensate Storage Tanks (CSTs) and the Suppression Pool. The CSTs provide reactor grade water for injection into the RV. Water pumped from either the CSTs or Suppression Pool via the HPCI/RCIC System is injected into the RV via feedwater piping and is distributed within the RV through the feedwater spargers.

Each of the two CSTs has a 75,000 gallon reserve for the HPC1/RCIC System. The HPCI/RCIC Systems' pumps take suction from the bottom of the CSTs while other service demands, including the backup source of water for the Core Spray System, are physically isolated by ,

suction piping raised to an elevation above the 75,000 gallon requirement. The Suppression Pool is the primary source of water for the Core Spray System and Residual Heat Removal (RHR) System / Low Pressure Coolant Injection (LPCI) mode, and is the secondary source of water for the HPCI and RCIC Systems.

A full flow functional test of the HPCI or RCIC System is performed by taking a suction from the CSTs suction header and discharging the flow through a test return pipeline to the CSTs return header. Each test return pipeline includes a restricting orifice and motor operated globe type valve. The HPCI System full flow test restricting orifice R0-2301-59 is a circular flat plate with a perforated conical central section. The RCIC System full flow test restricting orifice R0-1301-8 is a flat plate with a single perforation in the central section. The normally closed motor operated valve is throttled during testing to adjust the pump discharge pressure.

EVENT DESCRIPTION On June 30, 1993, at 0001 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , the HPCI System was declared inoperable and a seven day Technical Specification 3.5.C.2 Limiting Condition for Operation (LCO) was entered. The system was declared inoperable because the indicated pump flowrate was less than the ;

acceptance criteria during the monthly surveillance test. The indicated pump flowrate was approximately 4000 gpm at 1330 psig. The acceptance criteria is 24250 gpm at 1225 psig.

The test was being conducted in accordance with Procedure 8.5.4.1 (Rev. 41), "High Pressure Coolant Injection System Pump and Valve Monthly / Quarterly Operability".

The system was isolated and made inoperable on June 30, 1993, at approximately 1200 hours0.0139 days 0.333 hours 0.00198 weeks 4.566e-4 months , for investigative purposes.

Problem Report 93.9308 was written to document the low pump flowrate. The NRC Operations l Center was notified in accordance with 10 CFR 50.72 at 0135 hours0.00156 days 0.0375 hours 2.232143e-4 weeks 5.13675e-5 months on June 30, 1993. !

This event occurred while at 100 percent reactor power. The reactor mode selector switch was in_the RUN position. The Reactor Vessel (RV) pressure was approximately 1021 psig with the RV water temperature at approximately 533 degrees Fahrenheit.

NFC FORM 366A @ 02)

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FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) vEAn *fESEn" EME 3 of 7 PILGRIM NUCLEAR POWER STATION 05000-293 93 --015-- 00 TEXT (if more space is required, use additional copies of NRC Form 366A)(17)

CAUSE The cause of the low flowrate was foreign material that plugged a portion of the HPCI full flow test restricting orifice. Specifically, approximately seven pieces of plastic tie-wrap, three other pieces of plastic, three pieces of grinding wheel type material, and one piece of ferrous type material were found to be pluggino six to seven of the 70-75 perforations in the restricting orifice. The neutral colored plastic tie-wrap material was rectangular in shape and ranged in size from approximately one-eighth to one-half inch in thickness and approximately one-quarter to one inch in length. One of the three other pieces of plastic material was neutral colored and approximately one-sixteenth inch in diameter and three inches long. The other two pieces of plastic material were blue in color, irregular in shape and were approximately three-quarters of an inch at the largest diagonal. The grinding wheel material was irregular in shape and ranged in size from approximately one-half to three-quarters of an inch at the largest diagonal. The ferrous type material was dark in color with a lustrous sheen, roughly cylindrical in shape and approximately one-quarter inch in diameter and length.

The foreign material was discovered after the HPCI System instrumentation and HPCI turbine speed control system were calibrated in accordance with Procedures 8.E.23 (Rev. 26), "HPCI System Instrumentation Calibration", and 8.E.23.1 (Rev. 4), "HPCI Turbine Speed Control System Calibration", with satisfactory as-found results. After the calibrations, the HPCI System was tested June 30, 1993, at 1922 hours0.0222 days 0.534 hours 0.00318 weeks 7.31321e-4 months in accordance with Procedure 8.5.4.1 with the same results as on June 29, 1993, at approximately 2358 hours0.0273 days 0.655 hours 0.0039 weeks 8.97219e-4 months . ,

Subsequent investigation led to the inspection of the restricting orifice and discovery of the foreign material. The investigation included comparison of HPCI System operating characteristics such as pump flow and discharge pressure, turbine speed, and valve M0-2301-10 position. The investigation revealed the valve M0-2301-10 position, typically adjusted to approximately 50 percent open for surveillance testing, had changed during surveillance testing in 1992 and had to be adjusted to approximately 100 percent open to achieve the acceptable discharge pressure of 21225 psig with the turbine speed at approximately 4000 rpm. The HPCI turbine speed controller was believed to be related to the cause for the change and therefore, the speed controller was replaced during the 1993 refueling outage. During startup from the refueling outage, the HPCI System was surveillance tested with satisfactory results for pump flow and discharge pressure. The ,

valve M0-2301-10 position was approximately 100 percent open, the discharge pressure was approximately 1320 psig, and the turbine speed was approximately 4100 rpm. In contrast, ,

the valve M0-2301-10 position was approximately 100 percent open, the discharge pressure was approximately 1330 psig, and the turbine speed was approximately 4200 rpm on June 29 and on June 30, 1993. The HPCI main pump is rated at 24250 gpm at a speed of 4000 rpm and was manufactured by Byron Jackson Pumps (model DVMX,10X12X15, two stage, horizontal centrifugal type). The increased turbine speed was due to the foreign material that caused an imbalance between demand flow and sensed flow and consequent control system increase in turbine speed.

NRC f OAM 366A (5-02)

MC FDRM 36SA U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104

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FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)

YEAR & BR 4 Of 7 PlLGRIM NUCLEAR POWER STATION 05000-293 93 --015-- 00 l TEXT (tf more space is required, use additional copies of NRC Form 366A)(17)

Strainers are installed at the Suppression Pool end of the pump suction piping of the HPCI, RCIC, Core Spray and RHR Systems because of debris introduced by a design basis accident. The strainers are perforated with holes approximately one-eighth inch in diameter. The strainers are designed to ensure adequate pump net positive suction head during the recirculation phase of a loss of coolant accident. Strainers are not installed at the CST end of the pump suction piping of the HPCI, RCIC, and Core Spray Systems.

Research could not determine a definite reason for no strainers at the CSTs end of the pump _ suction piping. Apparently, the CSTs were assumed to be free of foreign material and therefore, no strainers were installed. The nonsafety-related CSTs are not seismically qualified. The CSTs are enclosed, welded steel cylindrical structures approximately 50 feet tall. Each CST is equipped with two bolted manways, a ladder fixed to the exterior surface, and piping connections / penetrations on the bottom surface. The lower manway is approximately four feet above ground level. The upper manway is located on the top surface. The interior of the CSTs including the piping penetrations are not readily accessible and the manways are seldom open. The HPCI/RCIC pumps' suction piping and test return piping are connected to suction and return headers. The headers are connected to the CSTs. The 18 inch HPCI/RCIC suction header penetrations are located approximately 10 feet from the periphery of the inwardly convex bottoms of the CSTs.

No work was performed during the 1992 mid-cycle outage or 1993 refueling outage (RF0 9) involving the opening of the boundary of the HPCI booster / main pump or HPCI pumps' suction and discharge piping and valves. Therefore, the direct source of the foreign material found in the HPCI restricting orifice is believed to be the CSTs.

t Procedure 1.4.35, " Personnel and Material Controls", is the administrative procedure governing housekeeping zone levels and personnel / material and tool accountability. A review was conducted regarding the effectiveness of the procedure. The review included -

interviews of personnel and review of significant maintenance activities having the potential for introducing foreign material into the Refuel Cavity or Reactor Vessel. The interviews with refuel floor co-ordinators for RF0 8 and RF0 9 indicated logs of ;

materials / tools were maintained. Loose parts / foreign material inadvertently introduced into the Refuel Cavity / Reactor Vessel or Suppression Pool during RF0 8 or RF0 9 were reported and retrieved or evaluated.

1 CORRECTIVE ACTION The HPCI full flow test restricting orifice was removed, cleaned, and reinstalled. The l HPCI System was subsequently tested in accordance with Procedure 8.5.4.1 with satisfactory results on July 1,1993. The system was declared operable and the seven day LC0 was terminated at 2001 hours0.0232 days 0.556 hours 0.00331 weeks 7.613805e-4 months on July 1, 1993.

Corrective action being planned when this report was prepared includes visual inspection of at least one of the CSTs for foreign material. Additional action, if any, will be based upon the results of the inspection. This report will be supplemented if the inspection reveals significant new information or if significant additional action is necessary as a result of the inspections (s).

NRO FORM 386A (5-92)

NRC FDRM 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104

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EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WfTH THIS LICENSEE EVENT REPORT (LER) 02lEa"nD!A"ntEM"iMoOindn%^2 TEXT CONTINUATION Nu"4E*Es's"1"$Nofo"E*Dc No"N OSES7MEEEDEDYGWAS?NGT N, FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)

YEAR R MB R 5 of 7 PILGRIM NUCLEAR POWER STATION 05000-293 93 --015- 00 TEXT (if more space is required, use additional copies of NRC Form 366A)(17)

PREVENTIVE ACTION i Preventive action being planned when this report was prepared includes the removal, inspection, and cleaning (if necessary) of the HPCI restricting orifice at a planned interval. Based on the results of the inspections, the inspection frequency may be increased, decreased, or eliminated. This action will be scheduled and tracked via the Preventive Maintenance Program.

SAFETY CONSE0VENCES This event posed no threat to the public health and safety.

The RCIC System, Automatic Depressurization System, Core Spray and RHR (LPCI) Systems were verified operable as specified by Technical Specification 3.5.D.2 during the period the HPCI System was declared and made inoperable. Moreover, the indicated pump flowrate of approximately 4000 gpm at 1330 psig, although less than the acceptance criteria of > 4250 gpm at >l225 psig, was due to the plugging of a portion of the restricting orifice located in the flow path that returns the water used for the test to the CSTs. The restricting orifice is not located in the flow path that injects water into the RV. The satisfactory as-found results of the calibrations of the HPCI instrumentation and HPCI turbine speed control system and satisfactory results of the testing conducted after the removal of the ,

foreign material from the restricting orifice, provides reasonable assurance the HPCI '

System was operable for core cooling when surveillance tested on June 29-30, 1993.

The foreign material in the HPCI System posed no threat to the public health and safety.

The HPCI System features a booster pump and main pump that are both driven by the system's steam turbine. The majority of the booster pump flow is directed to the main pump via a 12 inch pipe. A portion of the booster pump flow provides turbine lube oil cooling and gland seal condensing. This flow is directed from the booster pump via a two inch pipe and branch piping downstream of a two inch, self regulating pressure control valve (PCV-2301-46). The piping to the gland seal cooler (GSC) is equipped with a restricting orifice (R0-2301-60) having a single perforation approximately 0.635 inch in diameter.

The piping to the lube oil cooler (LOC) is equipped with a restricting orifice (R0-2301-61) having a single perforation approximately 0.477 inch in diameter. The GSC and LOC tubing is approximately 0.625 inch in diameter. The piping to or from the GSC and LOC is also equipped with pressure and temperature sensing instruments. These instruments provide alarm and/or indication functions but no turbine / pump trip functions. The RCIC System features a main pump that is driven by the system's steam turbine. A portion of the pump flow is directed to the turbine barometric condenser and lube oil cooler via piping, a restricting orifice, and pressure control valve similar to the HPCI System. The piping to or from the barometric condenser and lube oil cooler is equipped with pressure and temperature sensing instrumentation similar to the HPCI system. These instruments also provide alarm and/or indication functions but no turbine / pump trip functions. The HPCI GSC and LOC pressure and temperature, and RCIC barometric condenser and lube oil cooler pressure and temperature, are checked during monthly / quarterly surveillance testing. A performance review of the HPCI/RCIC auxiliary systems that could be impacted by foreign material indicated no detected degradation.

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FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) vtAR 1u"iER "EE 6 of 7 PILGRIM NUCLEAR POWER STATION 05000-293 93 --015- 00 TEXT (if more space is required, use additional copies of NRC Form 366A)(17)

The performance of pumps including the HPCI, RCIC and Core Spray System pumps is trended quarterly in accordance with Procedure 8.I.1.1, " Inservice Pump and Valve Testing Program". The trending includes pump vibration and hydraulic performance. Therefore, the effect of foreign material upon the performance of the system pump (s) is detectable and corrective action would be initiated if necessary.

The Core Standby Cooling Systems (CSCS) consists of the HPCI System, Automatic Depressurization System (ADS), Core Spray System and RHR System (LPCI mode). Although not part of the CSCS, the RCIC System is designed to provide high pressure core cooling, similar to the HPCI System. In the event the HPCI System was or were to become inoperable when its operation was necessary, the RCIC System would be initiated for high pressure core cooling. Alternately, in the event the RCIC System was or were to become inoperable when its operation was necessary, the HPCI System would be initiated for high pressure core cooling. In the unlikely event the HPCI and RCIC Systems were inoperable and high pressure core cooling was necessary, the ADS would be initiated to reduce the RV pressure '

for low pressure core cooling provided independently by the Core Spray System and RHR System /LPCI mode.

Foreign material similar to the material discovered in the HPCI restricting orifice would pose no threat to the public health and safety if introduced into the Reactor Vessel via HPCI/RCIC System operation for level control. Foreign material has been discovered in or inadvertently introduced into the RV during previous refueling outages. The foreign material has included material either identical or similar in material properties and size as the foreign material found in the HPCI restricting orifice. The assessments of the foreign material discovered in or not retrievable from the RV typically considered the potential for fuel assembly flow blockage and subsequent fuel damage, the potential for interference with control rod operation, and the potential for corrosion or other chemical action to reactor materials. The assessments have concluded safe reactor operation was not or would not be compromised by the presence of the foreign material.

This report is submitted in accordance with 10 CFR 50.73(a)(2)(v)(D) because the HPCI System was made inoperable.

SIMILARITY TO PREVIOUS EVENTS A review was conducted of Pilgrim Station LERs submitted since January 1984. The review focused on LERs involving foreign material in the HPCI, RCIC, or other system. The review identified no LERs involving foreign material in the HPCI and/or RCIC System, but did identify LER 85-001-00 that involved foreign material in the Standby Liquid Control (SLC)

System.

NHC FORM 366A $ S2)

NRC F'ORM 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104

. t 5-m , EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTTH THIS LICENSEE EVENT REPORT (LER) l"!3%lWWJSL"EME0Jf00,4%TTO '

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TO THE* PAPERWORK REDUCTION PRCUECT 0150 0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503, FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)

YEAR N 7 Of 7 PILGRIM NUCLEAR POWER STATION 05000-293 93 --015-- 00 TEHT (if more space is required, use additional copies of NRC Fum 366A)(17)

For LER 85-001-00, the SLC System was declared inoperable on January 1,1985, during power .

ascension from a refueling outage. The reactor, which was at approximately 22 percent power, was brought to a cold shutdown condition in accordance with Technical Specification 3.4.D. The system was being surveillance tested and the Train 'A' relief valve lifted at approximately 600 psig versus a setpoint of 1425 psig. Train 'B' was subsequently tested and, although the train met Technical Specification flow requirements, was declared inoperable when debris (i.e., rubber gloves and masking tape) was observed floating in the system's main tank and test tank. An investigation found the main tank manway cover and test tank cover were open at times during the recirculation pipe replacement and refueling outage. It was postulated that the debris either fell or was thrown into the tanks.

Corrective action taken included: disassembly, cleaning, and reassembly of the Train 'B' relief valve; visual inspection of the internals of other components; flushing; l reinstalling the test tank cover with locking mechanisms; and bolting the cover to the i main tank manway.

ENERGY INDUSTRY IDENTIFICATION SYSTEM (EIIS) CODES The EIIS codes for this report are as follows:

COMPONENTS CODES j Orifice OR Pump P SYSTEMS High Pressure Coolant Injection (HPCI) System BJ 4

b i

NRC FORM 368A (5 9a

._ _ _ . _

ML20046C313

Text

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