ML17326B548

Annual Operating Rept,1988. W/890228 Ltr

ML17326B548
Person / Time
Site:	Cook
Issue date:	12/31/1988
From:	Lauren Gibson, Will Smith, Svensson B INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 8903100329
Download: ML17326B548 (21)

Text

ACCELERATED DIUBUTlON DEMONSTR!0.i SYFTEkl REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:8903100329 DOC.DATE: 88/12/31 NOTARIZED: NO DOCKET FACIL:50-315 Donald C. Cook Nuclear Power Plant, Unit 1, Indiana & 05000315 50-316 Donald C. Cook Nuclear Power Plant, Unit 2, Indiana & 05000316 AUTH. NAME AUTHOR AFFILIATION STEVENSON,B.A. Indiana Michigan Power Co. (formerly Indiana & Michigan Ele GIBSON,L.S Indiana Michigan Power Co. (formerly Indiana & Michigan Ele SMITH,W.G. Indiana Michigan Power Co. (formerly Indiana & Michigan Ele R RECIP.NAME RECIPIENT AFFILIATION I

SUBJECT:

"Annual Operating Rept,1988." W/890228 ltr.

DISTRIBUTION CODE: IE47D COPIES RECEIVED:LTR ENCL SIZE:

TITLE: 50.59 Annual Report of Changes, Tests or Experiments Made W/out Approv S NOTES:

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD3-1 LA 1 0 PD3-1 PD 5 5 STANG,J 1 0 INTERNAL: AEOD/DOA 1 1 AEOD/DS P/TPAB 1 1 NRR/DLPQ/HFB 10 1 1 NRR/DOEA/EAB 11 1 1 N QDRBPp4 B 10 2 2 NUDOCS-ABSTRACT 1 1 LE 02 1 1 RGN3 FILE 01 1 1 EXTERNAL: LPDR 1 1 NRC PDR 1 1 NSIC 1 1 pgg; gysz/zest JEST 5~aw, 8 l I R

I NOTE TQ ALL "RIDS" RECIPZEZFS'IZASE HELP US V3 REDUCE HASTE! CDNZACr 'IHE DOCUME,'iZ CONTROL DESK, ROOM Pl-37 (EXT. 20079) K) ELIKCNATE YOUR NAHE FBCH DIST!KBVZIGN LISTS POR DOCUMENTS YOU DClN'T NEED!

TOTAL NUMBER OF COPIES REQUIRED: LTTR a/

~ ENCL lf

DONALD C.COOK NUCLEAR PLANT ANNUAL OPERATING REPORT 1988 COMPILED BY: ~c'. ~~

B.A. Svensson Licensi 'vities Coordinator REVIEWED BY:

L. . xbson Asst. Plant Mgr.-Technical Support APPROVED BY: UM r".J~

W.G. Smith, Jr.

Plant Manager ))(nS>

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TABLE OF CONTENTS TITLE PAGE NUMBER Introduction Personnel Exposure Summary Steam Generator Inservice Inspection Reports Changes to Facility Changes to Procedures 15 Challenges to Pressurizer PORVS and Safety Valves 16 Irradiated Fuel Inspection

INTRODUCTION The Donald C. Cook Nuclear Plant, owned by Indiana Michigan Power Company is located five miles north of'Bridgman, Michigan and consists of two nuclear power units. Each unit employs a pressurized water reactor nuclear steam supply system furnished by Westinghouse Electric Corporation.

The Unit 1 reactor is currently designed for a power output of 3250 MWt and the Unit 2 reactor is designed for a power output. of 3411 MWt, which are their licensed ratings. The approximate gross and net electrical outputs of Unit 1 are 1056 MWe and 1020 MWe and of Unit 2 are 1100 MWe and 1060 MWe, respectively. The main condenser cooling

. method is open cycle using Lake Michigan water as the cooling source.

The Cook Nuclear Plant was the first domestic nuclear facility to employe the ice condenser reactor containment system. The American Electric Power Service Corporation was the architect-engineer and constructor.

This Report was compiled by B.A. Svensson with the following individuals contributing information as follows:

D.C. Loope Personnel Exposure Summary C.A. Freer Steam Generator ISI Summary S.D. DeLong Changes to Facility J.B. Droste Changes to Procedures B.K. Worm Challenges to Pressurizer PORVs and Safety Valves R.W. Hennen Results of Irradiated Fuel Inspections Page 1

ANNUAL OPERATING REPORT RG 1.16 for 1988

/I PERSONNEL > 100mR TOTAL MAN-REM STAT. UTIL. CONT. STATION UTILITY CONTRACT REACTOR OPERATIONS & SURVEILLANCE Maintenance Personnel 0007 0000 0010 0000.971, 0000.000 0003.062 Operations Personnel 0056 0001'000 0033 0015.451 0000.155 0013.699 Health Physics Personnel 0018 0061 000>.353 0000.000 0022.952 Supervisory Personnel 0000 0000 0001 0000.000 0000.000 0000.130 Engineering Personnel 0010 0000 0004 0001.822 0000.000 0000.674 ROUTINE MAINTENANCE Maintenance Personnel 0101 0002 0241 0044.339 0000.267 0155,889 Operations Personnel 0019 0001 0022 0005.910 0000.338 0006.246 Health Physics Personnel 0006 0000 0013 0001.205 0000.000 0005.106 Supervisory Personnel 0002 0000 0001 0000.285 0000.000 0000.244 Engineering Personnel 0003 0003 0000 0000.357 0000.536 0000.000 IN-SERVICE INSPECTION Maintenance Personnel 0005 0000 0018 0000.954 0000.000 0003.448 Operations Personnel 0003 0001 0014 0000.630 0000.650 0008.586 Health Physics Personnel 0003 0000 0003 0000.494 0000.000 0000.929 Supervisory Personnel 0001 0000 0000 0000.170 0000 000 F 0000.000 Engineering Personnel 0000 0000 0001 0000.000 0000.000 0000.170 SPECIAL MAINTENANCE Maintenance-Personnel 0011 0031 0680 0001.722 0028.997 0541.633 Operations Personnel 0020 0010 0011 0004.961 0004.987 0010.332 Health Physics Personnel 0000 0000 0079 0000.000 0000.000 0103.184 Supervisory Personnel 0010 0010 0012 0001.102 .

0000.432 '0004.257 Engineering Personnel 0010 0010 0020 0000.145 0003.130 0022.918 WASTE PROCESSING Maintenance Personnel 0001 0000 0054 0000.620 0000.000 0019.097 Operations Personnel 0000 0000 0004 0000.000 0000.000 0002.429 Health Physics Personnel 0000 0000 0017 0000.000 0000.000 0004.597 Supervisory Personnel 0000 0000 0000 0000.000 0000.000 0000.000 Engineering Personnel 0001 0000 0001 0000.170 0000.000 0000.175 REFUELING Maintenance Personnel 0004 0000 0018 0000.528 0000.000 0008.664 Operations Personnel 0010 0001 0066 0004.750 0000.106 0033.389 Health Physics Personnel 0002 0000 0014 0000.205 0000.000 0005.743 Supervisory Personnel 0000 . 0000 0000 0000.000 0000.000 0000.000 Engineering Personnel 0001 0001 0000 0000.121 0000.151 0000.000 TOTALS Maintenance Personnel 0102 0002 0324 0047.532 0000.267 0192.357 Operations Personnel 0073 0002 0106 0026.741 0001.249 0064.458 Health Physics Personnel 0018 0000 0079 0009.257 0000.000 0039.327 Supervisory Personnel 0003 0000 0001 0000.455 0000.000 0000.484 Engineering Personnel 0011 0004 0005 0002.470 0000.687 0001.019 GRAND TOTALS 0511 0079 1913 0172.910 0004.406 0595.290 Page 2

1 '6 REPORT - WORK FUNCTION CATEGORIES Reactor 0 rations and Surveillance l

Those activities involved with operating the plant or monitoring it's operation, including chemistry, performance testing, plant may be at any power level, surveillance testing, etc. The including zero, and still have work falling into this area. Many STP's run during shutdown or refueling may still fall into this category.

Routine Maintenance All equipment or system maintenance, whether preventative or restorative, which does not involve significant modifications to equipment or systems. Included is I&C repair work, as well as work to adjust operable equipment to improve performance (adjusting fan blade pitch, for example).

Inservice Ins ction Inspections of 'equipment and systems to monitor changes that would be detrimental to function or integrity. Also included is all work required to permit such inspections, such as building required scaffolding, removing or replacing supports of insulation, or disassembly of valves, pumps, etc. Not included are inspection to

.assess or monitor normal wear, etc. For example, dissembly of a charging pump to inspect bearing wear would not be Inservice Inspection, but dissembly to inspect for rotor cracking or casing damage would be. Inspection of a weld on a newly added line is Special Maintenance, or inspection of a weld repair at a leaking fitting is Routine Maintenance.

S ecial Maintenance All work on equipment or systems performed to make significant modifications. Installation of new systems or equipment, replacement or addition of supports or hangers, addition of new lines or instruments, removal of existing equipment, replacement of existing. equipment. with significantly different equipment are all Special Maintenance. For example, replacement of a properly functioning, original equipment pressure transmitter with a different model with improved characteristics or certification would be Special Maintenance, but replacement or a malfunctioning pressure transmitter with a newer or improved model would probably be Routine Maintenance.

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Waste Proceaiin All work associated with decontamination of equipment, areas, systems, etc. (if not an integral part of another job, such as processing of waste, whether solid, pump repair), collection and liquid, or gas. Operations in support of waste handling are also included. For example, draining a filter to permit changing it, or venting it after changing are part of Waste Processing, but valving a clean filter into the system is Reactor Operations.

Repair of the Baler or drumming room crane is Routine Maintenance.

Refuelin All work is directly concerned with refueling the reactor, including all support operations, is classified as Refueling.

Testing the polar crane or installing the cavity filter rig is part of Refueling, as is cavity decon before or after flood-up.

Changing the cavity filter, however, is is Waste Processing Routine Maintenance.

and fixing the manipulator crane Page 4

STEAM OPERATOR TUBE INSERVICE INSPECTION REPORTS 1988

SUMMARY

REPORTS UNIT NO. 1 There were no inservice inspections of Unit No. 1's steam generators for the year 1988.

UNIT NO. 2 Unit No. 2 was removed from service on April 23, 1988, for the complete replacement of the four steam generator lower assemblies. A complete "preservice inspection" of the new lower assemblies was performed following the field hydrostatic test and prior to initial operation using equipment and techniques expected to be used during subsequent inservice inspection, pursuant to the requirements of Technical Specification 4.4.5.4.a.9.

There were no reportable indications identified during the preservice inspection.

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REQUEST FOR CHANGE CHANGES TO FACILITY Brief descriptions and summary safety'valuations for design changes (RFCs) made to the facility as described in the Donald C. Cook Nuclear Plant Final Safety Analysis Report (FSAR) are presented in this section.

These changes were completed pursuant to the provisions of Title 10, Code of Federal Regulations subsection 50.59 (a).

RFC-12-2859 BRIEF DESCRIPTION RFC-DC-12-2859 provided for the replacement of the South Boric Acid Evaporator Steam Coil Tube Bundle. Eddy current testin'g of the original tubes indicated severe pitting of the tubes. This pitting was most likely caused by Chlorides present in the evaporator bottoms. Although originally installed as a Boric Acid Evaporator, this evaporator is currently being used to process radwaste which has much higher chloride concentrations that can be expected in boric acid service. Based on recommendations from the Vendor, the tube bundle was replaced with a tube bundle manufactured from Incoloy 825 material. It is expected that the Incoloy 825 will perform better than the originally supplied 304SS bundle because of its superior resistance to chloride pitting.

SAFETY EVALUATION This RFC has been classified as Safety-Interface since the South Boric Acid Evaporator is a Seismic Class II component.

Nuclear Safety and Licensing has reviewed t'e change as per the review criteria in NS&L procedure No. 7. As a result of the Safety Review, there were no open items for this RFC. The Incoloy material was determined to have higher allowable stress limits than the stainless steel material.

The purpose of this review was for procurement, design and installation.

It was concluded, by the review, that this RFC did not constitute an unreviewed safety question as defined in 10 CFR 50.59, nor did it create a substantial hazard to the health and safety of the public.

RFC-12-2908 (Addendum f/3)

BRIEF DESCRIPTION RFC-DC-12-2908 (Addendum Pi3) provided for the installation of a new 150 ton/

20 ton Auxiliary Building Crane and bridge system. This crane was installed on the existing crane rails and was used in tandem with the existing crane that was modified under RFC-DC-12-2962 (also included in this submittal). Together, these cranes were used to lift and move the steam generators through the auxiliary building during the Steam Generator Replacement Project.

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Section 9.7 of the FSAR has been revised to incorporate the changes made to the auxiliary building. crane. Additional information concerning the cranes may be found in that section of the FSAR.

SAFETY EVALUATION This RFC has been classified as safety related because it involves modifications to the auxiliary building crane which is Seismic Class I equipment.

In addition to the safety evaluation provided for RFC-DC-12-2962, it was necessary to evaluate the consequences of a load drop during the installation of the crane components. Items covered were: (1) the individual crane components while they were being lifted by a boom crane, and (2) inadvertently dropping the boom crane components onto the auxiliary building structural elements or inadvertently hitting the auxiliary building structural elements. These evaluations were performed by the AEPSC Structural Design section.

Nuclear Safety and Licensing has reviewed these evaluations and found them to be acceptable with the following comments on the boom evaluation:

a) The boom crane has a capacity of 600K and the maximum weight lifted was that of the trolley at 144K. The safety factor available was 4.17. This was reasonably high for handling an occasional load.

It was noted that, in order to avoid any type of inadvertent human error, an additional operating engineer was available who also worked as an "oiler." It was recommended that a dedicated operating engineer be posted without any other assignments that would demand his attention.

c) After discussion with the AEPSC Material Handling Division, it was Nuclear Safety and Licensing's understanding that

,the maximum load lifted during the crane installation activity was 72 tons and the boom crane was certified to carry a test load of 110K of the maximum load as per ANSI B30.5. After the boom crane was installed at the site, an installation certificate was issued to document that the boom crane was installed per the manufacturer's guidelines.

d) The procedures for the crane installation were reviewed and approved by the appropriate engineering disciplines in AEPSC.

Based on the evaluation noted above, it is concluded that the installation of the cranes does not constitute an unreviewed safety question as per 10 CFR 50.59 Section (a)(2) and will not adversely affect the health and safety of the public.

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RFC-12-2962 BRIEF DESCRIPTION RFC-DC-12-2962 provided for modifications to the auxiliary building crane in order to conform to the single-failure-proof requirements of NUREG-0554. Specifically, the modification involved (a) replacing the original trolley (150T/20T) with a new trolley (150T) designed and built to single-failure-proof (SFP) requirements, (b) adding a second holding brake and an inching mechanism to the bridge drive, and (c) upgrading the crane runway girder in the auxiliary building to resist the higher wheel loads.

These modifications were performed for two reasons: (1) The crane must

=meet the requirements of NUREG-0612 and 0554 in regard to handling of heavy loads in the auxiliary building. (2) The single-failure-proof features of the modified crane were necessary for the movement of steam generators in and out of the auxiliary building during the S/G replacement project.

Section 9.7 of the FSAR has been revised incorporating the changes made to the auxiliary building overhead crane. Reference the revised text in Section 9.7 should any additional information be desired.

SAFETY EVALUATION The auxiliary building crane is a Seismic Class I component which performs various Safety-Related activities such as opening/closing of the containment equipment hatches and moving new and spent fuel assemblies and lifting and transporting steam generators during their replacement activities. Therefore, this RFC was classified as Safety-Related.

The safety memo specifically addressed the following items related to the crane modifications:

1. The modifications to the crane as noted in the original RFC have been completed. Specifically, (a) the new trolley has been designed and fabricated to meet the single-failure-proof requirements of NUREG-0554, (b) a second holding brake and an inching mechanism has been added to the bridge drive and (c) the crane runway girder has been modified to take the higher wheel loads.

2. KS&L has reviewed the stress/seismic analysis of the modified crane performed by Whiting and found it to be acceptable.

h

3. AEPSC Cognizant Engineers have visited the Whiting offices to review the vendor documents as per the requirements of specification DCC-MH-105-QCN and accepted the new trolley.

4. The auxiliary building crane is a seismic Class 1 component and all modifications to the crane have been procured and installed to meet the seismic Class 1 requirements.

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Based on the evaluation described above, it is concluded that the design modification performed to the auxiliary building crane does not constitute an unreviewed safety questions as per 10 CFR 50.59, Section (a) (2) and that it will not adversely affect the health and safety of the public.

The NRC has reViewed the modific'ations to the auxiliary building crane in amendment 100 to facility operating license No. DPR-74. Their concurrence reads as follows:

"Based on the licensee's demonstration of compliance to the guidelines of NUREG-0554 and the adequacy of the crane's structural component in meeting their allowable stress values, the staff finds that the proposed new crane installation is acceptable."

RFC-DC-12-4042 BRIEF DESCRIPTION RFC-4042 provided for the installation of an additional level indicator and low-level alarm for the Volume Control Tank (VCT) level control system. These modifications were installed to eliminate an undesirable situation should one of the two (2) originally installed VCT level channels fail.

A failure of the capillary reference leg on the VCT Level Controller (QLC-452) will cause the instrument to fail high. This high level signal causes the VCT Divert Valve (QRV-303) to open in an effort to reduce and maintain the VCT level within a normal operating band. As long as the Redundant VCT Level Controller (QLC-451) functioned normally, no alarm would have sounded.

This scenario had the potential of allowing the VCT to be pumped down until the charging pumps lost suction. The additional low level alarm added under this RFC is fed from QLC-451. This alleviates any concerns regarding the scenario described above.

SAFETY EVALUATION This design change was an NRC commitment to provide additional VCT level instrumentation to aid plant operations. Since this modification will enhance the safety function of the VCT, this RFC has been classified as "Safety Interface".

This RFC has been reviewed in accordance with NS&L procedure number 7 "Safety Review of Design Changes". Subsequent conversations with the IGC engineer permitted a conclusion that there were no open items with regard to RFC-DC-12-4042 and the modifications described above did not constitute an unreviewed safety question as defined in 10 CFR 50.59.

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MINOR MODIFICATIONS CHANGES TO FACILITY Brief descriptions and summary safety evaluations for design changes (Minor Modifications) made to the facility as described in the Donald C.

Cook Nuclear Plant Final Safety Analysis Report (FSAR) are presented in this section. These changes were completed pursuant to the provisions of Title 10, Code of Federal Regulations subsection 50.59 (a).

12-MM-010 REMOVE WASTE EVAPORATOR FILTER ELEMENTS BRIEF DESCRIPTION This Minor Modification removed the Waste Evaporator Filter elements from the Waste Evaporator Feed Filters. Most liquid radwaste at the plant is currently treated with the Duratek demineralization system rather than the radwaste evaporators. Although Duratek has its own filters prior to the demineralizers, wastes are currently also being passed through the waste evaporator filters. It has been determined that this extra filtration is not needed to obtain adequate clean-up with the Duratek system. The waste evaporator filters are difficult to change and result in worker exposures that are not consistent with the principles of ALARA.

SAFETY EVALUATION This change is classified as Safety Interface because it involves a system that handles radioactive wastes.

A technical evaluation was performed by the Chemical Engineering section.

Specifics of the safety review for the modifications are provided below.

1. The filter units are designated seismic Class II. Operation without the removable filter elements in place in the filter housing will not degrade the seismic rating of the piping.

2. All liquid effluents are routed to the monitor tanks, where they are sampled prior to discharge to the lake. The sampling ensures that the effluents will not violate 10 CFR 20 or Technical Specification limits. Thus, even if cleanup capability is reduced by removal of the filter elements, it would not result in releases exceeding those limits.

3. The waste evaporator filters are discussed in Chapter lj.l of the FSAR.

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The safety evaluation concluded that removal of the waste evaporator filter elements does not constitute and unreviewed safety question as described in 10 CFR 50.59, "Changes, Tests and Experiments," Section (a) (2) and that it does not significantly impact public health and safety.

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TEMPORARY MODIFICATIONS CHANGES TO FACILITY Brief descriptions and summary safety evaluations for Temporary Modifications made to the facility as described in the Donald C. Cook Nuclear Plant Final Safety Analysis Report (FSAR) are presented in this section. These changes were completed pursuant'o the provisions of Title 10; Code of Federal Regulations subsection 50.59 (a).

Temporary Mod. 819 (Unit 2)

BRIEF DESCRIPTION This Temporary Modification involves the lifting of Cables 3167C-2, 3168-1, 3153C-2, 3138-2 and 3139-2 which is the power feed for pressurizer heaters f/3, f/4, f!53, k/55 and /f56, respectively. These pressurizer heaters have a defective heater element. By disconnecting these cables, pressurizer heaters (J3, f/4, 853, f155 and //56 will not be able to perform their intended function.

SAFETY EVALUATION This Temporary Modification has been classified as safety related because it affects the Reactor Coolant System.

The Plant Nuclear Safety Review Committee (PNSRC) has reviewed this Temporary Modification per the review criteria of PMI-1040, Rev. 3 It was concluded by the review, that this Temporary Modification does not constitute an unreviewed safety questions as defined in 10 CFR 50.59, nor does it create a substantial hazard to the health and safety of the public.

Tem orar Mod. 835 (Unit 1)

BRIEF EVALUATION This Temporary Modification involves the lifting of Cable 3151-1 which is the power feed for pressurizer heater f!48. This pressurizer heater has a defective heater element. By disconnecting this cable, pressurizer heater f148 will not be able to perform its intended function.

SAFETY EVALUATION This Temporary Modification has been classified as safety related because it affects the Reactor Coolant System.

The Plant Nuclear Safety Review Commi.ttee (PNSRC) has reviewed this Temporary Modification per the review criteria of PMI-1040, Rev. 3.

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It was concluded by t'e review, that this Temporary Modification does not constitute an unreviewed safety questions as defined in 10 CFR 50.59 nor does it create a substantial hazard to the health and safety of the public.

Tem orar Mod. 836 (Unit 1)

BRIEF DESCRIPTION This Temporary Modification involves the lifting of Cable 3139-1 which is the power feed for pressurizer heater 856. This pressurizer heater has a defective heater element. By disconnecting this cable, pressurizer heater //56 will not be able to perform its intended function.

SAFETY EVALUATION This Temporary Modification has been classified as safety related because it affects the Reactor Coolant System.

The Plant Nuclear Safety Review Committee (PNSRC) has reviewed this Temporary Modification per the review criteria of PMI-1040, Rev. 3.

It was concluded by the review, that this Temporary Modification does not constitute an unreviewed safety question as defined in 10 CFR 50.59 nor does it create a substantial hazard to the health and safety of the public.

Tem orar Mod. f/43 (Unit 1 & 2)

BRIEF DESCRIPTION Installation of the Duratek Demineralization System using a mechanical jumper to route waste hold-up tank water to the 587'rumming room for processing. The effluent will be routed to the waste evaporator condensate tanks. All hose connections will have a working pressure at 300 PSI.

The system adds additional w'aste processing capability with a maximum feed flow of 55 GPM and will be used as an alternative to South Radwaste Evaporator operation.

SAFETY EVALUATION This Temporary Modification has been classified as Safety Interface because this system handles radioactive solids, liquids and gases. The system itself is entirely Seismic Class III.

The Nuclear Safety and Licensing Section has reviewed this proposed change as per the review criteria in NS&L Procedure No. 7. As a result of the Safety review, it was concluded that this Temporary Modification does not constitute an unreviewed safety question as defined in 10 CFR 50.59 nor does it create a substantial hazard to the health and safety of the public.

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Tem orar Mod. !/149 (Unit 1 & 2)

BRIEF DESCRIPTION This Temporary Modification involves the addition of a 15X Sodium Hypochlorite solution into the Circulating Water supply in lieu of the Chlorination System to control algae and slime and to regain cooling efficiency in the Circulating Water Condensers.

SAFETY EVALUATION The subject modification has been classified as Safety Interface since it involves adding chemicals that may interact with the Essential Service Water System, a Class I system.

It was concluded that the addition of Sodium Hypochlorite would not adversely affect the safety systems of the Plant. Further, it was pointed out that one of the reasons the system is described in the FSAR is that the NRC is the lead Federal Agency and addition of Sodium Hypochorite to the lake is an environmental matter normally handled by the Environmental Protection Agency.

This Modification does not constitute an unreviewed safety question as defined in 10 CFR 50.59 and will not adversely affect the health and safety of the public.

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CHANGES TO PROCEDURES A brief description of a procedure change implemented under the provisions of 10 CFR 50.59 and the associated safety evaluation is provided below:

Unit 2 Main Steam Safet Valve Set Point Verification Procedure The change in Unit 2 main steam safety valve set point verification consisted of a new procedure implemented through Special Procedure 12 MHP SP.126, Revision 1, which allows setpoint testing of main steam safety valves in Modes 1, Power Operation, through Mode 3, Hot Standby.

, One safety valve is tested at a time. During testing the valve is considered inoperable per T/S 3.7.1.1. Thus per the Action Statement, the Power Range Neutron Flux High Setpoint is reduced per Table 3.7-1.

If a problem should occur, the most plausible problem with the test would a safety valve sticking open. Since only one valve at a time is

,tested, any transient resulting from a stuck-open valve is bounded by the Unit 2 steam line break analysis found in Section 14.1.5 of the updated FSAR. This conclusion is valid in Modes 1, 2, and 3. Accident analysis assumptions which rely on the safety valve to open to relieve pressure, are maintained by causing a stuck-open safety valve during power operation, there is a possibility of a reactor trip. Although this potential exists, its possibility is minimized by provided a hydraulic closing device. Even if a trip occurs the consequences are bounded buy the existing accident analyses.

A change to Section 10.2.4 of the Updated FSAR has been made which states that "steam generator safety valve setpoints are checked periodically prior to or during scheduled outages".

The safety evaluation concluded that testing of the Unit 2 main steam safety valves in Modes 1, 2, or 3 does not constitute an unreviewed safety question as defined in 10 CFR 50.59.

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CHALLENGES TO PRESSURIZER PORV'S AND SAFETY VALVES There were no challenges to the Pressurizer PORV's or Safety Valves for either Unit 1 or-2 of the Donald C. Cook Nuclear Plant during 1988.

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Annual 0 eratin Re ort Irradiated Fuel Examinations During 1988, two separate'xaminations were performed on the irradiated fuel discharged from Unit 2 Cycle 6. These examinations were conducted in parallel with, or shortly after, the core was unloaded, and the intent was to determine fuel pin failures as well as gross structural defects in the assemblies.

The first examination was a routine binocular 'inspection of the fuel assemblies (**12 THP 6040 PER.353). As each assembly is downloaded to the Spent Fuel Pool,-

sides visually. The examiner is it is examined on all four looking specifically for tom or missing gridstraps, missing or damaged fuel pins, excessive clad hydriding, or rod bow to gap closure. This inspection is primarily intended to detect fuel damage caused by mechanical interaction between assemblies or baffle jetting, and is done during each refueling. There was no indication of any fuel damage.

Due to RCS chemistry levels indicative of several leaking fuel pins, a contract was let to Advanced Nuclear Fuels (ANF) to provide Ultrasonic (UT) examination of the assemblies making up the Unit 2 Cycle 6 core, as well as any replacement assemblies for Cycle 7. the Ultrasonic system works by a probe transceiver sending a high frequency sound wave into a fuel pin and measuring the strength of the returning signal, or "ring back". A fuel pin can be determined to have water in it by monitoring the relative strength of this ring back. In this way, not only can an assembly be determined to have leaking pins, but the numbers and locations of the bad pins can be identified.

Testing results:

Assemblies Tested: 196 AssemblieS. with Failures 6' Number of Failed Fuel Pins Assemblies Fuel Batch Vendor Tested Assemblies Pins M (1 time burned) W 3 0 0 R (3 times burned) W 1 0 0 S (3 times burned) ANF 12, 0 0 T (2 times burned) ANF 92 5 8 U (1 time burned) ANF 88 1 1 196 6 9 Three of the failed fuel pins (2 in assembly T24 and 1 in assembly U39) were on the periphery. Using the small camera mounted on the testing system, a short video inspection was performed. The rods in assembly T24 both showed signs of secondary hydriding. The rodlet in assembly U39 was inconclusive.

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Indiana Michigan Power Company Cook Nuclear Plant P.O. Box 458 Bridgman, Ml 49106 616 465 5901 INDIANA POPOVER NICHIGiAN February 28, 1989 United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555, Donald C. Cook Nuclear Plant Docket Nos. 50-315/50-316 License Nos. DPR-58/DPR-74 Document Control Manager:

Two copies of the 1988 Annual Operating Report for the Donald C.

Cook Nuclear Plant are being tiansmitted to you under this cover letter. The information contained in this report covers the activities delineated in the Donald C. Cook Nuclear Plant Technical Specifications, Section 6.9.1.5, and the requirements of 10 CFR 50.59.

Copies of this report have been transmitted to the Regional Administrator, the Director of Inspection and Enforcement, the Director, Office of Management Information and Program Control of the United States Nuclear Regulatory Commission and the NRC Resident Inspector as specified in 10 CFR 50.4 and 10 CFR 50.59.

Respectfully, W.G. Smith, J Plant Manager CC D.H. Williams, Jr.

M.P. Alexich P.A. Barrett S.J. Brewer J.F. Kurgan A.B. Davis, Regional Administrator, Region Director, Inspection and Enforcement III Director, Office of Management Information and Program Analysis NRC Resident Inspector, Bridgman G. Charnoff, Esq, R.C. Callen, MPSC D. Hahn g/~sz/2&SI'I INPO ~gy

~l [ g~ ZEs Qi~~e 4 I I Dottie Sherman, ANI Library /