IR 05000289/1987002

Insp Rept 50-289/87-02 on 870109-0206.No Violations Noted. Major Areas Inspected:Outage Activities in Operation,Maint & Surveillance,Including Refueling Operations,Annual Emergency Diesel Overhaul & Fire Protection Upgrading

ML20205G463
Person / Time
Site:	Three Mile Island
Issue date:	03/24/1987
From:	Blough A, Conte R, Young F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20205G435	List: IR 05000289/1987002 ML20205G430
References
50-289-87-02, 50-289-87-2, NUDOCS 8703310462
Download: ML20205G463 (26)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report N /87-02 Docket N License N DPR-50 Priority -- Category C Licensee: GPU Nuclear Corporation Post Office Box 480 Middletown, Pennsylvania 17057 Facility At: Three Mile Island Nuclear Station, Unit 1 Inspection At: Middletown, Pennsylvania

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Inspection Conducted: January 9, 1987, to February 6, 1987 i Inspectors: P. Bissett, Reactor Engineer, RI D. Johnson, Resident Inspector (TMI-1)

F. Paulitz, Reactor Engineer, RI J. Rogers, Resident Inspector (TMI-1)

Reporting Inspector:

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3 F. Youngf Resident Inspector (TMI-1)

2- EV-J9 Date Reviewed by M[

h R. Conte /S'enior Resident Inspector (TMI-1)

3 2V-17 Date Approved By: 'M A. Blough'(Chief 2 -~ D - P )

Date Reactor Projects Section No. IA Division of Reactor Projects l Inspection Summary:

Areas Inspected Resident and region-based NRC staff conducted safety inspections (243 hours0.00281 days <br />0.0675 hours <br />4.017857e-4 weeks <br />9.24615e-5 months <br />)

of outage activities, focusing on plant and personnel performance. Specif-

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ically, items reviewed in detail in the operation, maintenance, and surveil-

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lance areas were: refueling operations and annual emergency diesel generator overhaul. Other items included: control rocm habitability modification, up-

, grade of emergency feedwater to safety grade, inservice testing (IST) program

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review, fire protection upgrading, inservice inspection of secondary piant l piping wall thickness, and licensee action on previous inspection findings.

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Inspection Summary (Continued) 2 Inspection Results:

For the activities sampled, the inspectors noted that, in general, the licensee properly conducted evolutions and completed maintenance / surveillance activit'ies consistent with regulatory requirements. Although problems were experienced with the "1A" diesel engine overhaul, the licensee properly evaluated the cause and successfully completed engine testing. Modifications for control room

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habitability and the emergency feedwater system were being accomplished in accordance with regulatory requirement The IST program was found to be properl,y : implemented. The licensee's program for secondary system piping wall thinning was expanded from its initial scope in response to a recent event at another reacto No violations of NRC requirements were observed. Unresolved items were identi-fled regarding (1) the licensee's computer program for electrical load distri-bution analysis, and (2) licensee resolution of labor disputes. The licensee continues to keep the NRC Resident Office informed on internal matters, not required to be reported, but of potential importance to safet .

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DETAILS Introduction and Overview 1.1 NRC Staff Activities The overall purpose of this inspection was to assess licensee activ-ities during the cold shutdown mode as they related to reactor safety and radiation protection. Within each area, the inspectors documen-ted the specific purpose of the area under review, scope of inspec-tion and findings, along with appropriate conclusions. The inspec-tors made this assessment by reviewing information on a sampling

, basis through actual observation of licensee activities, interviews with licensee personnel, measurement of radiation levels, or' inde-pendent calculation and selective review of listed applicable document During this period, region-based inspections were conducted in the areas of control room habitability modification, emergency feedwater upgrades, and inservice testing (IST) progra .2 Licensee Activities During this period, the licensee maintained the plant in cold shut-dow The fuel shuffle evolution was completed and the refueling transfer canal was pumped back to the Borated Water Storage Tank (BWST). Preparations for reactor vessel (RV) head installation were being made. The licensee completed the second overhaul of the "1A" EDG and successfully completed a test "ren-in" evolution. Continued efforts were being made in upgrading the fire protection and emerg-ency feedwater system . Plant Operations 2.1 Criteria / Scope of Review The NRC resident inspectors periodically inspected the facility to determine the licensee's compliance with the general operating re-quirements of Section 6 of the Technical Specifications (TS) in the following areas:

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review of selected plant parametars for abnormal trends;

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plant status from a maintenance / modification viewpoint;

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control of ongoing and special evolutions, including Control Room personnel awareness of these evolutions;

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control of documents, including logkeeping practices;

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implementation of radiological controls;

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implementation of the security plan, including access control, boundary integrity, and badging practices; and,

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implementation of the fire protection plan, including fire bar-rier integrity, extinguisher checks, and housekeepin Because of additional resident office coverage at this facility, nore detailed and frequent reviews of operating personnel performance were conducted to determine that:

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operators are attentive and responsive to plant parameters and conditions;

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plant evolutions are used and followed as required by plant policy;

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equipment and status changes are appropriately documented and communicated to appropriate shift personnel;

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the operating conditions of the plant equipment are effectively monitored and appropriate corrective action is initiated when required;

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backup instrumentation, measurement, and readings are used as appropriate when normal instrumentation is found to be defective or out of tolerance;

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logkeeping is timely, accurate, and adequately reflects plant activities and status;

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operators follow good operating practices in conducting plant operations; and,

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operator actions are consistent with performance-oriented trainin Specifically, the inspectors focused attention on the areas listed belo General / Operations

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Control Room operations during regular and backshift hours, including frequent observation of activities in progress, and periodic review of selected sections of the shift foreman's log and control room operator's log and other control room daily logs

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Areas outside the Control Room, including important-to-safety buildings

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Refueling Ope. rations

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Selected licensee planning meetings Maintenance

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Fuel handling equipment repair

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DHV-5A and DHV-14A valve repair (disc replacement)

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EG-Y-1A - emergency diesel generator yearly maintenance As a result of this review, the inspectors reviewed specific areas in more detail as described in the sections that follo .2 Findings / Conclusions 2. Refueling Operations The inspector conducted a review of the licensee's comple-tion of refueling activities. After fuel shuffle and prior to installation of the plenum and reactor vessel head, the licensee conducted an inspection of the fuel and verifica-tion of fuel load patter This verification is accom-plished to assure that no damage resulted to the fuel dur-ing installation and that all fuel bundles were installed in the proper location The licensee uses a remotely controlled camera lowered into the refueling canal and the reactor vessel to conduct the inspection. The inspector witnessed various portions of this inspection to verify proper fuel installatio The licensee did not detect any deficient conditions during the inspectio The inspector reviewed Administrative Procedure (AP) 1023, Revision 9, dated April 16, 1985, " Cleanliness Requirement,"

which specifies the cleanliness levels to be maintained in the fuel area to verify that proper levels of cleanliness

were maintained during the fuel inspection.

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The inspector also questioned licensee engineering person-nel concerning recent events at some pressurized water reactor (PWR) plants where fuel degradation had occurred due to water jet impingement at fuel pins in outer areas of

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the core where the core baffle plates had separated. This allowed high velocity water to degrade fuel cladding, re-sulting in the exposure of fuel pellet Licensee repre-sentatives responded that the design of the Babcock &

Wilcox (B&W) reactor vessel baffle plates precluded this problem. Within the B&W reactor vessel baffle, a small amount of water flows through the baffle element that eliminates the pressure differential needed for this prob-lem to occur. No evidence of baffle plate separation or fuel damage had occurre The inspector concluded that the fuel inspection had been adequately completed and that fuel shuffle and core reload operations had been conducted in a safe manner. The licen-see's post-fuel load inspection was also conducted ade-quatel The inspector had no concerns in this are . Diesel Generator Maintenance Review The licensee commenced the annual inspection of the "1A" emergency diesel generator (EDG) in December 1986. During this inspection, the licensee planned to replace all twelve engine cylinder liners because of a water jacket leak that was discovered on one cylinder during the November 1985 inspection. During the last one year interval, the licen-see has monitored the engine lube oil for evidence of excessive water / anti-freeze solution in the oi No evi-dence of excessive leakage has been observe The inspector reviewed the documentation associated with this inspection, which included the following:

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Job Tickets (JT) CI-261 and CL-595;

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Special Temporary Procedure (STP) 1-87-0006, Revis-ion 1, dated January 24, 1987, "EG-Y-1A Run-In;"

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Surveillance Procedure (SP) 1301-8.2, Revision 29, dated December 3, 1986, " Diesel Generator "A" Annual Inspection;" and,

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Corrective Maintenance Procedure (CMP) 1405-3.2,

" Diesel Engine Maintenance."

The inspector also observed various phases of the repair /

inspection activities and portions of the subsequent testin .

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The diesel generator inspection and maintenance is accom-plished by on-site maintenance personnel with guidance from a Fairbanks-Morse maintenance specialist. Procedures used are licensee procedures which are supplemented by guidance from the vendor. Certain recommendations made by the ven-dor are incorporated into the repair sequence after evalua-tion / concurrence by on-site engineering personnel. As a result of this process, certain modifications and adjust-ments were made to the diesel engine to improve performance and certain operating parameters. Specifically, the engine timing was advanced from 38 degrees to 42 degrees. The intent of this was to reduce combustion chamber pressure Also, an additional oil wiper piston ring was added to the bottom of the lower pistons to reduce oil consumption. The comb.ination of these modifications with reduced piston-to-cylinder wall clearances from the installation of the new cylinder liners resulted in higher than normal exhaust temperatures and also higher cylinder wall temperature This resulted in significant damage and excessive wear to some pistons and cylinder walls during testing after the ,

initial engine overhau This testing was accomplished

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over the January 10-11, 1987, weeken A small injector fuel leak and a minor jacket coolant leak required engine shutdown for repair A subsequent inspection of the cylinder walls through an inspection port revealed the damag The licensee subsequently disassembled the engine and re-placed all cylinder liners and some pistons and connecting rod bushings that had been damaged during the initial test-in In addition, other modifications and setpoint adjust-ments were mad The additional oil wiper ring that was

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installed on the lower piston was removed. The new cylin-ders were dimension checked to verify proper clearances and

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uniform manufacture. A special engine "run-in" sequence

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was developed and, during this "run-in," engine timing was reset to 40 degrees. The vendor also recommended reducing engine coolant temperature gradient (delta T) to 10 degree Previously, a 16-20 degree delta-T was used. This involved I

changing flow control orifices in the coolant line Following the second overhaul, the licensee successfully tested the engine and it was declared operable. At the end of this inspection, the licensee was in the process of con-ducting the inspection on the "B" EDG. When this inspec-tion / overhaul is complete, the licaasee will issue a report detailing the analysis of the caua of the engine failur At present, the licensee and vendor are evaluating the data accumulated during the second run-in procedure from the "A" EDG.

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Final recommendations on engine maintenance procedures and settings will be determined from that evaluation and the results of the run-in in the "B" ED The NRC Resident Office will continue to routinely follow this are . Emergency power Requirements While Shutdown During the course of the EDG engine maintenance, the in-spectors observed that the licensee was maintaining the one EDG associated with the operating decay heat (DH) loop in

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an operable status. Durirg one 5-hour period, both diesels were inoperable due to a surveillance that was required to be accomplished on the operable EDG. Also, when the "A" EDG was being overhauled the second time, the licensee used the "B" decay heat loop as the primary means of decay heat removal, due to the fact that the "A" decay heat loop was required to be disabled for extensive maintenanc This prompted the inspectors to review the emergency power

requirements for the decay heat loops while the plant is shutdown. The licensee's technical specifications (as well as the technical specifications of numerous other plants)

do not specifically require that the emergency diesels be operable when the reactor is not critical which would allow them to function as back-up emergency power when off-site power is lost for any reason. This is primarily due to the fact that the licensee's technical specifications defini-tion of "0PERABLE" does not include the standard technical specification clause that all " normal and emergency" power be available to consider a certain safety-related compon-ent, which uses the EDG's as emergency power, operabl The inspectors were concerned that the licensee could re-move both EDG's from service and the only power available would be off site via the auxiliary transformers. This has been a reliable source in the past and, as a matter of scheduling, the licensee does attempt to maintain an oper-able EDG to power the operable DH loop that is being used while shut down. But this is not always the case, as de-scribed above. The inspectors confirmed that this type of situation was being evaluated as part of the generic issue of station blackout by the Office of Nuclear Reactor Regulation (NRR) for operating and shutdown modes.

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4 The inspectors discussed this concern with licensee manage-ment, who acknowledged that, as an administrative control when shutdown, an attempt is made to schedule maintenance on safety-related components that can use the EDG as an emergency power sourc The intent is to schedule mainte-nance such that an EDG is always available to power the operating DH loop if off-site power is lost. This is not a requirement and cannot always be practiced due to mainten-ance on certain safety-related component . Investigation of Drug Misuse Offsite In December 1986, the licensee 1,7 formed NRC Region I that they were investigating a concern of drug misuse by em-ployees off site and that they would inform Region I on the results of that investigatio The licensee's Fitness for Duty policy requires drug test-ing to occur when reasonable grounds exist to suspect pos-sible drug us Should the tests prove positive, the licensee has the option of imposing either a disciplinary suspension with counseling and a rehabilitation program or dismissal of the employee. The licensee's investigation involved a total of fourteen employee One employee resigned after refusing to be drug tested, three were tested positive, and ten were cleared after extensive drug testin Recently, the licensee informed NRC Region I that the three employees that tested positive for drugs received disci-plinary actio All three individuals were employed off site and were generally not involved with day-to-day plant

, operations. Site access for each individual had been re-voke Each individual is required to complete a drug rehabilitation program and each must agree to random drug testing for an indefinite perio Since the individuals involved were employed off site, the safety significance of this occurrence is minima The licensee plans to issue a final report documenting the above information and actions taken to resolve the issu NRC Region I concludes that the licensee's actions thus far have been appropriate and reflect a conscientious manage-

, ment approach to fitness for duty issue NRC Region I will follow subsequent licensee actions and will review the licensee's final report (289/87-02-01).

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2. Worker Dispute During the weeks of August 18 and 25,1986, licensee repre-sentatives briefed the NRC Resident Office regarding (1)

planned disciplinary action against an employee for inade-quate performance and (2) counter-charges by the employee implying harassmen This case was undergoing internal review within GPU Nuclear Corporation. By letter, dated September 9, 1986, NRC Region I requested a written summary on this matter once the licensee dispositioned this case and any associated issue During this inspection, the inspector requested a status of the subject repor Licensee representatives indicated that the matter was still before the U. S. Department of Labor (DOL) and that the issues are expected to be resolved later this year. The requested report will be issued sub-sequent to the DOL resolutio The final resolution and report on this dispute is unre-solved pending Region I review of the licensee's submitted report (289/87-02-02).

2.2.6 Steam, Feed, and Condensate System Wall Thinning In accordance with NRC Region I Temporary Instruction (TI)

87-02, " Steam, Feed and Condensate System Surveys," the resident inspector reviewed the licensee's program for the determination of wall thinning of large diameter steam, feed, and condensate piping. The TI was issued as a fol-lowup to IE Information Notice 86-106, "Feedwater Line Break," which alerted licensees to a potential generic problem with feedwater pipe wall thinning related to the feedwater line break at Surry, Unit 2, on December 9,198 The purpose of the TI is to determine what actions the l licensee may have taken as a response to the Surry feed-water line break and documentation of the following items:

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the extent of the licensee's current or planned secon-dary system pipe wall thinning identification program;

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the guidelines used to identify secondary piping con-figurations that could be susceptible to accelerated erosion / corrosion; and,

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results of measurements take ,

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The inspector reviewed the following licensee documents pertaining to the wall thinning inspection program:

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SP 1101-12-102, Revision 0, Technical Specification for Pipe Wall Thinning Inspection;"

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SP 1101-12-099, Revision 1, " Wall Thinning Inspections of Main Steam Line Piping;"

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Field Change Request (FCR) No. C 038229, dated

, January 5, 1987, " Erosion Inspections;"

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FCR No. C 38230, dated January 14, 1987, "

Erosion Inspections;" and,

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GPU Memorandum No. MSS-87-012, dated January 12, 1987, concerning identification of piping areas potentially

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susceptible to erosio Engineering Technical Specification SP 1101-12-102 and the GPU memorandum both state the organization, records keeping requirements, and specifications required to establish and implement an erosion / corrosion wall thinning inspection progra The following guidelines are used to identify piping configurations in the main steam, condensate, and

feedwater systems that could be conducive to accelerated erosion.

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High moisture content in steam

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High local velocity areas such as headers

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Piping undergoing multiple direction changes within 3-5 pipe diameters (except for opposing fitting within the same plane) and single directional changes greater l

than 90 degrees

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Piping containing operating fluid temperatures in excess of 210 F

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Flow measuring devices located within ten diameters upstream cf a directional change

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Any piping made of carbon steel

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Continuous usage process lines, but bypass lines and pump recirculation lines were not included

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The initial inspection program (SP 1101-12-201 and SP 1101-12-099) used engineering analysis to prioritize main steam lines into categories of most and least acceptable to wall thinning. In addition, several FCR's were issued to expand the initial inspection program in response to the subject feedwater line break accident. From the initial prioriti-zation, thirty-two areas were identified as inspection areas for possible erosion / corrosio The FCR's added approximately sixteen more areas of inspection from the main feedwater, heater drains, and condensate system Most of these areas of inspection have been completed dur-ing the current 6R outage. A wall thickness less than 8 percent requires an engineering evaluation to determine dispositio Less than 87.5 percent wall thickness indi-cations have been found on the discharge piping of the "A" and "C" heater drain pumps. The eroded section of the "A" and "C" heater drain pump discharge piping is being re-place One other area of less than 87.5 percent wall thickness indication has been found on several pipe elbows of the sixth stage extraction steam lin These pipe el-bows are being repaire The licensee has a secondary wall thinning identification program, which uses specific guidelines, that has identi-fied approximately forty-eight inspection areas for pos-sible erosion / corrosion of which sixteen areas were added after an event at another reactor sit Most of these inspections have been completed during the current outage and indications are being resolved. Based on the above, the licensee appears to have a program in place to detect secondary wall thinning and the action items required in NRC Region I TI 87-02 have been completed.

2.3 Summary i In general, licensee management continued their strong involvement in

site activitie Refueling operations activities were well control-led. Housekeeping and cleanliness of the plant were adequate with respect to the significant amount of work being performe The annual emergency diesel maintenance resulted in excessive wear to some pistons; however, the licensee quickly identified the discrep-l ancy and is resolving the problem with the vendor. The Operations Department continues to maintain a positive control over plant activ-i ities and remains knowledgeable of current plant condition .

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A substantial program was already established to examine for secon-dary plant pipe wall thinning and the licensee showed initiative to almost immediately expand the program to the feed and condensate system in light of a recent reactor even The licensee continues to keep the NRC Resident Office informed on

, internal matters, not necessarily required to be reported, but have the potential to affect safet . Inservice Testing Program 3.1 Criteria / Scope of Review The objective of the licensee's inservice testing (IST) program is to provide added assurance of the operational readiness of certain ASME (American Society of Mechanical Engineers) Code Class 1, 2, and 3 pumps and valves that are required to perform specific functions in shutting down the reactor or in mitigating the consequences of an accident. As required by 10 CFR 59.55a(g) and as documented, the

licensee's IST program is being conducted in accordance with the ASME B&PV Code, 1980 Edition with Addenda through Winter 1980. Specific relief from certain ASME Code requirements is documented within the licensee's IST program and is currently under review by the NRC Office of Nuclear Reactor Regulation (NRR).

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An on-site inspection was conducted to ascertain whether the licen-see's program, procedures, and work activities are consistent with regulatory requirements and licensee commitment The inspection

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details are discussed in the following paragraph This review of completed inservice testing was performed to verify the accomplishment of the following attributes during the performance of these tests:

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tests were performed using properly approved up-to-date

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procedures;

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tests were performed within the required time limits; i --

test results met acceptance criteria and, if not, appropriate action was taken; and,

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completed tests were reviewcd and evaluated for acceptability within 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> following completion of the test.

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3.2 Review / Implementation Various discussions were held with operations and engineering per-sonnel, including operators, engineers, and managers, to ascertain the extent of their responsibilities. Additionally, reviews were made of those administrative controls presently in place that are used to control the conduct of the pump and valve testing progra This included the scheduling, performance, documentation, and subse-quent review of completed test result Conduct of the licensee's IST progam is controlled through Admin-istrative Procedure (AP) 1041, "IST Program Requirements," and with AP 1001J, " Technical Specification Surveillance Testing Program."

Scheduling of pump and valve testing is primarily accomplished through the issuance of surveillance test program schedules. Actual testing is then performed by operations personnel utilizing specific surveillance test procedures written by engineerin During this inspection, various surveillance procedures, documenting the performance of required pump and valve testing were reviewed for completeness, technical adequacy, and compliance with applicable por-tions of the ASME B&PV Code. Also, those operating procedures in place that provide additional controls for the scheduling and even-tual completion of pump and valve testing during certain plant condi-tions were reviewed. Examples of the above surveillances reviewed included the following:

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Surveillance Procedure (SP) 1300-3C, Revision 13, dated October 10, 1986, " Decay Heat Closed Cooling Water Pumps Func-tional Test;"

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SP 1300-3K A/V, Revision 20, dated January 17, 1987, Reactor Building Emergency Cooling Pump Functional Test and Reactor Building Cooling System Valve Operability Test;"

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SP 1300-3P, Revision 11, dated August 26,1986, "IST of Check Valves During Plant Shutdown;"

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SP 1300-3R, Revision 15, dated September 3, 1986, "IST of Valves During Shutdown and Remote Indication Check;"

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SP 1300-3T, Revision 9, dated June 30, 1986, " Pressure Isolation Test of CF-V4A/B, CF-VSA/B, and DH-V22A/B;"

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SP 1300-30, Revision 0, dated August 16, 1986, "IST of Boric Acid Injection Valves - Boric Acid Recycle System;"

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SP 1303-11.21, Revision 7, dated December 3,1983, " Core Flood-ing Valve Operability List;"

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Operating Procedure (0P) 1102-2, Revision 76, dated January 16, 1987, " Plant Startup;" and,

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OP 1102-11, Revision 66, dated January 30, 1987, " Plant Cool-down."

The inspector discussed in detail with various plant personnel those

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controls presently in place that provided additional assurance that testing deferred due to plant conditions would be accomplished within the allowable time frame. Specific situations ' were chosen where instances of this nature had occurred. Licensee representatives sub-sequently demonstrated to the inspector that these testing require-ments were effectively tracked by several means to ensure eventual completio Mechanisms in place, in addition to the surveillance schedules, include the plan-of- the-day agenda, regulatory retest tags, the regulatory retest tag log, operations' review of outstand-ing Exception and Deficiency (E&D) sheets and department head in-formal revie .3 Observation of IST Field Activities During this inspection period, the inspector witnessed the perform-ance of the IST quarterly surveillance test 1300-3C, " Decay Heat

, Closed Cooling Water Pumps Functional Test." Performance of this test verified pump operational readiness. NRC observation of this surveillance was to verify that the following was accomplishe Applicable procedures were approved, up to date, and used throughout the conduct of the tes Appropriate personnel were notified prior to the start of the tes Installed instrumentation from which data was taken was i calibrate Calibrated measuring and test equipment was use Acceptance criteria were met and, if not, appropriate steps were take '

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During the review of completed test results, it was noted by the shift foreman in charge of the test that the pump radial and thrust bearing reference temperatures in comparison to their actual tempera-tures appeared to be reversed (i.e. , thrust bearing reference tem-perature should have been the radial bearing reference temperature and vice versa). Although both temperatures vere less than the maxi-mum allowed (180 F), the shift foreman felt that engineering per-sonnel should be contacted even though all test results met accept-ance criteri Engineering personnel normally review only those tests in which measurements fall within the " alert" or " required" action ranges (acceptance criteria is not met). This practice pre-sented some concern to the inspector, as detailed below.

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3.4 Other Findings / Conclusions It was determined that the licensee has instituted and implemented an inservice testing program that meets the requirements of Section XI of the ASME B&PV Code. However, some areas of concern were presented to licensee management which pertained primarily to licensee engi-neering involvement, test review, and trendin As noted earlier in this section, engineering personnel develop the particular surveillance tests for inservice testing and operations personnel subsequently perform these test Engineering involvement following completion of testing does not normally occur if test re-sults are acceptable. Since the engineering group does not review acceptable test results, they do not effectively trend pump and valve performance parameter Trending essentially occurs only after a problem or potential problem has occurred or has been identifie The inspector determined that this practice was nothing more than confirming that an adverse trend had indeed developed. The inspector questioned the usefulness of identifying a trend if a component had already faile The following examples were presented to the licensee to point out why the licensee engineering or management should become more in-volved in the review process of all completed inservice tests. These problems were identified by the inspector following a review of com-pleted inservice tests previously noted in paragraph Performance of test 1300-3KA/B on June 5,1986, was completed, reviewed, and noted as acceptable. The same test was.again per-formed on September 3, 1986. However, in comparing valve stroke times with the previous test results, it was noted that valve RR-V10A had exceeded its maximum stroke time during the June 5, 1986, test. Approximately three months had expired before this was identifie .

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Had technical support personnel been reviewing the results of all IST tests and had an in place trending program existed, this oversight by operations would not have gone unnoticed for three month For the same test, 1300-3KA/B, performed on March 10, 1986, "B" pump bearing temperatures had not stabilized prior to running the test as required by paragraph 6.1.29 and IWP 3500(b). Test results were considered acceptabl Even though the final test results were acceptable, the licensee should have required the test to have been rerun. As a minimum, test results should have been compared to those results of the previous test in which bearing temperatures had stabilized prior to conducting the tes SP 1300-30, performed on January 29, 1986, had an apparent reference temperature error for the thrust and radial bearing (refer to paragraph 3.3 of this report, " Observation of IST Field Activities).

Test results were acceptable and, therefore, no further engi-neering involvement was required. Had the shift foreman not been as conscientious and as observant as he was, engineering personnel may not have been contacted in reference to the ques-tionable bearing temperature ,

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SP 1300-3T, pressure isolation test performed on January 28, 1986, had an initial leakage test failure of a core flood valv '

Subsequent test results were within acceptable limits. A dis-crepancy (E&D) sheet was generated documenting the initial fail-

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ure; however, the decay heat system valve designator was used (3 times) vice core flood designator on the E&D write-up.

l This was an administrative error; but, nesertheless, another example of where additional licensee technical support involve-ment might alleviate problems of this natur The licensee acknowledged the inspector's comments and stated that this area would be reviewe There is a heavy involvement by operations department supervisory and engineering personnel in the surveillance area. However, their involvement is not completely effective in identifying the types of errors noted abov Plant engineering personnel could provide an additional measure of independence in the review process of accept-able test results. More significant errors of this nature could lead to violations of IST program / procedure requirement _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

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3.5 Summary ,

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The established IST program is essentially in copformance with regu-latory requirements. However, several licensee exemption requests need to be reviewed by NRC staff to assure full dompliance with those requirements. The program appears to be properly implemented. How-ever, for acceptable test results, the lack of a trending program represents a weakness in technical support for this program. As a result, errors associated with acceptable results are allowed to persist for excessive periods of tim . Control Room Habitability Modifications 4.1 Background By 10 CFR 50 Appendix A, General Design Criteria 19," Control Room",

the licensee is required to assure that control room operators will be adequately protected against the effects of accidental release of toxic and radioactive gases and the nuclear plant can be safely oper-ated or shutdown under design basis accident condition Listed below is the licensee's design to protect the control room operators from radioactive gas and other toxic gases from on-site and off-site source Radioactive gas will be precluded from the control room by maintain-ing a positive pressure of 0.1 inch WG within the control room using the emergency fans and filtration unit The total outside air entering the control room will be limited to 12,000 cubic feet per minute (cfm).

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Chlorine gas will be prevented from entering the control room by detecting 5 PPM chlorine or greater at either the air intake or near the chlorine receiving and storage location. Upon detec-tion of chlorine, the control building ventilation system (CBVS)

will be isolated in ten seconds. An alarm is received in the control room for 1 PPM chlorine detectio Ammonium hydroxide gas will not be a hazard to the control room operator for a postulated leak of the storage tank if the con-tent of the tank is confined to a seismic designed dike and the total surface areas of this dike does not exceed 550 square fee Analysis to assure no adverse impact from toxic gases from off-site sources; e.g., nearby railroad accident and chemical spil _ _ - . _ .

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1S The detection of radioactive gas and automatic actuation associated with this detection are not within the scope of this inspectio The inspector rev.iewed the following documentation as part of this inspection:

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TMI-1 Final Safety Analysis Report (FSAR), Update 4, 7/85, Sec-l tion 9.8, " Ventilation Systems;"

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Technical Specifications (TS) Amendment No. 76, TS Section 3.15,

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" Air Treatment Systems;"

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NUREG 0737, Item III.D.3.4, " Control Room Habitability;" and,

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NRR Safety Evaluation Reports on Control Room Habitability in

letters dated August 22, 1984, December 23, 1985, and August 14, 1986, 4.2 Design Requirements The licensee conducted a test to evaluate the capability of the con-trol building HVAC systems to maintain a post-accident habitable

environment in the control roo These test results are documented in a Technical Data Report (TDR) No. 728, Revision 0. As the result

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of this test, the licensee has committed to t.he followin Permanently isolate six control building HVAC system air regis-ters, which were blanked off during the test

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Maintain weather stripping on control room doors

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Revise Operating Procedure (OP) 1104-19, " Control Building Ventilation System" and Abnormal Procedure EP 1203-34, " Control Building Ventilation System":

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to assure a positive pressure of 0.1 WG inch in the control

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room during a single failure of damper AH-D-28; damper

AH-D-39 must be opened to a predetermined position; and,

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to assure a positive pressure of 0.1 inch WG in the control room during a single failure of damper AH-D-41A(B) located in the normal fan system, the opposite emergency fan must be starte The licensee also replaced the ducts between the filtration units and i their respective emergency fans. This replacement was done because

there was signs of deterioration of the duct and damper The new ducts are designed to withstand all operating conditions (standby positive 2-inch WG differential pressure (d/p); emergency negative 6.25-inch WG d/p due to clogged filter condition),

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4.3 Findings / Conclusions The inspector reviewed the documentation associated with the above scope and found them acceptable with the following exception System Design Description (SDD) T1670F, Division II, Revision 0, " Chlorine Detection System (CDS) for the Control Building Ventilation System (CBVS). Section 1.5.1 states, in part, that the chlorine probes CE-776-2 and CE-777-2 are located below grade in the air intake structur The inspector has visually verified that these probes are located in the air intake struc-ture above grade at the 320-foot elevatio An SDD revision will be processed by the licensee to account for this chang The licensee's program allows design changes to be made through FCR actions, updating of the SDD after completion of madifica-

, tions design.

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SDD TI 670F Section 1.6.6.2.2 states, in part, the "A" channel control circuit consists of a reset pushbutton located on sec-tion A of the H&V pane The inspector visually verified that the push button was located on the B section of the H&V pane An SDD revision will also be made by the licensee for this

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chang Procedures OP 1104-19, Revision 18, and EP 1203-34, Revision 9, have not been revised. The licensee has plans to review the These procedure revisions may be reviewed in a future inspec-l tion.

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Field Questionnaire No. C 038553, April 10, 1986, (Chlorine s Detection System) indicated that the solenoid operated valve 1 (SOV EP3 Cabinet A) was specified and purchased as a nuclear l grade device with a pressure range of 15-150 psig. However, the i intended response time could not be me A commercial grade ASCO S0V, 8320 B17666 stock symbol number 987-062-5001-1 was to be used. This SOV had an operating range of 0-50 PS The

, inspector visually verified that this commercial grade SOV had l been mounted in Cabinet A. It appeared to the inspector that the commercial grade SOV was unacceptable for this applicatio After further discussion with the licensee, it was determined that the modifications in question were not to be seismically r qualified as accepted per NRR safety evaluation reports for this

! item; therefore, the use of the commercial grade SOV was accept-

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able. The SDD T1-670F, Division II, was revised (Revision 2) to reclassify (seismically) the portion of the chlorine detection system.

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.The inspector visually verified that the following registers had been

"reinoved and blanked of R-1A and S-1A associated with AH-E-17A

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R-2A and S-1B associated with AH-E-17A

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R-9 inside of the control room near door 48 The register R-3, located in the control room hallway near door 47, could not be visually verified as to being removed and blanked off because the suspended ceiling nad been placed in front of i How-ever, the inspector did not feel or hear any air movement from the space next to the duct and would, therefore, conclude that the register had been blanked of The inspector examined the door seals of doors 47 and 48, which are part of the control room pressure envelop Door 48 is a double door, which would only be used in an emergency. There was no indi-cation that air was leaking out of this door. However, the inspector could hear air passing around the edge of door 47 and its frame. A visual inspection revealed that the door and the door jam does not appear to be plum. The seal is only applied to part of the door ja Later in the inspection period, the resident inspector noted that the door had been repaired. The licensee has a procedure for the main-

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tenance of these doors, and it is conducted on a six-month interva The chlorine detection system was inspected with the following finding The detection probes were not mounted and connecte The tubing to SOV EP3 Cabinet A was not connected to the solenoi (NOTE: The above-noted conditions were expected, based on status at the time of the inspection.)

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The conduit supports, ICCH-Cll,12,13, and 14 were verified to be located as specified on Drawing 5130 205-746, Revision IA/ The Channel A and B alarms were located on the H&V panel annun-ciators as specified on Drawing C-224-668, Revision IB/ The wires of the "A" reset pushbutton, located on the "B" H&V

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panel, were covered with a flame retardant as specified.

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The two-man cable pulling rule was changed to one-man for pull-ing the detection cable This was done to comply with the vendors pulling tension of 40 pounds and sidewall pressure of 214 pounds. It has been physically demonstrated that one elec-trician can pull 75 pound; therefore, licensee stated that field QC was further instructed to not allow excessive pull tension by the installe The ammonium hydroxide spillage dike was reviewed and physical meas-urements were made of the dike with the following finding The volume measured was 1209 cubic feet. This exceeded that re-quired to contain 7000 gallons which would be 900 cubic feet based upon a specific gravity of The measured horizontal area was 468 square feet which met the requirement that the_ area not exceed 550 square fee The replacement of the duct system between the filters and the emergency fans was inspected with respect to the dampers, damper actuator, and control tubin This replacement was found acceptabl .4 Summary The NRC staff's Safety Evaluation Reports (SER's) concluded the con-trol building HVAC system would be able to maintain a habitable post-accident environment in the control room. i r.e inspector concluded that the modifications and upgrades reflected the intent and the ob-jectives of the licensee's documented design. The noted problems and incomplete work were being reviewed by the licensee. This area will be further reviewed in a subsequent inspection (TMI TAP III.D.3.4.)

5. Auxiliary Feedwater System Modification 5.1 Background As the result of post-accident design review by the NRC, it was established that the auxiliary feedwater (AFW) system should be treated as a safety-related system. Modifications to the AFW system, now referred to as emergency feedwater (EFW) system, had been made prior to restar These modifications were implemented to upgrade the EFW system to a safety grade system in order to provide increased reliability in its capability to mitigate the effects of design basis accidents when the main feedwater :ystem 13 not available. The foi-lowing are additional modifications to or studies of the EFW system within the scope of this inspectio .

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~ ,, Pipe support in the reactor building;

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Seismic system interaction

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Emergency bus loading

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Turbine-driven EFW pump overspeed trip alarm

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Condensate storage tank level The inspector reviewed the following documentation during this inspection:

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FSAR Sections 7.1, 7.2, 7.3, 7.4, 8.2, 10.3, 10.4, 10.5 and 10.6;

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Technical Specification Sections 3.5, 3.7. 4.8, 4.9 and 4.15;

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NUREG 0737 Task Action Plan (TAP) II.E.1.2, " Auxiliary feedwater System Automatic Initiation and Flow Indication;"

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TMI-1 Restart Safety Evaluation Report, NUREG-0680 and 0680 Supplement No. 3; and,

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Draft Safety Evaluation Relating to NUREG 0737, Item II.E.1.2,

" Emergency Feedwater System Review."

5.2 Design The design for the condensate storage tank level is contained in the installation specification T1-IS-4120224-006, Revision 1, " Emergency

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Feedwater Heat Sink Protection System Electrical Modification." The

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design for the turbine-driven emergency feedwater overspeed trip alarm is contained in the installation specification T1-IS-4120224-004, Revision 0. These specific designs are as follow Provide redundant safety grade level indication and low-low level alarm in the control room for each condensate storage tan Provide a overspeed trip position switch on the turbine-driven EFW pump to actuate an annunciator alarm in the control roo As the result of a previous licensee system walkdown, it was observed that the emergency feedwater pipe support EF-111 for "B" loop line inside the containment was connected to the building frame through the steel stair framin Analysis of the support considering its interaction with the steel stair framing had shown over-stress of the stair members under seismic loading. This modification removed sup-port EF-111 and strengthened support EF-016 in order to maintain

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seismic integrity of the EFW system. Details are given in the in-stalletion specification T1-IS-115302-023, Revision 0, " Emergency Feedwater Support Modification." The results of the EFW system seismic interaction walkdown are contained in GPU letter 4211-84-2160, dated July 16, 1984, to J. F. Stolz, NRC. The above pipe sup-port interaction is listed in Item 1 In addition, the following items have been identified as requiring modificatio Item 4 - Cubical containing instrument air compressor IA-P-IA.

, Radiation Monitor RM-A-2 will be anchored to the floor to pre-clude sliding impact with power and control cables for EF-V-28 and power cable for EF-P-2 Item 15 - Ladder aounted on the reactor building wall in the EFW-P-2B pump roo Ladder mounting bolts will be replaced to assure SSE qualificatio The licensee was requested in NRC letter, dated June 21, 1984, and discussed in the licensee's plans in their letter dated August 23, 1983, to review the emergency power bus loadings as affected by the long term EFW modification .3 Findings / Conclusions The inspector reviewed the documentation associated with the above scope and found them acceptabl The inspector visually observed that the condensate storage tank level transmitters were located in the intermediate building at ele-vation 295 feet. The redundant tratismitters associated with the "B" condensate storage tank, LT-1062 (red) and LT-1063 (green), were located and connected to the "B" condensate line at the west end of the hallway which provides access to the EFW pumps. The redundant transmitters associated with the "A" condensate storage tank, LT-1060 (red) and LT-1061 (green) were located and connected to the "A" con-densate line at the east end of the hallway. The transmitters were identified and installed as specifie the licensee stated that level correction would be applied since the transmitter location is lower than the storage tank. The inspector observed that there was no potential interacting systems in the area of the transmitters to cause common failures of redundant transmitters. The inspector ver-ified the cables and conduit identification associated with the transmitters was as specified. The inspector observed that the two dual condensate storage tank level indicators were not installed in the main control boar .

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The inspector visually observed the turbine-driven EFW pump overspeed position indication device was mounted as specifie Changes had been made to the original mounting to assure operability (FCR No. C 038527). These changes were discussed with the QC personnel, who witnessed the installation and testing and was found acceptabl .

Also, a change to the annunciator window for the alarm was made (FCN No. C 033044). The original location was on annunciator D-2- The

"0" annunciator is located some distance from the EFW system con-trols. The new location on annunciator J-3-4 is directly in front of the EFW controls and is acceptabl The inspector observed that Radiation Monitor RM-A-2 had been seis-mically anchored to the floor in the instrument air compressor IA-PIA cubical, elevation 295 feet in the intermediate building. The in-spector walked down the EFW system area at elevation 295 in the in-termediate building and did not identify any additional potential system interactions, including seismi The licensee's response to the emergency power bus load was contained in letter 4211-84-2304, dated January 11, 1985, to J. F. Stolz, NR The licensee stated that the results of GPU computer program GEE 01400,

"TMI-1 Electrical Load Analysis," indicated that the ESF system mod-ification did not cause bus loadings exceeding acceptable limits under worst case loading conditions. The inspector was informed in November 1986 by Engineering at the Parsippany Office, that GPU has a new in-house computer program, " DAPPER" (Distribution Analysis for Power Planning, Evaluation, and Reporting). The inputs for TMI-1 have been completed into the DAPPER program. The inspector noted that loads were added between January 1985 and 1987 and the DAPPER program was properly used. This is unresolved pending additional Region I review for adequacy of the above-noted computer program (289/87-02-03).

5.4 Summary Within the scope of this review, the licensee design is essentially in conformance with commitments made and NRC-established design cri-teri Items remaining to be reviewed in future inspection of the EFW system upgrade are (289/85-20-01):

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review the calibration and test of the condensate storage 'ank level indication and low-low level alarm;

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verify the removal of EFW loop "B" pipe support EF-111, includ-ing the connection between the stair framing and the building column in the containment (reactor building) elevation 331-345 feet near Column C116;

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verify that the rod support of support EF-16 was replaced with a strut support of EFW loop "B", elevation 331-345 feet in the

. containment;

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verify that the ladder, located on the reactor building wall in the EFW-P28 pump room, mounting bolts have been replaced to assure seismic qualification; and,

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safety grade initiation and control function (heat sink protec-

, tion system (HSPS).

6. Exit Interview 1'

The inspectors discussed the inspection scope and findings with the licen-see management at a final exit interview conducted February 6,1987. An interim exit occurred on January 30, 1986, for the IST program revie Senior licensee personnel attending the final exit meeting included the i followin P. Christman, Manager, Plant Administration, TMI-1 G. Kuehn, Manager, Radiological Controls, TMI-1 C. Incorvati, TMI-1 Audit Manager L. Robinson, TMI-1 Communications C. Shorts, Manager, Technical Functions, TMI-1 C. Smyth, Manager, TMI-1 Licensing R. Toole, Operations and Maintenance Director, TMI-1 The inspection results as discussed at the meeting are summarized in the

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cover page of the inspection repor Licensee representatives indicated

! that none of the subjects discussed contained proprietary or safeguards

, information.

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7. Unresolved Items are matters about which information is required in order to ascertain whether they are acceptable, violations, or deviations. Unre-solved items discussed during the exit meeting are addressed in paragraphs 2.2.5 and 5.3.

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