NG-88-0424, 1987 Annual Rept of Facility Changes,Tests,Experiments & Safety & Relief Valve Failures & Challenges

1987 Annual Rept of Facility Changes,Tests,Experiments & Safety & Relief Valve Failures & Challenges

ML20147F185
Person / Time
Site:	Duane Arnold
Issue date:	12/31/1987
From:	Rothert W IES UTILITIES INC., (FORMERLY IOWA ELECTRIC LIGHT
To:	Murley T NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM), Office of Nuclear Reactor Regulation
References
NG-88-0424, NG-88-424, NUDOCS 8803070264
Download: ML20147F185 (44)

Text

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Iowa Electric Light and Power Company

' February 29, 1988 NG-88-0424 Dr. Thomas Murley, Director Office of Nuclear Reactor Regulation-U.S. Nuclear Regulatory Conunission

~ Attn: Document Control Desk Washington, DC 20555

Subject:

Duane Arnold Energy Center Docket No: 50-331 Op. License No: DPR-49 1987 Annual Report of Facility Changes, Tests, Experiments, and Lafety and Relief Valve Failures and Challenges File: A-118e

Dear Dr. Murley:

In accordance with the requirements of Appendix A to Operating License DPR-49,10 CFR Part 50.59(b), and NUREG-0737 (Item II.K.3.3), please find enclosed the subject report covering the calendar year 1987.

Very truly yours, William C. Ro her Manager, Nuclear Division WCR/DJM/pjv*

cc: D. Mienke L. Liu L. Root R.

J. R.McGaughy(NRC-NRR)

Hall A. Bert Davis (Region III)

NRC Resident Office Commitment Control 870021 8803070264 071231 PDR ADOCK 05000331

.DCD R

General office

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• Cedar napids, lov s2 toe *329/39a44s1 Y I

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y. .$e SECTION A - PLANT DESIGN CHANGES r

.This section contains brief descriptions of and reasons for plant design changes completed during the calendar year 1987 and summaries of

-the safety evaluations for those changes, pursuant to the _ requirements of.10 CFR Part 50.59(b) .

The basis for inclusion of a Design Change Package (OCP) in this report is closure of the package at the Duane Arnold Enerav Center-(DAEC) in the calendar year of interest. It is noted that portions of some DCPs listed were partially closed in orevious years.

DCP No. 803 Addition of Chemical Treatment Eauipment to the Circulating Water System Description and Rasis for Chanae: A significant reduction in the heat transfer efficiency of the main condenser and an increase in the corrosion rate of the main condenser and the water heat exchanger surfaces were caused by (1) ageneral service build up of scale due to high calcium concentrations in the circulating water, (2) a thick mat of biogrowti which could not be adequately controlled by chlorination of the circulating water system, and (3) a build up of silt in areas of low flow velocities. To solve these problems tanks, pinos, valving and controls for the addition a biocide (2) a dispersant and (3) a stabilizer that of (1) the crystaline growth of calc,ium carbonate were installed.

stons Sununary of Safety Evaluation: This change was not safety-related.

The installed equipment does not interface with any safety-related system. Additionally, reduction of the corrosion rate improved the reliability and heat transfer characteristics of the-circulating water system. All piping installed meets ANSI B31.1 power plant piping requirements. The chemical storage tanks are mounted on a concrete pad with a spill protection dike and 2" drain valve. No unreviewed safety questions existed.

DCP No. 1033 Radwaste Solidification Process Pipino Description and Rasis for Change: All radwaste shipped by the OAEC for burial must meet the burial site reauirements outlined in NRC Information Notice 80-24 Beginning iluly 1,1481, the burial sites required that all spent resin which exceed activity levels of 1 uC1/aram must be solidified to insure no leakage after burial. Under this nCP oicing was installed to deliver resin slurry from the process,pioing'in the centrifuge room to the solidification equipment. This new oiping also provides the capability to samole the resin mixture.

Summary of Safety Evaluation: This DCP was not safety-related and does not affect the safe shutdown of the reactor. The new nipinq meets the NRC criteria concerning radwaste piping, Regulatory Guide 1.143, except the methods used for seismic aualification were consistent with original plant design and not uoaraded to Regulatory Guide 1.143 standards. Because this system ties into ASME Section III, Class 3 and ANSI B31.7, Class 3 piping, the interface points have been modified to conform to the original construction code. No unreviewed safety questions existed.

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DCP No.-1065 Orywell Radiation Monitors Finstrip Heaters and Improved Detectors Description and Basis for Change: The drywell radiation monitor detectors had a history of premature f ailures due to moisture intrusion. The air sampling system radiation monitors provide an early indication of leaks in the drywell or torus and serve as an alternate to the drywell leak detection system. The solution to this problem was to seal the detectors to prevent moisture damage, and to heat and insulate the detector assemblies to prevent condensate from forming around the detectors by the addition of circuit-protected finstrip heaters and fire retardant insulation.

Summary of Safety Evaluation: This change was not safety-related.

The detectors do not interf ace with any safety-related components, nor do they initiate any automatic protective functions. The air sampling piping for the drywell radiation monitors is isolated at the drywell penetration during a Group III Isolation. The modification did not alter the operation or design function of the system. Additionally, it increased the availability and reliability of the system such that the margin of safety defined  %

by Technical Specification (T.S.) 3.6.C.2. is improved. No unreviewed safety questions existed.

DCP No. 1086 Replace Radiation Element / Radiation Indication Monitor (RE/ RIM) 6101 A/,B & 7606 A/B Description and Basis for Change: DCP 1086 corrected a number of problems associated with this equipment, including seismic qualification inadequacies, poor equipment reliability and availability due to adverse weather conditions, and difficulties in performing the required calibration of the equipment. The solutions to these problems were, as follows, (1) new seismically-qualified instruments for RE/ RIM 6101 A/B and 7606 A/B were procured and installed, (2) RE/ RIM 6101 A/B were relocated from an outdoor location to an indoor location downstrean of the standby filter unit (SFU) pre-heat coils, which is in the normal air flow path, (3) temperature inputs to the SFU lockout relay were removed, but the existing low temperature alarm was retained, and (4) an access hole (for the insertion of a calibration source) was provided to better f acilitate calibration.

Summary of Safety Evaluation: This modification was safety-related. It did not change the original design intent or function of RE/ RIM 6101 A/B and 7606 A/B. The new instruments were seismically qualified in accordance with IEEE-344.

Relocating the detectors downstream of the preheat coils eliminated their exposure to rain and winter weather. Upgrading the instrtsnents provides more reliable operation and more accurate readings because the upgraded equipment was better qualified to operate during and af ter a seismic event. Habitability of the control room is not affected by the removal of the temperature inputs that initiate auto startup of the SFUs because the alarm function vas retained and the operator will have sufficient time to manually start the SFUs upon receipt of an alarm. No unreviewed safety questions existed.

.,  ; c DCP No. 1131 _

.Radwaste Cuno Filters Piping Install'ation Description and Basis for Change: The piping to/from the Cuno filters from the radwaste surge pump had the potential for leakage of radioactive. fluid which could have resulted in potential equipment and personnel contamination. Additionally, that piping returned filtered liquid to the radwaste surge tank. This design change replaced the piping with permanent steel pipe and.added cross-tie piping to return filtered water to the radwaste-collector tank, the floor drain collector tank, or the chemical waste tank for added flexibility and efficiency in the processing of liquids.

Summary of Safety Evaluation: This change reduced personnel radiation exposure and reduced the likelihood of leaks or spillage of contaminated fluid, while adding operational flexibility to the radwaste system and produced no conditions not previously analyzed in the FSAR. No changes in plant technical specifications were required. No unreviewed safety questions existed.

DCP No.1149 Modification to Diesel Generator Air Start System Description and Basis for Change: Various problems to the diesel generator air start system existed before DCP 1149. These problems were: (1) leakage of system air via the relief valves due to relief valve chatter. The relief valve chatter was caused by too small of differential pressure between the diesel air start system operating pressure and- the relief valve setpoint; (2) pressure switch f ailure due skid vibrations and air pulsations from the reciprocating compressors; (3) unregulated battery _

chargers for the diesel starting air compressor batteries caund premature failure of the batteries; (4) air leakage from the diesel air start regulator located on the crank case of- the emergency diesel generators (due to repeated disconnection and reconnection of the copper air supply tubing to the oil booster tanks during testing of the system); and (5) particulate build up on a solenoid valve downstream of the air filter. Additionally, small mounts of corrosion were detected in the assembly, indicating the presence of moisture. DCP 1149 made the following modifications: (1) the maximum operating pressure of the diesel generator starting air system was lowered from 240 psig to 225 psig, new relief valves were installed and the low pressure alarm switch setpoints (PS-3232A&B) and (PS-3233A&B) were reset from 200 and 225 psig to 175 and 200 psig respectively; (2) pressure switches PS-3234A&B were relocated away from their respective compressor skids; (3) the unregulated battery chargers were replaced with regulated battery chargers; (4) instrument valves were installed upstream of the oil booster tanks for emergency diesel generators 1G-21 and 1G-31 to preclude air leakage; and (5) new air filters, housings and flanges were installed.

Additionally, moisture drain valves and flexible tubing were installed in the air lines of the diesel generator air start system. The flexible tubing (which was installed from the air compressors to the rigid piping of the diesel air start system) prevents damage to the rigid piping due to vibration of the air compressor skids.

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OCP No. 1149 Sumary of Safety Evaluation:- This modification was safety-related.

(Continued) The reliability of the diesel generator air start system was increased by lowering the maximtsn operating pressure and installing superior relief valves and regulated battery chargers. The original design intent of the diesel generator air start system was not affected by DCP 1149. The capacity of the system still meets the original design requirements described in section 8.3.1.1.2 of the Updated FSAR. Installation of instrument valves upstream of the oil booster tanks eliminated the necessity to disassemble and reassemble the piping during system testing. No unreviewed safety questions existed..

DCP No.1163 Personnel-Monitoring Stations Description and Basis for Change: Radiation protection procedures require that all personnel exiting contaminated areas survey themselves for radioactive contamination. The purpose of DCR 1163 was to install six additional personnel monitoring stations in the turbine and reactor buildings. Permanent walls and roofs were constructed to provide shielding to reduce the area background radiation levels. The walls were constructed of high density grout-filled concrete blocks with the roof of reinforced concrete on steel deck. Access to each station is through an open doorway.

Summary of Safety Evaluation: The added station structures do not house any safety-related items and are not essential to plant operation. The design of each station structure ensures that a possible, but unlikely collapse 9 the stations as a result of a seismic event would not impact ihe integrity of the surrounding building structure or any safety-related system components. Each structure will retain its integrity under load during Design Basis Earthquake (DBE) conditions. The enclosures had a negligible affect on the structural integrity of the existing buildings. No unreviewed safety questions existed.

DCP No.1180 Turbine Lube Oil Conditioner IT-39 Description and Basis for Change: The existing Turbine Lube Oil Purification system did not clean the oil adequately and was a frequent maintenance item. To improve the quality of the lube oil, a larger more efficient lube oil conditioner was instalied. The existing system was retained as a backup. The new conditioner can be operated alone or in parallel with the existing system.

Summary of Safety Evaluation: This modification was not safety-related. High quality turbine lube oil is essential for turbine and generator protection. The new system was installed in accordance with ANSI B.31.1, "Power Piping Code." The probability of a turbine trip due to the failure of the Turbine Lube Oil System is reduced, because the system is more reliable and has improved conditioning capabilities. Additior. ally, curbing was added to the lube oil tank area to accommodate the additional fluid capacity of the new system in the event of fluid leakage or spillage. The additional capacity was analyzed for its impact on the Fire Hazards Analysis and was found acceptable. No unreviewed safety questions existeJ.

m s DCP No.1186 Acid Pump Replacement & pH Control Modifications Description and Basis for Change: The sulfuric acid feed system is intended to control the acidity of the Circulating Water System.

The syst- that existeo before DCP 1185 was a gravity feed arrangement that performed erratically and had permitted water acidity to become high enough to cause corrosion damage to equipment in the Circulating Water System. DCP 1185 installed acid feed pinnps and improved controls to allow more precise control of the introduction of acid into the circulating water to maintain optimtsn pH. Additionally, DCP 1185 upgraded the sulfuric acid feed system piping to stainless steel. Stainless steel is better suited for handling acid beca/se of its excellent resistance to corrosion. DCP 1185 also added a catch basin for the sulfuric acid tank and a safety shower / eyewash station.

Summary of Safety Evaluation: DCP 1185 was not safety-related and does not interf ace with ar,y safety-related equipment. The installation of ccerosion resistant piping and better pH control increases the reliability of the Acid Feed System. The installation of a catch basin for the sulfuric acid tank and a safety shower / eyewash station improves personnel safety. The remote location of the tank and planps (outside the pumphouse) assures that safety-related equipment will not be affected in the event of a piping f ailure. The Circulating Water System is protected from low pH by interlocks that shut off the pumps and isolate the Acid Feed System should the Circulating Water System pH f all below 6. High pH alarms are provided although high pH is not damaging to equipment. No unreviewed safety questions existed.

DCP No. 1237 Replacement of Pressure Indicating Switch (PIS) 3887 Description and Basis for Change: Pressure indicating switches (PIS 3887 and 3901) monitored line pressure at the Collector Filter (1F-207) inlet in the Liquid Radwaste Treatment System. Before DCP 1237, jumpers were installed in control room panel IC96 to bypass both pressure inlet switches and indications because of previous concerns with the pressure switches either becoming inoperable or losing calibration. Design Engineering concluded that the pressure switches were not needed for the operation of this filter.

Therefore, the pressure switches and indicators were eliminated.

! Summary of Safety Evaluation: The removal of these switches did not adversely af f ect tne operation of tt e liquid radwaste system. No other control functions for the radwaste system were adversely affected by this modification. This change did not alter system operation as described or implicit in the Final Safety Analysis Report. No unreviewed safety questions existed.

DCP No.1238 Replace Recirculation System Flow Transmitters Description and Basis for Change: Flow transmitters FT-4631 A-D and FT-4632 A-0 for the reactor recirculation system had a history of zero shift drifting that caused sperious half-scrans and could have resulted in incorrect biasing of the scram setpoint. The existing transmitters were no longer manuf actured and were replaced with suitable substitutes.

DCP No.1238 S;mmary of Safety Evaluation: This modification was safety-related.

(c6ntinued) Tvaluation of the DAEC's temperature profile for the specific locations and applications in the specific loops, showed that the new transmitters were suitable replacements for the existing transmitters. The transmitters were punhased as Class 1E units and were qualified to IEEE-323 and IEEE-3/ . andards and no new f ailure modes were introduced. Seismic calculdions were performed to attest to the seismic mounting of the new transmitters. The instrtinent racks are Seismic Category I and have been seismically qualified (see FSAR Section 3.10.1.1). No um eviewed safety questions existed.

3 OCP No. 1245 Access Control Modifications Description and Basis for Change: OCP 1245 authorized the modification of specific areas of the grounc floor of the Administration Building. The modificatinns primarily involved the Access Control function of the building and resulted in more efficient traffic flow in and out of the reactor building.

New office and storage space was created, allowing more effective utilization of Health Physics personnel. The capability of process'.ng contaminated personnel and equipment was also improved, as was emergency equipment access and storage. These improvements were recommended in NRC Appraisal Audit Repcrt Number 50-331/80-21.

Additionally, this modification resulted in a significant improvement in housekeeping and appearance of the administration building. The scope of DCP 1245 included modifications to the existing architecture and pltrnbing as well as Heating, Ventilating and Air Conditioning (HVAS and electrical systems of the administrhtion building. The modification was essentially a reutilization of existing space and no t.hanges were made 'n the overall structural integrity of the building.

Sunmary of Safety Evaluation: This modification was not safety-related and did not 4fect any safety-related equipment or functions of tne plant. The ' attallation, use or f ailure of any modification made by DCP 1245 di Description and Basis for Change: This DCP evaluated and controlled the flapping process on eight weld overlays on the recirculation system. The weld overlays were originally completed .

under DCP 1310. However, new Electric Power Research Institute (EPRI) ~ standards for ultrasonic examinations require a smoother surface on the weld overlays. Flapping provided a smoother surface finish to produce a more reliable ultrasonic examination of the weld overlays.

- St.inmary of Safety Evaluation: This modification was safdty-related. The plant was in the cold shutdown condition when this modification was performed. The original design intent for each weld overlay was not changed by this modification. The flapping process was evaluated and controlled by this DCP to ensure that the integrity of each full-structural weld overlay was maint ained. Flapping these weld overlays improved their surface finish and increased the reliability of their ultrasonic examinations. No unreviewed safety questions existed.

OCP No.1390 Remote Shutdown Panel Structural Improvement t Description and Basis for Change: The structural modification of the remote shutdown panel (lC388) consisted of removing sections of 1/8" thick exterior plate in the vicinity of the corner joints, welding gusset plates directly to the angle iron structure, smoothing the surf ace contour for appearance, and painting all exposed metal surf aces. The addition of gusset plates and welds was necessary due to inadequate butt welds used in joining the structural members at the corner joints.

This deficiency was reported pursuant to 10CFR Part 21 in LER ,87-008.

Summary of Safety Evaluation: This modification was not safety-related. This design change ensured that the supporting structure of panel IC388 met the original seismic design. This modification did not affect any other systems or components. It  !

did not affect the design function of the remote shutdown panel "

but merely provided adequate seismic support for the panel. No unreviewed safety questions existed.

DCP No.1391 Replace Discs in Pressure Safety Valves PSV-4403 and PSV-4404 Description and Basis for Change: Disc damage was discovered during testing of main steam safety valves PSV-4403 and PSV-4404.

DCP 1391 was initiated to repair the discs in main stean safety valves PSV-4403 and PSV-4404. The disc in PSV-4403 was constructed of ASTM A565, Gr.616, condition "HT" material and was repaired. The disc in PSV-4404 was constructed of ASTM A-479, type 410 material. Rather than repair this disc, it was replaced with a disc constructed of ASTM A565 Gr.616, condition "HT"  ;

material, which was evaluated as a suitable replacement.  !

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.- n DCP No.1391 . Summary of Safety Evaluation: This modification was (continued) s afety-rel ated. It was performed when the plant was in the cold shutdown condition. The new disc did not affect the operation or function of the main steam safety valve. The replacement disc either meets or exceeds the requirements for the original disc as stated in the ASME Code,Section XI,1980 Edition Through Winter 1981 Addenda. The main steam safety valves were retested satisf actorily prior to their installation. No unreviewed safety ;

questions existed.  !

r DCP No.1392 Replace HGA Relays Description and Basis for Change: General Electric, in Service Advisory Letter (5AL) 174.1, identified a potential problem with -

its HGA relays under seismic conditions. Specifically, the contacts, which are closed while the relay is deenergized can chatter at horizontal accelerations above 0.35gs. The DAEC FSAR  !.

requires that all instrumentation required for nuclear safety must '

be qualified to seismic acceleration levels of 1.59s horizontal and 0.5gs vertical over a frequency range of 0.25 to 33 bz. The results of an indepth evaluation identified fourteen relays that required modification to prevent possible intermittent operation of a safety-related circuit during an accident condition while a seismic event is taking place. These fourteen HGA relays were r replaced with GE HFA series relays. The replacement HFA relays -

were seismically evaluated and were found acceptable. .

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! This design deficiency was reported pursuant to 10CFR Part 21 in l LER 87-021. .

Sumary of Safety Evaluation: This modification was

. safety-related. The FSAR does not specify a particular type of .

relay for a given application. Both the HGA and HFA relays are  ;

used in safety-related circuitry including the logic channels that  ;

were affected by this modification. The HFA relays met Class 1E l electrical qualifications and IEEE 344-1975 seismic

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qualifications. Installation of these new HFA relays in panels 1C30,1C32, IC33, IC43, and 1C44 did not affect the seismic

! qualification of che panels. The replacement HFA relays meet and i exceed the seismic qualification requirements of the DAEC FSAR  !

and perform the same functions electrically as the HGA relays. [

This modification returns the plant to a condition within its -

! original design basis. No unreviewed safety questions existed. l l

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m s SECTION B - PROCEDURE CHANGES During 1987, various procedures as described in the safety analysis report were revised and updated. All changes were reviewed against 10 CFR Part 50.59 by the DAEC Operations Comittee. No procedure changes were performed that constitute unreviewed safety questior.s.

All Special Test Procedures (SpTPs) performed in 1987 were also reviewed by the DAEC Operations Committee. No unreviewed safety questions were found to exist. Summaries of these special tests and their safety evaluations are found below.

TEST NO. TITLE / DESCRIPTION SpTP No. 129 Reactor Water Cleanup (RWCU) System Area Differential Temperatures The purpose of this test was to collect operating data on the temperature differential bet.een air entering the reactor water cleanup area and air exiting this area. This data was used to ensure that the setpoints for the steam leak detection logic were correct as per Technical Specification Table 3.2-A. The test also served to resolve an NRC open item, identified in NRC Inspection Report 50-331/86002 relating to the use of a plastic cover on door >

249 that restricted air flow into the reactor water cleanup system heat exchanger room.

This test was performed on March 2,1987.

Summary of Safety Evaluation: The operation of the plant during the test was within design requirements and Technical Specification limits. Additionally, the plant was protected from an unnecessary isolation of the RWCU system and, in the event of a RWCU system leakage condition, the Group V primary containment isolation function would have occurred per system design. At no time during this test were both sides of the RRCU steam leak detection system inoperable simultaneously. Based on the above, no unreviewed safety questions existed.

SpTP No. 131 Main Control Room Ventilation System Exhaust Isolation Damper Leakage Test In the event of a Loss of Coolant Accident (LOCA), the Control Room is isolated and required to be maintained at a positive pressure such that habitability can De maintained. The purpose of this special test was to determine at what positive pressure the control room atmosphere can be maintained during Standby Filter Unit (SFU) operation. Specifically, the test would determine whether air leakage past control building isolation damper 1V-AD-31A reduces the maximum attainable differential pressure inside the Control Room.

This test was performed on January 15, 1987.

Summary of Safety Evaluation: Since only one SFU train was activated during the test, one was always available to perform the required protective function during reactor operation. The use of one train during artual accident. conditions has previously been evaluated in the FSAR. Based on the above, no unreviewed safety questions existed.

TEST NO. TITLE / DESCRIPTION SpTP No. 132 Effect of Battery Room Exhaust Flow on Control Room Pressure In the event of a loss of Coolant Accident (LOCA), the Control Room is isolated and required- to be maintained at a positive pressure such that habitability can be maintained. The purpose of this test was to determine the effect of reduced battery room exhaust flow on the positive pressure in the Control Room during Standby Filter Unit (SFU) operation. By temporarily reducing the Battery Room Exhaust flow during the test, a substantial increase in control room positive pressure was achieved. A permanent change to the position of the battery room danper was not made and.

the damper was returned to its normal position upon completion of the test.

This test was pe' formed on February 11, 1987 and February 22, 1987 Summary of Safety Evaluation: Since only one SFU train was activated during the test, one was always available to perform the required protective function during reactor operation.

Additionally, the throttling of battery room air flow did not reduce the plant margin of safety since the 1-1/2 hours required to perform the test was significantly less than the time required to reach the hydrogen concentration by volume (in air) necessary to form an explosive mixture. Based on the sbove, no unreviewed safety questions existed.

SpTP No. 133 Motor-0perated Valve M0-2202 Opening Operability Testing At Abnormal Differential Pressure The purpose of this test was to denonstrate valve M0-2202 operability at its maximum-expected accident differential pressure of 1110 psid and to determine whether any adjustments or repairs were required. The valve operability demonstration at 1110 psid was based on an NRC Bulletin 85-03 requirenent.

This test was performed on May 23, 1987 Summary of Safety Evaluation: This test was performed while the reactor plant was in the cold shutdown condition. The test pressure was isolated from the reactor, High Pressure Coolant Injection (HPCI) turbine and other connected systems and interf acing piping was vented to atmosphere. Additionally, the test was performed while the HPCI system was shutdown for normal maintenance and normal design backup systems were still available.

This test did not effect other ECCS systems. The test pressure of 1110 psig was below the system design pressure of 1365 psig and relief protection was provided on the hydrostatic test rig (pressure source). Based on the above, no unreviewed safety questions existed.

% s TEST N0. TITLE /0ESCRIPTION SpTP No. 134 Motor-0perated Valve M0-2312 Opening Operability Testing at Abnormal Differential Pressure The purpose of this test was to demonstrate valve M0-2312 operability at its maximum-expected accident differential pressure of 1289 psid. The valve operability test at 1289 psid was performed to meet an NRC requirement in Bulletin 85-03 to denonstrate valve operability at its maximum-expected accident differential pressure and to determine whether any adjustments or repairs were required.

This test was performed on May 28, 1987.

Sumary of Safety Evaluation: This test was performed while the reactor plant was in the cold shutdown condition. The test pressure was isolated from the reactor, High Pressure Coolant Injection (HPCI) pump and Condensate Storage Tank (CST) and other connected systens and interf acing piping was vented to atmosphere.

Additionally, the test was performed while the HPCI syste,n was shutdown for normal maintenance and normal design backup systems were still available. This test did not effect other ECCS systems. The test pressure of 1289 psig was below the maximum service condition pressure of 1590 psig and relief protection was provided on the hydrostatic test rig (pressure sourca). Based on the above, no unreviewed safety questions existed.

SpTP No. 135 Motor Operated Valve M0-2317 Opening Operability Testing At Abnormal Differential Pressure The purpose of this test was to demonstrate valve M0-2318 operability at its maximum-expected accident differential pressure at 1416 psid. The valve operability test at 1416 psid was performed to meet an NRC requirement in Bulletin 85-03 to demonstrate valve operability at its maximum-expected accident differential pressure and to determine whether any adjustments or repairs were required.

This test was performed on May 23, 1987.

Summary of Safety Evaluation: This test was performed while the reactor plant was in the cold shutdown condition. The test pressure was isolated from the reactor and the High Pressure Coolant injection (HPCI) pump and other connected systems and interf acing piping was vented to atmosphere. Additionally, the test was performed while the HPCI system was shutdown for normal maintenance and normal design backup systems were still available.

This test did not effect other ECCS systems. The test pressure of 1416 psig was below the maximan service condition pressure of 1590 psig and relief protection was provided on the hydrostatic test rig (pressure source). Based on the above, no unreviewed safety questions existed.

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- s TEST NO. - TITLE / DESCRIPTION SpTP No. 136 Motor Operated Valve M0-2322 Opening Operability Testing At Abnormal Differential Pressure The purpose of this test was to demonstrate valve M0-2322 operability of its maximum-expected accident differential pressure of 123 psid. The valve operability test at 123 psid was performed to meet an NRC requirenent in Bulletin 85-03 to demonstrate valve operability at its maximum-expected accident differential pressure and to determine whether any adjustments or repairs were required.

This test was performed on March 30, 1987.

Summary of Safety Evaluation: This test was performed while the reactor plant was in the cold shutdown condition. The test pressure was isolated from the reactor and other connected systems and interf acing piping was vented to atmosphere.

Additionally, the test was performed while the HPCI system was shutdown for normal maintenance and normal design backup systems were still available. This test did not effect other ECCS systems. The test pressure of 123 psig was below the initial system hydrostatic test pressure of 188 psig and relief protection was provided on the hydrostatic test rig (pressure source). Based on the above, no unreviewed safety questions existed.

SpTP No.137 Motor Operated Valve 60-2512 Opening Operability Testing At Abnormal Differential Pressure The purpose of this test was to demonstrate valve MO-2512 operability at its maximum-expected accident differential pressure at 1417 psid. The valve operability test at 1417 psid was performed to meet an NRC requirement in Bulletin 85-03 to demonstrate valve operability at its maximum-expected accident differential pressure and to determine whether any adjustments or repairs were required.

This test was performed on May 28, 1987.

Summary of Safety Evaluation: This test was performed while the reactor plant was in the cold shutdown condition. The test pressure was isolated from the reactor and other connected systems and interfacing piping was vented to atmosphere. Additionally, the test was performed while the RCIC system was shutdown for normal maintenance and normal design backup systems were still available. This test did not effect other ECCS systems. The test pressure of 1417 psig was below the maximum service condition pressure of 1460 psig and relief protection was provide 6 on the hydrostatic test rig (pressure source). Based on the above, no unreviewed safety questions existed.

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_ TEST NO. TITLE /0ESCRIPTION

'SpTP.No._i38 Motor Operated Valve M0-2517 Opening Operability Testing At Abnormal Differential Pressure The purpose of this test was to demonstrate valve M0-2517 operability at its maximum-expected accident differential pressure of 128 psid. -The valve operability test at 128 psid was performed to meet an NRC requirement in Bulletin 85-03 to denonstrate valve operability at its maximum-expected accident differential pressure and to determine whether any adjustments or repairs were required.

This test was performed on April 10, 1987.

Summary of Safety Evaluation: This test was performed while the reactor plant was 'in cold shutdown condition. The test pressure was isolated from the reactor and other connected systens and interf acing piping was vented to atmosphere. Additionally, the test w&s performed while the RCIC system was shutdown for normal maintenance and normal design backup systems were still available.

This test did not effect other ECCS systens. The test pressure of 128 psig was below the initial system hydrostatic test pressure of 188 psig and relief protection was provided on the hydrostatic test rig (pressure source). Based on the above, no unreviewed safety questions existed.

SpTP No. 139 Procedure for Low-Amplitude Testing Of Class 1E Control Room Panels The purpose of this test was to determine the dynamic properties of the control panels in the control room. In order. to evaluate the control panel anchorage and the seismic adequacy of the Class 1E equipment mounted in the control panels, the control panels were subjected to low-amplitude vibrational forces by the employnent of a mechanical shaker.

This test was performed during the time period of March 31, 1987 through April 8, 1987.

Summary of Safety Evaluation: The maximum input from the mechanical shaker to the panel did not exceed the previously evaluated values for the vertical and horizontal accelerations of the 786'-0" floor elevation in the control building during the design basis earthquake. Dynamic anplification of the input motion did not exist because the input motion of the shaker was applied at (or near) the top of the control panels. Additionally, random vibration input to seismic category I equipment had been previously evaluated and enveloped by the existing analysis discussed in the FSAR. Based on the above, no unreviewed safety questions existed.

TEST NO. TITLE / DESCRIPTION SpTP No.140 ANI 5-Year Diesel Fire Pump Operability Test The purpose of this test was to denonstrate the operability of, and verify the capacity of, the diesel-driven fire pump.

This test is required by the Anerican Nuclear Insurers (ANI).

This test was performed on May 20, 1987.

Summary of Safety Evaluation: The performance of this test was not safety related and did not alter the function or operation of any safe shutdown equipment. This test was performed outdoors so as not to damage any safety-related equipment. There is no cross-connection between the fire protection water and any potentially contaminated systems. The river water supply p2nps renained operable throughout this special test procedure such that intake basin level remained above the low limits. Based on the above, no unreviewed safety questions existed.

SpTP No. 141 ANI 5-Year Electric Fire Pump Operability Test The purpose of this test was to demonstrate the operability of, and verify the capacity of, the electric motor-driven fire pump.

This test is required by the American Nuclear Insurers (ANI).

The test was performed on May 20, 1987.

Summary of Safety Evaluation: The perform ece of this test was not safety-related and did not alter the functinn or operation of any safe shutdown equipnent. This test was performed outdoors so as not to damage any safety-related equipment. There is no cross-connection between the fire protection systen and any potentially contaninated systems. The river water supply pumps renained operable throughout this special test procedure such that intake basin level remained above the required limits. Based on the above, no unreviewed safety questions existed.

SpTP No.143 Motor Operated Valve MD-2316 Operability Testing at Normal Differential Pressure of 300 PSIG and 1200 PSIG.

The purpose of this test was to denonstrate the operability of valve M0-2316 and to determine the valve operator thrust requirements for full pressure operation. This test was performed to meet an NRC requirement in Bulletin 85-03.

This test was performed on July 1,1987 Sumary of Safety F ealuation: The test was performed while the High Pressure Coolant Injection (HPCI) system was operating for normal surveillance testing and nonnal design backup systens were av ail able. The maximum test pressure of 1200 psig was below the system maximan service condition pressure of 1590 psig. The test pressure was isolated from the reactor and other connected systems. This test did not ef fect other ECCS systems. The test was performed within the design capabilities of the HPCI system during normal plant surveillance testing. Based on the above, no unreviewed safety questions existed.

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TEST NO. TITLE / DESCRIPTION

-SpTP No. 145 Containment Spray logic Functional Test of Selected Relay

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-Contacts.

The purpose of this test was to denonstrate operability of all initiation logic components and contacts of the containment spray subsysten of the Residual Heat Removal (RHR) systen that were not normally tested under the scope of the normal Surveillance Test Procedure (STP) program. This test was performed in response to NRC Inspection Report 50-331/87004.

This test was performed on June 5,1987.

Summary of Safety Evaluation: The function and operation of the RHR systen was not altered by the special test procedure. The test involved testing certain contacts of the systen logic by use of jumpers and/or test switches to simulate closed contacts where open contacts existed due to there being no actuation / initiation signal. This test ensured all contacts in the logic circuits were tested in accordance with the original design intent. Precautions were taken such that no valve cycling occurred, and the logic was returned to its pre-test configuration upon test completion.

Based on the above, no unreviewed safety questions existed.

SpTP No. 146 LPCI Trip System Logic Functional Test of Selected Relay and contacts

- The purpose of this test was to denonstrate operability of all Low Pressure Coolant Injection (LPCI) system logic components and contacts that were not tested under the scope of the normal Surveillance Test Procedure (STP) program. Specifically, verification of the opening and closing of all contacts utilized in the LPCI mode of the Residual Heat Removal (RHR) system logic was performed. This test was performed in response to NRC .

Inspection Report 50-331/87004.

This test was performed on June 5,1987 Summary of Safety Evaluation: The function and operation of the RHR system was not altered by the special test procedure. The test involved testing certain contacts of the systen logic by use of jumpers and/or test switches to simulate closed contacts where open contacts existrd due to there being no actuation / initiation si gnal . This test ensured all contacts in the logic circuits were tested in accordance with the original design intert. Precautions were taken such that no valve cycling occurred, and the logic was returned to its pre-test configuration upon conpletion of the test. Based on the above, no unreviewed safety questions exist ed.

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i TEST NO. . TITLE / DESCRIPTION SpTP No. 147 Core Spray Trip System Logic Functional Test of Selected Relays- ,

and Contacts The purpose of this test was to demonstrate operability of Core Spray system initiation logic components and contacts that were not normally tested under the scope of the . normal Surveillance Test Procedure (STP) program. Specifically, verification of the opening and closing of contacts within the core spray logic was performed . This test was performed'in response to NRC Inspection Report 50-331/87004.

This test was performed on June 4,1987.

Summary of Safety Evaluation: The function and operation of the Core Spray system was not altered by the special test. procedure.

The test involved testing certain contacts of the system logic by use of jumpers and/or test switches to simulate closed contacts where open contacts existed due to there being no actuation /

initiation signal. This test ensured all contacts in the logic circuits were tested in accordance with the original design intent. Precautions were taken such that no valve cycling occurred, and the logic was returned to its pre-test configuration upon completion of the test. Based on the above, no unreviewed safety questions existed.

SpTP No.148 Feedwater Level Set Test This test provided a mechanism for establishing the optimun liquid level in the 'ow pressure feedwater heaters at the OAEC. Because of recent poblems experienced at the DAEC with vibration induced failures of feedwater heater tubes, it was recommended that the liquid level be raised. Raising the liquid level in a feedwater heater will preclude the occurance of steam entrainment in the drain cooler zone of a feedwater heater, and thus eliminate a root cause of the vibration damage mentioned above.

Establishing an optimum feedwater heater liquid level will improve plant heat-rate, since there is an optimun level where extraction stean can provide the most benefit in preheating the feedwater.

This test was performed during the time period of July 17, 1987 through August 6, 1987.

Sumary of Safety Evaluation: The consequences of a malfunction of a feedwater heater level control are a loss of a feedwater heater or a turbine trip. Both of these events have been previously evaluated in the Updated FSAR and bound any event that could be caused by the performance of this test. Feedwater heater liquid levels were varied in only one feedwater heater at a time

\'TESkN00 TITLE / DESCRIPTION SpTP No. 148 and the incremental' level changes made were small relative to (continued) level variances experienced during operational transients (the test was performed at constant reactor. pressure levels). The original feedwater heater liquid level was restored upon l completion of the test. Based on the above, no unreviewed safety _  !

questions existed.

SpTP No. 150 HPCI System to CV2315, Test Valve Data The purpose of this test was' to demonstrate and record the valve position of CV2315 and the turbine speed for various High Pressure Coolant Injection (HPCI) flows 'and pressures. Additionally, this test provided' data for the monthly and quarterly surveillance test procedure to meet an NRC commitment to incorporate cold quick starts of the HPCI turbine into the applicable STPs.

This test was performed on November 17, 1987 Summary of Safety Evaluation: Performing this special test did not adversely affect the HPCI system. This special test required operation of the HFCI pump above its normal flow of 3000 gpm, but below its maximum rated flow at maximum rated turbine speed (3900 rpm). The purnp discharge pressure was varied above and below the normal pressure, but below the rated pressure at the maximum rated 1 turbine speed. The auto-initiation function of the HPCI was unaffected by this test. The probability of an inadvertent HPCI

- injection was unchanged by following the normal Surveillance Test l

Procedure (STP) for valve manipulation. Based on the above, no unreviewed safety questions existed.

SpTP No.151 RCIC System to M02515. Test Valve Data The purpose of this test was to demoristrate and record the valve position of M02515 and the turbine speed for various Reactor Core Isolation Cooling (RCIC) pump flows and pressures. Additionally, this test provided data for the monthly and quarterly surveillance test procedure to meet an NRC comitment to incorporate cold quick starts of the RCIC turbine into the applicable STPs.

4 This test was performed on Ncvember 13, 1987 Surunary of Safety Evaluation: Performing this special test did l not adversely affect the RcIC system. This special test required operation of the RCIC pump above its normal flow of 400 gpm, but below its maximum rated flow at maximum rated turbine speed (4500 rpm). The pump discharge pressure was varied above and below the normal pressure, but below the rated pressure at the maximum rated turbine speed. The auto-initiation function of the RCIC system was unaffected by this test. The probability of an inadverttint RCIC injection was unchanged by following the normal Surveillance Test Procedure (STP) for valve manipulation. Based on the above, no unreviewed safety questions existed.

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SECTION C LEXPERIMENTS -

This seitionihas'been prepared in accordance with.the requirements of 10 CFR Part 50.59(b). No experiments were conducted during calendar year.1987 h

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SECTION 0 - SAFETY AND RELIEF VALVE FAILURES AND CHALLENGES

' This section contains information concerning relief valve and safety valve failures and challenges for calendar- year 1987 in accordance with the requirenents of Technical Specification 6.11.1.e. Note that any instance in which a main steam relief or safety valve was manually cycled open, for surveillance testing or other reasons, is included for your information. There were no safety valve failures or challenges during 1987. There were no relief valve failures during 1987. There was one relief valve challenge during 1987. This event is described below:

Date Event Description June 29, 1987 . Relief valves PSV-4400, -4401, -4402, -4405, -4406 and -4407 were opened and closed during the satisf actory completion of a normal surveillance test.

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