IR 05000289/1986019

Safety Insp Rept 50-289/86-19 on 861003-31.Violation Noted: Failure to Include in 860228 Rept Assessments of Doses from Effluents to Individuals Working Inside Site Boundary

ML20207N229
Person / Time
Site:	Three Mile Island
Issue date:	01/07/1987
From:	Conte R, Dante Johnson, Young F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20207N212	List: IR 05000289/1986019 ML20207N210 ML20207N216
References
50-289-86-19, NUDOCS 8701140063
Download: ML20207N229 (48)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report N /86-19 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 i Inspection Conducted: October 3, 1986, to October 31, 1986 Inspectors: R. Conte, Senior Resident Inspector (TMI-1)

S. Davidson, Radiation Specialist, RI D. Johnson, Resident Inspector (TMI-1)

M. Miller, Radiation Specialist, RI W. Oliveira, Reactor Engineer, RI

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J. Rogers, Resident Inspector (TMI-1)

R. Struckmeyer, Radiation Specialist, RI A. Weadock, Radiation Specialist, RI

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F. Young, Resident Inspector (TMI-1)

Reporting Inspector: O M 8cb D. Johnson, Resident Inspector (TMI-1)

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Date Reviewed By: bT b , / 7 F. Young, Senior Resident Inspector (TMI-1) D' ate'

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l Approved By: b R. Conte, Chief

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Date

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Reactor Projects Section No. IA l

Division of Reactor Projects j

Inspection Summary:

Resident and region-based NRC staff conducted safety inspections (369 hours0.00427 days <br />0.103 hours <br />6.10119e-4 weeks <br />1.404045e-4 months <br />)

of power operations, focusing on plant and personnel performance. Speciff-

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

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lance areas were: reactor coolant pump seal problem and power coastdown

, evolutions. Other items included: corporate pre-outage work; non-licensed i training for engineers and technical staff; radwaste program review and j independent effluents measurements; safety grade emergency feedwater in plant

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I 8701140063 870107 '

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review; technical and safety reviews of licensee procedure changes; pre-outage health physics review; and, licensee actions on previous inspection finding Inspection Results:

For the activities sampled, the inspectors noted that, in general, the licensee properly completed evolutions, maintenance, and surveillance activities consis-tent with regulatory requirements. The power coastdown evolutions to end of cycle was accomplished in a properly controlled manner although the procedure change implementing the required operator actions was approved late. The li-censee action on previous inspect:nn findings was acceptabl Overall, there is proper implementation of radwaste and radioactive monitoring programs. The licensee appears to be prepared for the upcoming refueling outag One apparent violation of NRC requirements was identified in the effluents measurements reporting program (paragraph 8.5).

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DETAILS Introduction and Overview 1.1 NRC Staff Activities The overall purpose of this inspection was to assess licensee activi-ties for the power operation mode as they related to reactor safety and radiation orotection. Within each area, the inspectors documented the specific purpcse of the area under review, scope of inspections, along with appropriate conclusions. The inspector made this assess-ment by reviewing information on a sampling basis through actual observation of licensee activities, interviews with licensee person-nel, measurement of radiation levels, or independent calculation and selective review of listed applicable document During this period, regional health physics inspectors completed remaining inspections for the close of the SALP (Systematic Assess-ment of Licensee Performance) II period (May 1,1986, to October 31, 1986). The independent effluents measurements van was on site con-ducting evaluations of selected radwaste and Reactor Coolant System (RCS) radiochemistry samples. A programmatic review of the licensee's radwaste programs was also conducted. Additionally, a region-based inspection was conducted of the licensee program for training of technical staf .2 Licensee Activities During this period, the licensee operated the plant at full power until the commencement of power coastdown during the last week of the inspection period. A problem with the "A" Reactor Coolant Pump (RCP)

No. I seal leakoff was noted but was closely monitored by the licensee. Licensee preparation for the coming outage continued with emphasis on equipment installation for the containment integrated leakrate test (CILRT). Plant Operations 2.1 Scope of Review The NRC resident inspectors periodically inspected the facility to determine the licensee's compliance with the general operating requirements of Section 6 of the Technical Specifications (TS) in the following areas:

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review of selected plant parameters 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

. barrier integrity, extinguisher checks, and housekeepin Because of additional resident office coverage at this facility, more 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 defec-tive 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

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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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Selected licensee planning meetings Maintenance

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DH-V-4A/B and DH-V-5A/B motor replacement Surveillance

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Reactor protection channel checks

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EF-P-1 and MS-V-6 testing

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Thermal power heat balance calculations

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Control rod exercise evolution As a result of this review, the inspectors reviewed specific areas in more details as described in the sections that follo '

2.2 Findings 2. Power Coastdown On Wednesday, October 22, 1986, the licensee commenced Reactor Coolant System (RCS) average temperature reduction due to boron depletion. At end of cycle, boron concentra-tion was 2.0 ppm and boron removal using the installed capability of the deborating demineralizers and batch addition of pure water was not sufficient to maintain average temperature at 579 F. Using the Integrated Control System (ICS) T-AVE controller, T-AVE was reduced in 1 F increments until approximately 574 + 2 F was reached at which time it was necessary to reduce power in order to add positive reactivity to account for fuel burnu Power reduction commenced on October 28, 1986, and, at the end of the reporting period, was at approximately 95 percent. The unit will be shutdown commencing October 31, 1986. This evolution will be documented in the next routine inspection repor The licensee issued Temporary Change Notice (TCN) 1-86-0151, dated October 21, 1986, to Operating Procedure (0P) 1102-4, Revision 42, to give guidance to plant operators for the coastdown evolution. The inspectors reviewed the TCN and associated safety evaluation (SE) required by Technical

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Specifications (TS). Generally, the procedure change and safety evaluation addressed the appropriate parameters tha would change during the coastdown evolutions. One portion of the safety evaluation stated that T-cold would be reduced to the Final Safety Analysis Report (FSAR) design nominal value of 554 F and that T-AVE would be reduced to 574 In fact, T-cold was reduced to an average value of 551 F during portions of th2 coastdown evolution and T-AVE reached 573 F. The inspectors questioned licensee engineering per-sonnel about this apparent discrepancy. The licensee representative acknowledged that a 2 F tolerance should have been placed on the above limits for operator contro A second procedure change was initiated to allow the opera-tors a 2 F band about 574 F T-AVE in which to operate. The inspectors had no safety concern with operation in this manner. With respect to the minimum FSAR required T-cold of 554 F, the licensee responded that this was not a design low limit for safety reasons but an operational limit asso-ciated with the T-AVE reduction of 5 F. In fact, the minimum T-cold allowed for critical operation of the plant is 525 F, which is specified in the technical specification to assure a negative moderator temperature coefficient of reactivity at all allowable boron concentrations and power level The inspector also conducted discussions with operators on the conduct of this evolution. The operators performed the evolution over the two-week period with no adverse conditions. Operators were aware o' possible system param-eter changes; e.g. , once-through steam generator (OTSG)

level, feed flow, T-hot, T-cold, and reactor power imbalance, and seemed prepared to control the evolutio The inspector noted that procedural guidance and associated safety evaluation could have been enhance The inspectors also noted that the actual procedure TCN was not delivered to the control room until the day that T AVE reduction had already commence Generally, the evolution of power coastdown with associated T AVE reduction, complete rod withdrawal, and power reduc-tion was conducted in a safe manner. No safety issues were identifie . "A" RCP Seal Problems The "A" Reactor Coolant Pump (RCP) No. I seal leakoff flow exhibited a marked decrease in flow rate during the inspec-tion period. Normally, the flow rate from a RCP No. I seal is in the range of 1-5 gallons per minute (gpm). The flow i

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rate from the No. I seal of the "A" RCP had decreased to the range of 0.3-0.5 gpm at the end of the inspection period. The flow rate had been constant at approximately 2.0 gpm for the past few months of operation. Also, evi-dence of seal problems was a increase in the leakoff temperature to approximately 181 F. Normally the seal leakoff temperature would be in the range of 120-140 Additionally, the labyrinth seal delta-P had increased to 60-inch H 2O from a nominal 30-inch H 2O that had been present during the previous few months. This was an indi-cation that a larger portion of the seal injection flow was entering the RCS via the labyrinth sea The licensee commenced an increased monitoring of the above seal parameters, including trend recorder monitoring of seal leakoff temperature and pump radial bearing temperature. A limit of 190 F was established for the maximum allowable temperature that would be allowed prior to removal of the pump from service. Aoditionally, a limit of 0.2 gpm was established as a lower limit for seal leak-off flow. At the end of the reporting period, the flows and temperature had appeared to stabilize. The licensee is planning to shut the pump down as a result of the upcoming outage. The operators were provided with amplifying guid-ance in the Operations Department Night Order Book on appropriate steps to follow in order to take the "A" RCP out of service should the need arise. The licensee con-sulted with the seal vendor to arrive at the above limitation The inspectors reviewed licensee action in this area, con-sisting of review of RCP seal leakoff data, discussions with operators and review of the RCP technical manua Although the pump was running at the upper allowable band for two of its operating parameters, it appeared to the inspector that continued operation within the established bands would no*. present a safety concern. Ample time would be available to reduce reactor power and secure the pump and prevent pump damage or challenges to plant safety systems should the seal parameters vary outside their allowable bands. A review of the recommended technical manual parameters revealed that the pump was operating within allowable guidelines. Additionally, the plant is designed to respond to the loss of one RCP and automati-cally run back to 70 percent reactor power and the No. 2 RCP seal is designed to function against the resulting delta-P should the No. I seal fai The inspectors concluded that operation at full power with the No. I seal of the "A" RCP in its present condition did not pose a significant safety concern. Additionally, the

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pump will be required to operate for a short period of time in this condition until the commencement of the 6R refuel-ing outage at which time the pump seal will be repaire . Reactor Coolant System (RCS) Leak Rate The inspector selectively reviewed RCS leak rate data for the month of October. The inspector independently calculated certain RCS leak rate data reviewed using licensee input data and a generic NRC " BASIC" computer program "RCSLK9" as spect-fied in NUREG 1107. Licensee (L) and NRC (N) data are tabulated belo TABLE 1 RCS LEAK RATE DATA (All Values GPM)

DATE/ TIME (NUREG 1107) CORRECTED DURATION Lg Ng Ng N U 'O 10/8/86 .201 .21 .09 .0144 .0052 2 Hours 10/13/86 .2748 .28 .08 .1844 .1758 2 Hours 10/23/86 .2973 .30 .1044 .1030 2 Hours G = Identified gross leakage U = Unidentified leakage L = Licenset calculated N = NRC calculated Columns 2 and 3; 5 and 6 correlate 1 0.2 gpm in accordance with NUREG 110 (Nu corrected by adding 0.1044 gpm to N due to leakoff from the No. 3 seals of the RCP's.)

The inspector concluded that the licensee leakrate determinations were in good agreement with those calculated by the NRC staff progra . ESF Ventilation for the Fuel Handling Building 3.1 Background The inspector conducted a review of the design and operation of the new fuel handling building Engineered Safeguards Features (ESF)

ventilation system. The purpose of this review was to determine if, during operation, any failure of a single component could cause both

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trains of the system to become inoperable. This review was accom-plished in accordance with guidance in Regional Temporary Inspection Instruction (RI-86-01), " Inspection of Standby Gas Treatment System."

This temporary instruction was generated in response to a situation at another Region I facility (Pilgrim) where an engineering review of the standby gas treatment system (SGTS) revealed the following condition. The failure of one temperature element in the fire sup-pression system could have caused both trains of the SGTS, in conjunction with loss of power to fail open dampers in the low flow air bleed system, to become inoperabl This would have been due to the deluge system actuation for both trains. This condition could have led to an unfiltered releas .2 NRC Review The fuel handling building ESF ventilation system at TMI-1 is presently still under construction. It is required to be operable prior to commencing fuel movement in the upcoming outage. This system has been installed to mitigate the consequences of a postu-lated fuel handling accident as described in the FSAR. The inspec-tor reviewed the applicable system design description SDD-TI-845B, Division II (Revision 1) and Burns and Roe Drawing M0004, Revision 4, " Flow Diagram ESF Ventilation System - Fuel Handling Area." As yet, no operating or surveillance procedures have been implemente The system testing has not been completed and, as yet, no maintenance history is availabl .3 Findings The ESF fuel handling building ventilation system is a two train system with each train capable of supplying 100 percent of required flow. The system also incorporates a 10 percent low flow bleed through the non-operating filter bank when one fan is operating. The system uses piping, motor-operated " fail as is" dampers, charcoal and particulate absorbers, roughing filters, air heaters, motor-operated

" Vortex" flow control dampers, single speed fans, and appropriate instrumentation for detection of filter delta-P, flow and temperatur A radiation monitor for noble gas and detection capability for particulate and iodine activity is provide A review of postulated failure of instrumentation used for control of the system revealed no situation where a single detector or instrument failure could cause a loss of both trains. The fans are manually operated and are required to be manually loaded onto the emergency diesel generators upon loss of off-site power. Failure of low flow bleed motor-operated dampers will not result in improper flow distribution. Two dampers (sized for 100 percent flow) in parallel from the fuel handling building with separate electro-thermal links and combustible gas analyzers will shut each damper on high temperature or high hydrocarbon levels. Fire protection for the system is provided in the form of temperature sensing elements down

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stream of the charcoal filters. No automatic initiation of fire protection systems is present. Any fire must be mitigated using installed adjacent hose stations and fire brigade personne .4 Conclusion The inspectors review of this ventilation system revealed no condi-tions where a single instrument or component failure would cause a total failure of the system. The system appears to be designed with appropriate redundancy and independenc Should subsequent testing and operation by the licensee identify any concerns, they will be documented in future reports. The inspectors continue to follow and observe the pre-operational testing of this syste . Failures of General Electric Breakers Several failures of General Electric (GE) type AKF-2-25 breakers to open upon demand have occurred during the last three years at various facilities. This concern was identified and is being tracked as Regional TI-86-0 The use of this type breaker varies with plant design. At TMI-1, GE AK-2 type breakers are used in safety-related reactor protection system (RPS) and main generator field exciter unit. Specifically, RPS a.c. breakers CB-10 and CB-11 are GE AK-2A-25 breakers. The four RPS breakers, CB 1 through CB-4, and the main generator field exciter unit output breakers are GE AK-2A-15, a similarly designed breake The inspector reviewed the application and failure history of this type breaker for TMI-1 in safety-related systems to determine if there was a generic desigr. problem. The inspector did not review licensee's use and control of this type breaker in non-safety-related application A review of the machine ry history noted that RPS a.c. breakers have failed to function properly on several occasions. Only on one occasion did a breaker of this type fail to open due to the mechanical portion of the breaker. The mechanical failure occurred on January 14, 1986. This event was reviewed by the NRC and is documented in NRC Inspection Report 50-289/86-01. The licensee reviewed the GE response on the cause of the failure and determined that their characterization was accurat The remaining failures to open were associated with the electronic /

electric tripping circuitr The failures of the electronic tripping devices can be characterized as single component failures and are unique to the special circuitry at TMI-1 on this type breaker application. For the main generator field exciter breaker, which is not a safety-related application, no maintenance problems have been noted. The breaker falls under the licensee's balance of plant maintenance program and is tested on a less frequent basi .

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The inspector reviewed the licensee's preventive maintenance and surveil-lance testing to determine what was being performed for RPS breakers in these areas. In the area of surveillance, all RPS breakers are tested on a monthly basis. After the mechanical binding on the CB-10 and CB-11 breakers was resolved, no significant problems were noted with any RPS breakers. When a breaker is tested, surveillance. test data is collected and trended by on site engineering for adverse conditions. In conjunction with B&W Owners' Group, the licensee participated in a B&W Owners' Group Reactor Trip Breaker Symposium on January 24, 1986, to discuss improve-ments in reactor trip breaker reliability. The information from the symposium is being used in trend analyses data on breaker performanc The inspector found that the licensee was taking a positive approach toward improving the reliability of RPS breakers. Maintenance personnel were well versed in maintenance and operational problems that have been noted at other B&W facilities. The preventive maintenance is being per-formed and adequately documented. In general, the licensee's program in this area has been successful in obtaining adequate reliability of RPS breaker . Corporate Inspection 5.1 Refueling Outage Preparations / Control The inspector conducted a review at the GPUN corporate office in Parsippany, New Jersey, of TMI-1 Cycle 6 refueling outage prepara-tions and control. Part of the inspection also included a review of the project status meeting and refueling outage meetings on sit The inspector was to:

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assess overall preparations and control for the Cycle 6 refuel-ing;

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review problem areas or incomplete engineering work with dates for resolution and/or completion and also the knowledge of design or project engineers on these issues; and,

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in conjunction with an NRR project manager for TMI-1, review outstanding licensing actions and those actions needed to be accomplished by the licensee and/or NRC staff for the Cycle 6 startup (after Cycle 6 refueling).

In conjunction with the corporate engineering training review, the inspector interviewed cognizant engineers or project engineers overseeing the various engineering topics selected, and he also met with the PWR Licensing Manager and Long-Range Planning Manager to discuss outstanding licensing actions. The inspector also reviewed the licensee's internal list of "GPUN/NRC Licensing Activities."

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The inspector found that the various preparatory activities of the licensee reflect well on the overall coordination effort. There was evidence of a high level of management attention to the refueling and startup preparations. Presentations to the Board of Directors were being prepared with respect to schedular, procurement, and design problems on outstanding items. Engineers interviewed were knowledge-able in their related area The most notable engineering problem was a design deficiency with the reactor water level instrumentation. Licensee representatives identi-fied this during a design verification. The present design of the reference leg would cause varying pressure situation in the reference leg and, therefore, cause flashing of water in the reference leg, giving an erroneous indicatio Engineers were working on a solution and the final design will be submitted to NRC staff to support the staff's pending safety evaluation. The resident inspector will con-tinue to routinely follow this ite The licensee representatives prioritized the list of licensing actions into two categories. One list was those for which NRC staff action was needed for Cycle 6 startup; e.g., Cycle 6 core reload Technical Specification (TS). The other list was those items which were desired but not essentially neede Examples include TS on TMI Task Action Plan (TAP) items required to be implemented for Cycle 6 startup. Operability would be assured by procedures. The listing prioritization is tentatively acceptable pending further review by NRC staf The inspector also noted that a majority of licensing action items on the list are highly dependent on licensee submittals scheduled to be issued to NRC staff within the next two month .2 Corporate Audits While at the corporate office, the inspector selected certain Quality Assurance (QA) audits of corporate activities, which included audits of contractors. As an acceptance criteria, the inspector used selec-ted sections of ANSI N-45.2-1971, Section 19. The inspector focused on the:

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details and thoroughness of the audits;

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findings appropriately classified;

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auditors were responsive to findings and were independent in determining the acceptability of or unacceptability of respons-es; and,

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responsiveness of Quality Assurance (QA) organization to recent NRC Performance Appraisal Team (PAT) finding .

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In addition to interviews with QA personnel, the inspectors reviewed the below listed audit files, which included responses to audit finding No. 0-TMI-85-03, conducted August 26, 1985, to October 1, 1985, Fire Protection

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No. 0-COM-85-08, conducted December 9, 1985, to April 11, 1986, Technical Function and Architect Engineer Calculations

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No. 0-0C-85-06, conducted September 23, 1985, to October 24, 1985, Headquarters, Oyster Creek Site Engineering Projects

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No. 0-COM-85-07, conducted November 6 - 11, 1985, Teledyne Isotopes Midwest Labs

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No. 0-TMI-86-01, conducted January 22 - 23, 1986, Design Control, TMI-1 Projects Gilbert Associates

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No. 0-TMI-86-04, conducted May 20 - 22, 1986, Mini-Mod Program (TMI-1)

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No. 0-TMI-86-05, conducted May 20-22, 1986, Document Control The audit filea reflected thorough reviews with detailed checklists usedwhichcor{elatedtoappropriateregulatoryrequirements. The reviews were critical of licensee organizations and contractor It appeared tha't findings were appropriately characterized. An initiative was noted in that strengths and weaknesses were identi-fied along with recommendations. Further, audits are becoming more technical-based in distinction to management- based. For the most part, audited organizations were responsive to the recommendations, although no response was require The inspector also noted a number of QA audit findings were precur-sors to the PAT I findings. On the other hand, PAT I follow-up audits were responsive to NRC staff findings and they provided the

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licensee with performance insight As an example, a post-PAT I audit dealing with design verification confirmed a weak technical function procedure that was not specific on when design verification was to be done. As a result, design g verification was excessively delayed or not completed with no action

, to track completion. On the other hand, a post-PAT I audit confirmed

, a sound program with respect to the " mini mod" process. Minor imple-

- mentation problems were noted as with PAT I. The inspector noted that the implementation problem also appears related to the company-

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3 Another post-PAT I audit identified relatively few deficiencies with the licensee's " Computer Assisted Records and Information Retrieval System (CARIRS). This contradicted the PAT I and most recent PAT II findings. This area will be further reviewed upon NRC follow-up to-the PAT II finding .3 Procedure Compliance Task Group In their response letter, dated April 21, 1986, to the interim Systematic Assessment of Licensee Performance (SALP), the licensee addressed the issue of proper use and quality of procedures. They presented a nine point list of corrective actions taken or planne Many of these actions were acknowledged and/or verified by the NRC staff in a recent inspection which addressed the closeout of related inspection findings in this are During this inspection, the inspector focused on licensee actions related to Corporate Procedure Compliance Task Group (PCTG) (referred to as Corporate Task Force in the licensee's letter). The PCTG was established to review safety-related procedures and their implementa-tion in all divisions. The PCTG issued various reports between May and August 1986 with a Corporate Summary Report (undated) issued in August 1986. The inspector reviewed selected key sections of these reports and discussed licensee's findings, conclusions, and recom-

mendations with cognizant licensee personnel. The inspector's view of licensee results of PCTG effort are documented below.

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The PCTG reviewed and conducted interviews with personnel from

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eight of ten division within GPUN with respect to: (1) identifying the existence of division policies or procedure adherence and the consistency of that policy with the corporate policy; and, (2) survey

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personnel attitudes toward procedure adequacy and adherence. The two eliminated divisions were Comptroller General and Communication Divi-sion, which do not have direct responsibility in implementing the

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Quality Assurance Plan (QAP) mandate for procedure adequacy and adherenc In addition to the administrative procedural reviews

noted above, the PCTG interviewed selected personnel from three levels in the organization

workers (doers); supervisory; and, upper / middle managers within the division. A consistent set of questions related to the area of concern were asked to avoid biasing answers. Senior licensee personnel, especially within the Nuclear Assurance Division, participated in the PCTG as reviewer The PCTG found that the division policies reflected a varied implementation of the corporate policy on strict compliance with procedures. Some division policies were inconsistent with the corporate policy in that vague words; such as "if appropriate" or

"should," were used in the division policies. Certain other divisions had a " verbatim" policy. In other divisions, no policy existed in which the corporate policy was implied.

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The inspector also noted that the existing corporate policy had a recommendation ("should") on stopping the implementation of an evolution if the procedure could not be properly implemented. As with the finding of PCTG and by past inspections, the inspector concluded that the strict procedure compliance policy was not uniformly implemented by the divisions of GPUN due, in part, to the above-noted reason The PCTG also noted that the corporate procedure was not specific on how minor administrative changes ("on-the-spot" or " temporary changes") were to be handled. The inspector noted that this may be why one operating division; such as, TMI-1, could not effectively operate under a verbatim compliance policy. Thus, procedure change flexibility was built into the TMI-1 temporary / permanent change methodology. The inspector also noted that, in the past, the established change system was not always properly implemented and this was one of the reasons for previous negative comments by NRC staff in previous SALP's. On the other hand, other divisions opted not to establish temporary change methodologies and these minor procedure inadequacies challenged personnel to properly implement those procedures. The licensee will need to establish a firm and consistent policy among divisions, coupled with providing the divisions with flexibility to establish a viable "on-the-spot" or

" temporary" change methodology, in accordance with the Technical Specifications. The PCTG recommendations are oriented to this en The recommendations of the PCTG were divided into short-term and long-term items. The short-term items consist of disseminating and reinforcing personnel responsibility to properly follow procedure and reinforcing company receptiveness to changing or improving procedure Other actions include training and nuclear assurance enhanced moni-toring of quality deficient reports (QDR) attributable to procedure nonadherenc The long-term actions include adopting the PCTG recommended procedure adherence policy, which appears to be firm but flexible. Other actions include: pre-implementation procedure walkthrough; holding personnel accountable through the Personnel Evaluation Process; procedure writer's guide; additional training; regular monitoring of QDR performance with respect to established goals. The licensee is actively reviewing these recommendations to adopt acceptable and viable corrective actions. The President of GPUN indicated that the essential elements of the PCTG recommenda-tions are being rigorously adopted and implemente The inspector views the PCTG report very favorably with respect to licensee response on a somewhat complex, repetitive, and negative issue. The inspector is encouraged by the recommendations of the PCTG and licensee upper management attitude and adoption of the essential elements of those recommendations. The effectiveness of licensee corrective actions remain to be verifie c

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The inspector notes no target completion date specified by the licensee for these corrective actions. It is recognized that a completion date for certain long-term items is hard to pinpoint, but the general feeling of licensee managers interviewed and who were involved with PCTG is that actions should be complete within the next year. This is unresolved pending licensee finalization and implementation of licensee completion of corrective actions in response to PCTG recommendations and pending NRC Region I review (289/86-19-01).

5.4 Conclusion The licensee appears to be prepared for the refueling outag However, a number of licensee and NRC staff actions are needed for the startup following this outage, which is scheduled for March 198 Corporate QA audits were appropriately conducted and contain many licensee initiatives. The QA organization is involved with GPUN upper management in resolving the procedure adherence proble . Safety-Grade Emergency Feedwater In plant Review As a result of NRC staff inspection of spaces housing the emergency feed-water (EFW) system and EFW system components, a number of questions were formulated by the staff. These questions were attached to NRC Inspection Report No. 50-289/85-26; and, in the cover letter to that inspection report, the licensee was asked to respond to those questions within sixty days from the date of that letter (December 12,1985).

The licensee responded to those questions in a letter, dated February 11, 1986. The questions, the licensee response, and the residual issues are addressed belo Item 1: The NRC staff expressed concern about the seismic instal-lation of ducting, piping, and other components installed above the redundant two-hour backup instrument air banks in the "B" diesel generator room. The licensee responded by indicating that the air supply ducting was upgraded to seismic criteria in accordance with Final Safety Analysis Report (FSAR) commitments. No piping is above the air bank and the remaining cable and conduit, although not specifically seismically mounted, by engineering judgement, would not fall and render the banks inoperable. The inspector noted that the basis for this engineering judgement was not detailed. This area is unresolved pending further review by Region I during its seismic /

seismic interaction review of the Cycle 6 safety grade emergency feedwater system (289/86-19-02).

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Item 2: The NRC staff needed additional information on the licen-see's proposed change in failure mode of the EFW flow control valve The latest design was for the valves to fail closed on a loss of

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instrument air instead of'failing opened as discussed during the TMI-1 restart hearing. The licensee responded with their evaluation of the acceptability of this design. The latest design was to allev-iate concerns with respect to control problems on valves failing open. This design change is significant and currently under review by NRR in which additional information has been requested and received from the licensee. NRR will disposition this matter in their Safety Evaluation Report (SER) for the final design review of Cycle 6 EFW system.

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Item 3: The staff expressed concern about various indications o flooding in the intermediate building when it learned that the tendon access gallery flood detection would not be completely safety grad The licensee's response described their indications and actions for intermediate building flooding using other instrumentation available in the control room. The pit level indicator will have Class IE components and be environmentally qualified for submergence and for a feedwater break, but they will not be redundant. The licensee assured NRC staff that a simple indicator malfunction would not lead to immediate operator action. Accordingly, their response is accept-able, but licensee final procedures on the operation of the pit level indication will be reviewed to confirm licensee assumptions and statements. This is unresolved (289/86-19-03).

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Item 4: The staff expressed concern on the additional impact on safety with the storage of hydrogen and oxygen calibration gas bottles in the intermediate building in the vicinity of the EFW system. The licensee provided a fire hazards analysis (FHA) summary, and they concluded that the installation was acceptable per codes and the FHA. The analysis is tentatively acceptable; however, this item is currently being followed by an unresolved item (289/85-25-04).

This will be reviewed during the upccming 10 CFR 50 Appendix R revie In summary, the licensee appropriately responded to the staff concern Residual issues will be reviewed in a future inspection or by the staff's SER on the final design review of EFW syste . Radiological Preparations for the 6R Outage Inspection effort in the area of radiological controls focused on a review of the licensee's activities in preparation for the upcoming 6R outag This included a review of instrumentation, staffing specialized training, and ALARA (as low as reasonably achievable) plannin .1 Facilities and Instrumentation Licensee instrumentation and facility upgrades planned for the outage are generally directed towards improving personnel access and egres Two additional control points with full computer access capabilities have been designed and will be used during the outage. Additional

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portal alarms will be used to speed egress from the areas. In addi-tion, a second gamma-spectroscopy system, dedicated for the counting of health physics samples, has been procured. This system should resolve the difficulties observed during the last outage resulting from a large sample counting backu .2 Outage Staffing and Training Licensee staffing upgrades for the outage include the addition of eight senior radiological controls (RC) technicians, twenty techni-cian assistants, two group radiological controls supervisors (GRCS),

and an additional radiological engineer. These upgrades appear adequate to handle the planned scope of work. The inspector noted that all additional personnel, with the exception of one senior con-tractor technician, were drawn from Unit 1 or Unit 2 personnel and are, consequently, already familiar with station procedure A potential weakness among the licensee's radiological controls staff is their lack of experience with post-critical plant refuel-ing activities. The licensee has recognized this situation and the following actions have been taken to ensure staff familiarity with refueling activities. A six-hour refueling training module was developed and presented to all field operations technicians during their cycle training. Personnel from the field operations and radio-logical engineering staff were also recently sent to another B&W utility to observe ongoing refueling activitie .3 ALARA Planning Discussions with licensee radiological engineers indicate ALARA preparations for the outage were initiated in a timely manne " Cognizant" engineers have been assigned ALARA responsibilities for each major task occurring inside the radiologically controlled are The engineering staff has developed a radiological controls " punch list," which details the numerous procedure changes, equipment procurements, etc., which must be performed prior to the outag The " punch list" is used to ensure each item is assigned to, tracked, and closed by the cognizant enginee The inspector discussed radiological preparations for several specific jobs with the cognizant radiological engineers and deter-mined that planning for each major job had been extensive and appropriate " lessons learned" had been incorporated, when applica-ble, from previous experien:e. The inspector did note that these preparations were not formally documented on ALARA reviews and that specific job exposure estimates had not been developed as of three weeks prior to the outage. The licensee indicated this documenta-tion had been held pending the receipt of additional man-hour estimates from the work groups and current survey information and would be available by November 1, 198 .

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7.4 Dosimetry Discrepancy Investigations On October 15, 1936, licensee radwaste personnel were involved in the transfer of a spent filter from a transfer pig to a shielded filter cask. The primary worker, located on the platform around the filter cask, was wearing a self reading dosimeter (SRD) and thermoluminescent dosimeter (TLD) on the chest, a Xetec digital dosimeter on the right upper arm, and extremity TLD's on both hand A health physics (HP) technician was providing constant coverage for this evolution. At the conclusion of the work activity, it was noted that the worker's chest SRD indicated a dose of 60 mrem, while the Xetec worn on the upper arm indicated a dose of 211 mrem. At this point, a Dosimetry Investigation Report (DIR No.86-131) was initi- l ated by the licensee's HP technician to investigate the discrepanc NRC review of the above incident included the following:

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review of RWP 032124, " Dispose of M/U Filter 2A," and associat-ed ALARA review 86-03-02; and,

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discussion with cognizant radiological engineering and protec-tion staf Licensee investigation into the discrepancy included reading the individual's whole body and extremity TLD's (thermoluminescent dosimeter) and checking the response of both dosimeters. These results indicated the initial discrepancy appeared accurate; i.e.,

the upper arm received a higher exposure than the chest. The inspector discussed dose assessment and dosimetry record update with the cognizant radiological engineer and identified that the appro-priate exposure value was being credited to the worke The 1:1spector noted that placement of the whole body TLD on the individual's chest appeared inappropriate; in that, it did not record the highest whole body dose during the filter transfer operation. The inspe~ctor noted that the controlling radiation work permit (RWP) and ALARA review for this task did not address place-ment of the TLD on the body. The licensee indicated that the HP technician covering the job typically directed placement after performing a surve Identification of the discreparcy was made promptly by licensee personnel and initiated an investigation report, which requires additional follow-up by the licensee. Hcwever, during the time frame of this inspection, licensee investigation into the discrepan-cy had been preliminary at best. No assessments as to the adequacy of their dosimetry placement for this task had been mad .

.

Due to the licensee's identification of this discrepancy, a determi-nation as to the adequacy of the licensee's whole body exposure monitoring during the filter transfer evolution will remain unre-solved pending licensee investigation and corrective action (289/86-19-04). Adequacy of the licensee's follow-up will be reviewed during a subsequent inspectio .5 Conclusion Overall, the licensee's planning and preparation for the upcoming outage appears to be effective. Additional personnel and the off-site training received by radiological engineering and field operations personnel should provide an adequate staff to accomplish i outage-related activitie . Radwaste program Review and Confirmatory Measurements Evaluation 8.1 Confirmatory Measurements Program

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8. Reactor Coolant Water Chemistry The inspector reviewed selected procedures and records for chemical and radiochemical analyses of primary and secondary coolant. The analyses include those for dissolved oxygen, fluoride, chloride, I-131 dose equivalent, E-bar, and gross activity in the primary coolant, and I-131 dose equivalent in the secondary coolant. The inspector also reviewed the

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licensee's method for scheduling and verifying that required surveillances are performed. The review indicated that required surveillances were performed on time and that the

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results of analyses were within Technical Specifications limits. The inspector noted that many analyses are done more frequently than require . Confirmatory Measurements Data During the inspection, actual liquid and gas samples, as well as simulated particulate filter and charcoal cartridge samples, were split between the licensee and NRC for the purpose of inter-comparison. Where possible, the split samples are actual effluent samples or in plant samples, which duplicate counting geometries used by the licensee for effluent sample analyses. The samples were analyzed by the licensee using normal methods and equipment and by the NRC:I Mobile Radiological Measurements Laboratory. Joint analyses of actual effluent samples are used to verify the licensee's capability to measure radioactivity in effluent samples with respect to Technical Specification requirements and other regulatory requirement <

r y .__.-,_-.---__v ,_.___.-___---r.. ,-- - - _ _ - - - - , - , _ _ y ,-. _ , _ - _ . _ , - - . _ . _

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In addition, a liquid effluent sample was sent to the NRC reference laboratory, Department of Energy Radiological and Environmental Sciences Laboratory (RESL) for analyses requiring wet chemistry. The analyses to be performed on the samples are Sr-89, Sr-90, gross alpha, tritium, P-32, and Fe-55. The results will be compared with the licen-see's results when received at a later date and will be documented in a subsequent inspection repor The results of the sample measurements comparison completed during this inspection indicated agreement, with the excep-tions noted below. The results of the comparisons are listed in Table I and the agreement criteria are included as Attachment The first attempt to compare measurements of a sample from the Waste Gas Decay Tank B (WGDT-B) resulted in disagreement for the three identified radionuclides -- Xe-133, Xe-133m, and Xe-135. The licensee's results ranged from 33 to 50 percent above the NRC's results. The licensee and the NRC use different sample containers for analyses of gas sample The difficulty of obtaining identical samples in the dif-ferent sample containers may account for some of the discrepancy in the results. When a single sample in the licensee 6 cc vial was analyzed by the licensee and by the NRC, closer agreement was obtained, although the licensee's results were still biased somewhat high. The inspector reviewed the licensee's calibration method and data and found that a solid source is used for the calibration of the 6 cc vial geometry. The licensee stated that no correc-tions were made to take into account the attenuation of the low energy photons within the solid source. The inspector stated that the licensee's efficiency file in this geometry would be inaccurate for low energy photons due to the attenuation. The calculated efficiencies at low energies will be lower than if the calibration were performed using a gas calibration source. When an actual gas sample is analyzed, the calculated quantity of low energy radio-nuclides will be overestimated to the extent that effici-encies were underestimated. The inspector further stated that the licensee should attempt to obtain a gas calibra-tion source for this geometry; or, if this is not possible, they should consider utilizing a 33 cc glass bulb for gas analyses. Actual gas calibration sources are available and are widely utilized in the industry. The licensee stated that it would investigate this matter. Because the licen-see's results are conservative; i.e., the quantity of radioactivity in gaseous effluents is overestimated, no violation of regulations occurred.

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A disagreement also resulted when the licensee analyzed a uniformly- loaded simulated charcoal cartridge supplied by the NRC. The analysis of a similar, but face-loaded, cartridge resulted in agreement. At the time of analysis, the licensee was not given any information to indicate the distribution of radioactivity in either cartridg The licensee's initial results for the uniformly-loaded cart-ridge were 18 to 29 percent below the known values. This analysis was performed by analyzing the cartridge as though it were face loaded. When a re-analysis was performed by counting each face and averaging the results, closer agree-ment was obtained, although the results were still biased low. The licensee stated that it does not normally encounter uniformly-loaded cartridges, due to the low levels of activity in its effluents, and has no calibration file for a uniformly-loaded cartridge. The inspector dis-cussed with the licensee the importance of being able to accurately analyze a uniformly-loaded cartridge. Such a need may arise if effluents were to increase well above normal levels for any reason. The inspector stated that the licensee should develop, as a part of its analytical procedures, a method to enable the analyst to quickly determine whether a cartridge is uniformly or face-loaded in order to perform the proper analysi The licensee's ability to accurately analyze gas and char-coal cartridge samples will be reviewed in a future inspection (289/86-19-05).

8.1.3 Laboratory QA/QC The inspector performed a selected review of the licensee's program for the quality assurance of radioanalytical measurements. Included in this review were procedures and data pertaining to analyses performed within the licensee's chemistry laboratory and analyses performed by its vendor laboratory. The following procedures were reviewed:

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N-1828, Revision 0, dated August 1985, " Quality Assurance Program for Radiological Effluent Monitoring;"

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N-1990.1, Revision 6, dated May 1986, "High Resolution Gamma- Ray Spectroscopy Using Canberra Industries Jupiter System"; and,

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N-1990.2, Revision 3, dated May 1986, " Calibration of the Canberra Industries Jupiter System."

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The review of data included:

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efficiency calibrations performed during 1986;

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inter-laboratory and intra-laboratory comparisons performed during 1986;

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results of vendor analyses of QC samples; and,

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control charts for daily source checks and background checks for 1986 of the gamma spectroscopy system in the chemistry counting roo Calibration records for all detectors and geometries were available. Those selected for review appeared to be accurate and complete. The licensee normally calibrates its gamma spectrometry equipment once per year, but may do so more frequently if problems, such as instrument drift or failure, are experience Procedure N-1828 requires several types of inter-laboratory, intra-laboratory, and vendor QC checks. Included among these are requirements for the vendor laboratory and the licensee's technicians to perform analyses (including gamma-spectroscopy) on liquid samples from a waste evapo-rator condensate storage tank (WECST). The inspector noted that the WECST samples have typically contained little or no gamma-emitting radioactivity; therefore, no meaningful comparisons are possible using such a sample. The inspector discussed with the licensee the possibility of obtaining samples with measurable levels (>LLD) of gamma activity and, further stated that if the licensee cotid not obtain actual samples from the plant, consideration should be given to preparation of spiked samples for analysis as unknowns by the vendor laboratory or technicia Procedure N-1828 also requires that Quality Assurance (QA)

data be entered on Form N-1828-1. Section 3.7 of the pro-cedure covers spiked sample anslyses by site technicians and by the vendor laboratory. The inspector noted that while these analyses have been performed, as required, the results were not entered on Form N-1828-1. The licensee stated that these results are monitored by a different individual from the one who monitors the rest of the requirements in this procedure. The licensee further stated that this procedure was undergoing revision and that one option was to remove the requirement to record the data from Section 3.7 on Form N-1828-1. The inspector stated that the licensee's method for recording and reviewing this QC data would be reviewed in a future inspection

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(289/86-19-06).

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8.2 Technical Specifications Surveillances 8. Effluent and Process Monitor Surveillance The inspector reviewed the licensee's program for effluent and process monitor surveillances required by Technical Specifications. This encompassed a walkdown of effluent and process monitor locations in the plant, checking system operability, local and remote setpoints, flow rates, and strip charts. The inspector also reviewed the following procedures:

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Surveillance Procedure (SP) 1303-4.10, "RML-12 Inter-lock Test;"

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SP 1303-4.158, "RMS Monthly Test, Liquid Channels;"

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SP 1303-4.15A, "RMS Monthly Test, Atmosphere Channels;"

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SP 1302-17.1, "RML-12 Calibration;"

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SP 1302-3.1, "RMS Calibration;"

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Maintenance Procedure (MP) 1407-3, " Assessment of the Adequacy of the Preventive Maintenance Program;" and,

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Operations Procedure (0P) 1101-2.1, "RMS Setpoints" for adequacy and conformance to Technical Specification requirement The inspector selected a random sample of records of the surveillances required and source and channel checks for review. These were examined for timeliness, completeness, and results. The inspector noted that when "as-found" conditions required actions on the part of the licensee, reviews of safety significance and restorative actions were made promptl The inspector discussed the frequency of failures noted in surveillances. The inspector determined that the frequency of failures was not excessive but queried the licensee on what systematic evaluations were made. The maintenance scheduling coordinator (GMS) reviews these occurrences for frequency of failures of components in common within the various systems and trends these. When a trigger level of failures is achieved an evaluation is made of the affected components for systems in which they are used and the use of alternative parts is reviewed, i

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The radiation monitor setpoint procedure and methodology was reviewed and the setpoints determined by the methods described were found to be adequate. Depending upon system operability status, setpoints determined during system walkdowns were found to be consistent with procedural requirement The inspector had no further questions in this are No violations or deviations were foun . Effluent Release Records The inspector reviewed the effluent release records by reviewing a selection of the appropriate procedures:

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Radiological Control (RC) Procedure 1621, " Releasing Radioactive Liquid Wastes;"

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RC 1622, " Releasing Radioactive Gaseous Wastes;" and,

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RC 1676, " Radiological Controls Responsibilities for Non-Routine Release."

Also, the inspector selected a random sample of routine and non-routine release records for scrutiny and observed a radiological engineering staff member as he entered data from a recent non-routine release into a computer, verify-ing hand calculated doses and released curies for various reporting periods. This included a spot check on the cal-culating method for consistency with the Off-site Dose Calculation Manual (0DCM) for both this particular release and the selected samples. All of the release permits reviewed conformed to procedural requirements, calculations were made properly for tank recirculation duration and analysis results, as well as related parameters to imple-ment and comply with applicable procedures and regulatory requirement In addition, the inspector discussed the flow path of release requests by operations, chemistry sampling and analyses and radiological evaluations pertaining to routine releases and found the systematic methods for releasing radioactive effluents comprehensive and timely. Intra-departmental communication of information relevant to ideas were similarly effected. The inspector had no further question in this area. No violations or deviations were found.

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8. Sampling Program The inspector reviewed lii + see surveillance records and discussed grab and continuous sampling procedures to deter-mine if effluent sampling and analysis was being conducted as required by Technical Specification Tables 4.22-1 and 4.22-2. The inspector noted that all surveillance require-ments were scheduled by a program coordinator and results were reviewed by a chemistry foreman and a program coordin-ator prior to filing in the control roo Procedures reviewed included:

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N-1857, Revision 1, " Liquid Waste Disposal System Sampling;"

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SP 1301-6.3, Revision 18, " Waste Evaporator Condensate Storage Tank Sampling;"

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9100-ADM-4250.09, Revision 6 " Tracking Continuous Releases from Turbine Building Sump;"

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SP 1301-4.9, Revision 11, " Liquid Continuous Release Composite, Weekly, Monthly, and Quarterly;"

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N-1858, Revision 1, " Sampling and Tracking Continuous Liquid Releases from Turbine Building Sump;"

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SP 1301-5.9, " Condenser Vacuum Pump Release Sampling;"

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N-1853, Revision 1, " Sampling of Waste Gas Decay Tanks;"

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9100-IMP-4200.03, " Revision 4, "Radcon/ Chemistry Actions Required When RMS Malfunctions;" and,

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RC 1622, Revision 35, " Releasing Radioactive Gaseous Waste."

These procedures addressed all required sampling surveillances. The inspector determined that all analyses were completed for the period January 1986 through October 198 The inspector noted that the licensee's gross beta measure-ments did not meet the required Lower Limit of Detection (LLD). The gross beta sample preparation method, which was performed by Unit 2 staff and procedures, had not been reviewed by Unit I chemistry personnel for this specific application. However, no violation was identified because the licensee was performing an isotopic beta analysis for the same period. The licensee agreed to evaluate their

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effluent control program to determine if any other sample preparation / analytical procedures were being performed by Unit 2 and had not been reviewed. This item will be reviewed during a subsequent inspection (289/86-19-07).

8. ODCM Implementation The inspector reviewed procedures which implemented the Off-Site Dose Calculation Manual (ODCM, 9100-PLN-4200-01, OP 1101-2.1 (Setpoints), RP1621 (Liquid Wastes), and RP 1622 (Gaseous Wastes) and found consistency within these and the appropriate NRC Regulatory Guide 1.109, "Calcula-tion of Annual Doses to Man from Routine Releases of Reactor Effluents n r the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I," Revision 1, dated October 1977. In addition, many instances where conservatism in setpoints are employed provide several magnitudes of safety in releasing significant quantities of radioactive materials from the plant. Similarly, the inspector reviewed recent semi-annual effluent release reports for the latter half of 1985 and first half of 1986 and determined that such measures resulted in actual release quantities and corre-sponding doses which were quite lo The inspcctor had no further questions in this area. No violations or deviations were foun .3 Radioactive Waste System Operation 8. Management Controls The inspector reviewed the management structure and depart-ment interfaces as it pertains to the liquid and gaseous radioactive waste processing program. The radwaste opera-tions manager is assisted by the senior radwaste engineer to coordinate and trend radwaste processing performance and to recommend primary water movement. The radwaste support group is part of the operations department, which operates the radwaste system The inspector discussed with the radwaste manager the licensee's commitment to maintain equipment availability, reduce sump inleakage and minimize radioactive waste. The inspector noted the monthly summary reports are distributed to other support groups, including chemistry and radio-logical engineering. These reports detail the monthly and cumulative accomplishments of the radwaste group, provide comparisons to annual goals, and summarizes radwaste release and dose commitment data. The inspector noted an improved communications network between the chemistry, radiological engineering and operations departments, since the previous inspection of this are 'o

8. Liquid and Gaseous Effluent Processing Systems The licensee's system for processing of reactor coolant and miscellaneous (sump) wastes and the licensee's gaseous waste hold-up system was reviewed with respect to Technical Specifications 4.22.1.3.2 and 4.22.2.3.2 and design require-ments documents in the TMI Unit 1 FSA The inspector discussed with the radwaste senior engineer and radwaste manager system operability performance and whether any major changes had been made within the last two year Within the scope of this review, the inspector determined that the waste systems, which had been operable to reduce effluent releases, were being maintained through a preven-tive maintenance program and that there were no changes to the radioactive waste processing systems with the last two years. The inspector also noted that for each evaporator run, a performance report detailing system cperation, chemistry specifications, and dose limits had been prepared for 198 Within the scope of this review, no violations were identifie .3.3 Air Treatment System Testing The inspector reviewed the licensee's air treatment system tests for the period January 1985 to date with regard to Technical Specifications 3.15.1, 2, 3 and 4.12.1, 2, The inspector reviewed the results of the high efficient particulate air (HEPA) filter and charcoal adsorber in-place filter tests, air flow distribution, and design air flow and differential pressur Laboratory test results for charcoal efficiency determinations were also reviewe The above tests were required by Technical Specifications for the following systems: emergency control room; reactor building purge; and, auxiliary and fuel handling exhaust air treatment system The inspector determined that all required and event triggered testing had been completed and corrective actions taken when filter conditions did not meet acceptance criteri The inspector also reviewed the installation of the ESF ventilation system for the fuel handling building and dis-cussed the functional and in place filter testing schedul The licensee had not received the effluent radiation detector for this system and some rework has been scheduled

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for_ Novembe Vendor testing has been scheduled for early December 1986. This area.will be reviewed during a sub-sequent inspection prior to fuel movemen .4 Audits The inspector discussed auditing activities with cognizant licensee personnel. The following GPUNC Audit Reports were reviewed:

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S-TMI-85-02, "00CM;"

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S-TMI-85-11, " Operations;"

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S-TMI-86-01, " Maintenance;"

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S-TMI-86-02, "SAI-4.15;"

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S-TMI-86-05, " Functional Audit of Safety Systems;" I

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S-TMI-86-14, " Operations;" and,

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S-TMI-86-17, " Chemistry."

It was determined that the licensee was using a schedule for audits and safety reviews of both on-site and off-site (vendor) activities to implement the requirements of Technical Specifications, Section 6.5.3, in a satisfactory manner. Owing to the fact that several organizations are used to implement radioactive effluent releases and execute the surveillances required for the entire radiation monitor system (RMS) and system operability checks, it was necessary to review selected portions of the respective audits. The inspector determined that audits were being performed in a timely manner and required frequencies were being met. During the discussions with the auditing staff, it was determined that they are knowledgeable of relevant subject matter and use effective audfting technique The inspector also discussed the surveillance audits with cognizant staff in the Operations Quality Assurance (0QA) group. These are field audits made during actual performance of procedures and these were generally found to be adequate for monitoring of activitie The inspector had no further questions in this area. No violations or deviations were foun .5 Semi-Annual Effluent Release Report Review The inspector reviewed copies of the Semi-Annual Effluent and Release Report for the last half of 1985 and the first half of 198 In general, these reports contain the information required by the licen-see's Technical Specifications (Section 6.9.5) and Regulatory Guide 1.21, consisting primarily of a summary of the quantities of radio-

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30 active liquid and gaseous effluents and solid waste produced by the unit. However, the inspector noted that certain information required in the report covering the last half of 1985 (submitted February 28, 1986) was not included in that report. These are listed in Technical Specification 6.9.5.2, which requires:

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an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year;

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an assessment of the radiation doses from radioactive liquid and gaseous efflue.1ts to individual due to their activities inside the site boundary during the report period;

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all assumptions used in making these assessments (i.e., specific activity, exposure time, and location); and,

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an assessment of radiation doses to the most likely exposed real individual from reactor releases and other nearby uranium fuel cycle sources (including doses from primary effluent pathways and direct radiation) for the previous twelve consecutive months to show conformance with 40 CFR 19 The review of the report submitted on February 28, 1986, indicated that the assessment of doses for the previous calendar year was provided on a quarter-by quarter basis and divided into doses due to liquid effluents, iodines /particulates, and noble gases. The year's totals were not provided, but the reader is referred to the annual Radiological Environmental Monitoring Program (REMP) report. The assumptions used in making these assessments were not provided in the semi-annual report. No assessment was provided of doses to indivi-duals from effluents due to their activities inside the site boundary, nor were the assumptions used for making such an assessment. The assessment to show conformance with 40 CFR 190 is given in the annual REMP report. The inspector stated that since the doses for the pre-vious calendar year, including the conformance to 40 CFR 190, are given in the REMP report, the lack of this information in the semi-annual rr. port would not be considered a violation. However, the inspector stated that the lack of an assessment of doses to indivi-duals inside the site boundary is an apparent violation of the licensee's Technical Specifications (289/86-19-08).

9. Technical Staff and Managers Training 9.1 General The purpose of this portion of the inspection was to assess the effectiveness of the licensee's Technical Staff and Manager (TS&M)

Training Program. As a representative sample, the inspector chose the training programs of plant engineers, site liaison and corporate engineers, and a non-licensed operating engineer. The effectiveness of implementation was astessed by reviewing the following activities:

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Implementing Safety Evaluation (SE) No. 113300-003, Revision 0, " Packing Reduction and Live Loading of Valves" SE prepared on FORM 5000-ADM-1291.01;

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Integrated Leak Rate Test (ILRT) Outage Planning Summary;

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Environmental qualification (EQ) of a dual element RTD;

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Control problems with the emergency feedwater (EFW) pump;

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Engineering review for motor replacement for decay heat (DH)

valves 4A & B, 5A & B;

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Technical design review (TDR) of the Instrument Air System;

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Revision to an Engineered Safeguard System Emergency Sequence and Power Transfer Test 1303-11.10;

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Response to a Change Modification (CM) Request No. 0635M for the Precoat Filter Sampling;

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Mini Modification (MM)-056 for the Integrated Control System (ICS) Pump Speed Control; e

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Design deficiency with reactor coolant inventory tracking system instrumentation;

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EFW nozzle replacement safety evaluation;

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Appendix F: outage planning work: and, i --

ICS power supply B&W Owners' Group revie .2 Details of the Review The effectiveness of the training program was determined by: review-ing the activities performed by the engineers at the job site, where practical; interviewing the engineers and managers as to whether or not they were adequately trained, qualified to perform the activities; and what continued training were they receiving to maintain or upgrade their proficiency. The inspector also reviewed the training program with the TS&M Training Coordinator, including the qualifications of the instructors, the lesson plans, student and instructor evaluations, the training records, and the status of the Institute of Nuclear Power Operations (INPO) accreditation effort. The results were discussed with supervision and training. No violations or deviations were identified. The observations were as noted belo _ _ - _ _ _ _ _ - - - . __ _ _ _ _ __ _ ___ , _ _ _ _ _ _ _ _

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9. Observation of Work Activities The inspector witnessed or walked down the activities identified in the preceding Section The activities were conducted by personnel knowledgeable in the require-ments and Technical Specifications. This was especially evident in the followin In implementing the SE No. 113300-003, the mechanical engineer was required to meet with the mechanics, discuss their concerns and suggestions for the method proposed by Oyster Creek for repacking valves with graphite packing. The communications and the rapport with the mechanics insured that the method would be objectively evaluated. The preparation of the SE included a 10CFR50.59 review and changes to the FSA The engineer was aware that the procedure 1000-ADM-1291.01, Revision 1, " Procedure for Nuclear Safety and Environmental Impact Review and Approval of Document" for preparing SE's was changed. The engineer did not know whether the present or revised procedure required the engineer to be notified that his changes to the FSAR were in fact incorporated into the FSA The Outage Planning Summary for ILRT was developed by the senior mechanical engineer and is used by other engineers to develop their summaries of major task The summary identified not only the activities involved but also the responsible personnel and codes including quality assurance (QA) and quality control (QC) inspec-tion personnel. Training tapes were also developed by the engineer for use by the Training Center to instruct engineers in the use of the planning summar The instrument and control (I&C) engineer was tasked to insure the replacement RTD was environmentally qualified (EQ). He reviewed the procurement request (PR), researched, and selected a vendor in accordance with Regulatory Guide (RG) 1.9 A Special Temporary Procedure (STP) 1-86-0005, "MS-PCS Testing on EF-P1 was developed by another I&C engineer for testing the controls and switches on the EFW pumps. The problem was that original proportional controller for returning the inlet pressure to the turbine set point to its normal position was not working. One test was to use a proportional plus integral type controller. The inspector discussed the 10 CFR 50.59 issue with the engineer in his prepara-tion of the SE. Since they were making temporary fixes and testing same, a response to the 10CFR50.59 was not necessary. Upon finding a suitable controller design, the safety issue would be addresse . _

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The engineering review of the decay heat (DH) valves included: a Verification Plan / Status Sheet V-1101-212-025, "LPSI/ Decay Heat Remosal System; operator stroke times and motor torque requirements; support information such as performance curves and sizing a motor for an operator per memo 86-3644 (3/7/86); and input data sheet an Torrey Pines latest calculation The liaison (electrical) engineer was conducting the review in accordance with procedure 5000-ADM-7311.02 (EP-009), Revision 2, " Design Verification" and Limitorque Corporation Procedure " Gate and Globe Valve Selection".

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The non-licensed operating engineer was preparing a technical design review 813, " Evaluation of PAT Findings Regarding 21R Back-Up (BU) Instrument Air (IA) System. The review paralleled a field change request (FCR) C044907, Removal of Normal and BU IA Lines to EF-V-30C/D initiated by a SE that deleted these lines. The FCR was being reviewed by a liaison (mechanical) enginee An electrical engineer was reviewing revision 16 to the Engineered Safeguard System (ESF) Energy Sequence and Power Test 1303-11.10. The test included simulating as ESF actuation and verify that all functions go to the ESF mode. Also simulated was an under voltage (UV)

condition, closing a diesel generator (DG) breaker and verifying loading sequence. The engineer stated that he as a " procedure owner" must review all changes to his procedure and must be present for the conduct of the test A mechanical engineer was preparing a response to change modification (CM) request 0635M, "Precoat Filter Sampling". The guidelines for the response attachment 1, " Typical CM Response Format". The guidelines were developed by the lead mechanical engineer. The preparation effort included a safety issue response regarding 10CFR50.59, and preparing drawing s

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A liaison (I&C) engineer was tasked by Technical Function Work Request (TFWR B 00945) to review mini mod MM-056, ICS FW Pump Speed Control. Using the MM procedure EMP-02 and SE procedure 1000-ADM-1291.01, the engineer developed the installation package for the field to do the work including installing fuses in the ICS cabine _ __ . . - _ . _ _ . __ _ _ _ _ , . _ . _ . _ _ - _ _ - . - -

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9. TMI Training Department The TS&M Training Coordinator provided the inspector with a copy of procedure 6250-PGD-2650.01, Revision 0, "TS&M Training Program Description. The procedure's effective date is July 3, 1986 and is the principal document in the TS&M Training Program. The program will received another INPO review in November, 1986 before a final INPO Board Review in December 1986. All prior concerns by INPO have been resolved and accreditation is expected in December 198 The inspector reviewed the Training Coordinator's respon-sibilities described in paragraph 5.3 of the procedure. The Coordinator was able to produce examples as requested by the inspector. The examples included: lesson plans, student and instructor evaluations, exams, training given by the Coordinator himself, TS&M Training Program Tracking Sheet, remedial training, and maintenance of training record The pilot training that is being given includes Basic Principles Simulator, Diagnostic Techniques including computer assist diagnostics (CAD), and Management Oversight and Risk Tree (MORT), as well as the basic plant specific engineering training recommended by INP The continuing training is divided in three parts. Part I will deal with changes / modifications, Part II will deal with Licensee Event Reports (LER's), and Part III will deal with Regulatory changes, e.g Appendix R. Also being con-sidered is the EQ training in the continuing training effor The training records of the engineers and their managers are being automated (computerized). The automated records, except in one case, were complete and current, and provided the information to support the activities witnessed by the inspector. Later the Coordinator was able to provide the hard copy of the incomplete automated recor The Coordinator produced evidence that course critiques and evaluations are reviewed for the pilot training. In one case the students requested that the course time be extended and future training will be extended. "Back on the job" evaluations are scheduled for November, 198 The Coordinator has five qualified instructors to support the TS&M program. He can also rely on Oyster Creek Training Department to provide additional support on as needed basis. Analysis is underway for possibly increasing the staf ._. , _ _ _. . . _. ____ _ _ _ _ _ _ _ _

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The engineer interviews confirmed well qualified and trained engineers. They were experienced individuals and

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.- they were knowledgeable in the areas of their cognizanc They feel that they had sufficient training to perform the jobs that they did. They confirmed that the licensee l management was supportive of formalized internal courses and outside courses- ~Many recognized the training aspect

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of their participating in B&W Owners' Group activities, 4 which was also fully supported by the licensee.

L Although not controlled by the training department, engi-neering services assures that personnel review procedure /

procedure changes at the corporate level on a periodic-

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basis. The list of personnel versus procedure is com-puterized to assure personnel remain updated on procedure

changes. There is also an indoctrination and training.

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program for entry level engineers, which takes two years

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to complete and includes at.least one year at the sit Other, more experienced, personnel are required to review key technical function procedures governing their employ-ment activities.

i In response to PAT I findings, the licensee conducted a i seminar on those findings which-focused attention on the 1 - adequacy of safety evaluations and ANSI-45.2.11 design control requirements. All personnel interviewed confirmed

that the training was educational and heightened their-

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- awareness of regulatory requirements governing their

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activitie The technical personnel and management formal

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training courses also include discussion on regulatory requirements and/or licensing basis documents. It was i' apparent that prior to this year the training program for engineers was weak in assuring a complete understanding of regulatory requirements in the design control are Also, in discussing-the results of the interviews with the Training Coordinator, the inspector noted a weakness of the engineers to fully understand and follow procedures, especially administrative procedures. Lack of adherence to procedures was also addressed in the SALP issued in July 1986. In most of the activities reviewed, the engineers were using a similar report or review package for guidance or guidance given to them by their lead engineer or manager. The training coordinator found the same problem during the review of the questionnaires returned by the engineers and the subsequent job / task analyses. The correc-tive actions included adding a new course " Fundamentals of Procedure Writing", to supplement the procedure usage cours .

9. QA/QC Interface With TMI Training Two QA audits were selected by the inspector for revie The audits were comprehensive and the responses were timely-and adequate. QA Audit No. S-TMI-85-12 " Plant Engineering" issued in March 1986, contains a section regarding train-ing. The auditors did not report any findings or concern Similarly, QA Audit No. S-TMI-85-10,"TMI-1 Project Engi-neering, Start Up and Test Engineering", contains a section regarding training. Again the QA auditors did not report any findings or concern The inspector reviewed the QA engineers training with a O'

Engineering Manager. The manager has received manageriai training and showed the inspector a 1986 catalog " Manage-ment Development Section" used for his continued trainin His QA engineers receive indoctrination training (required reading and tests), formal training in Procurement QA, EQ, Fire Protection, Start Up and Test Program (prior to Start UP). The QA engineers also participate in the TS&M Training Progra During the review of the ILRT Outage Planning Summary, the senior plant engineer described Quality Control's (QC) role in the ILRT effort. Other engineers were aware of the QA/QC role in the plant operations and engineerin . Technical / Safety Review of Licensee's Procedure Changes During the PAT II inspection, the inspector became aware of a new review and approval system for station procedures. This area was also reviewed and documented in NRC Inspection Report No. 50-289/86-17. As noted in that inspection report, the adequacy of this new review process is still under NRC revie The new program appears to be less restrictive than the previous review process and the potential exists for changes to procedures to occur that can have an adverse effect on plant safet On a sampling basis, the inspector reviewed Procedure Change Requests (PCR's) and Temporary Change Notices (TCN's) to station procedures that were processed during this period. This review was conducted to ensure that procedure changes being processed under the licensee's new program as described above did not have an adverse affect on safe plant operation Each sampled PCR or TCN was reviewed for the purpose, content of the change, and the technical merits of the change. In addition, the inspec-tor reviewed the changes to ensure that the changes would not cause another approved procedure to conflict with the proposed changes to other procedures. The inspector reviewed thirty-four generated PCR's and four TCN's.

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Within the review, the inspector did not find any changes that would have caused an adverse affect on safety. All PCR's reviewed were properly complete No inconsistencies were noted for any reviewed TCN's or PCR's processed on the new system. This area will be reviewed again after the management meeting with the licensee is conducted and unresolved issues are discusse . Licensee Action on Previous Inspection Findings 11.1 (Closed) Unresolved Item (289/85-25-07): NRC Region I to Complete Review of NSCC Activities. The inspector completed the review of Nuclear Safety and Compliance Committee (NSCC) activities by examining NSCC staff to NSCC interface activities and NSCC verbal and written reports to the GPUN Board of Directors. As a basis for the review the inspector used Restart Condition No. 1.t and the Commission Restart Order CLI 85-2, dated February 25, 1985. Addi-tional basis was licensee letter dated June 29, 1984 from P. Clark, GPUN, to H. Denton, NRR, the NSCC Charter (enclosed in the June 29, 1984, letter) and the NSCC Staff Guidelines, dated July 1, 198 The inspector focused on:

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NSCC and/or NSCC staff review on a broad base of regulatory areas;

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NSCC and NSCC staff maintaining an appropriate degree of inde-pendence in their activities, in dispositioning their findings, and in reviewing licensee management response to their findings;

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identify what types of safety or compliance issues are present-ed to the GPUN Board of Directors; and,

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evaluate the responsiveness of licensee management to NSCC findings and evaluate the effectiveness of the NSC The inspector interviewed NSCC staff members and met with the NSCC, along with the GPUN President, a member of the Board of Directors (B of D). The inspector also reviewed the following documents on NSCC activities:

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monthly meeting minutes and notes between NSCC staff and NSCC for March - September 1986;

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NSCC staff monthly reports to NSCC for April to September 1986;

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NSCC Report No. 4, dated April 5, 1986, for period October 1, 1985, to March 30, 1986;

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NSCC Report No. 5, dated October 15, 1986, for period April 1, 1986, to September 30, 1986;

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licensee internal. memorandum from P. Clark to R. Laney, dated October 8, 1986, response to NSCC Report No. 4 l --

licensee internal memorandum from R..Fasula to' Division Direc-

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tors, dated October 14, 1986, Commitments Made to NSCC;.

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Board of Directors meeting minutes excerpts on monthly NSCC reports -- July 1984 to August 1986;

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NSCC staff activities schedule, Revision 4, September 19, 1986

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for July - December 1986;

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TMI-R-86.001, dated January 24, 1986, " Makeup and Purification i- Pump. Shaft Seal Failures;"

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-TMI-R-86.002, dated May 1, 1986, " Evaluation of the TMI-1 Eddy L ' Current Outage;"

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.TMI-R-86.003, dated March- 13, 1986, Overview of Plant Engineering;"

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OC-R-86.004, dated February 11,1986, 'f0yster Creek Control of Welding;"

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TMI-R-86.004, dated February 12,1986, " Evaluation of Control

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Room Operator Training Program;"

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TMI-1R-86.005 (0C-R-86.005), dated January 1986, " Effective

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Maintenance Indicators; I

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TMI-R-86-006, dated April 16, 1986, " Review of Material Non-l Conformance Report and Quality Deficiency Report Programs;"

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TMI-R0-86.007, dated August 4, 1986, " Status of Training l Programs Being Prepared for INP0 Accreditation;"

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TMI-L-86.081, dated July 7,1986, " Replica Simulator Status;"

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TMI-R-85.017, dated October 28, 1985, " Evaluation of TMI-1 Audit Program;"

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TMI-L-86.082, dated July 22, 1986, "0yster Creek Control Room Deficiency Tag Situation;"

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TMI-L-86.028, dated March 18, 1986, "NSCC Staff Review of GPUN Response to NSCC Report No. 3;" and,

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various internal correspondence between NSCC staff and NSCC for

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The NSCC staff is a contractor composed of highly experienced professionals. They develop a schedule that review all the func-tional areas of NRC SALP process, except for safeguard / securit Their evaluations are thorough and their findings somewhat parallel those of other licensee review group or even NRC staff; such as, with respect to the procedure adherence issu The NSCC staff generates monthly status reports to the NSCC and periodic evaluations are sent to the NSCC and form the basis for monthly discussions with the NSCC. The NSCC staff also assist in the preparation of the verbal NSCC reports to the GPUN Board of Directors at their monthly meetings, along with the written semi-annual report For the most part, the significant issues identified by the NSCC staff are passed on to the B of D by the NSC These and other sub-tier issues are not formally tracked, but the staff uses a manual system of reviewing past reports within an area to identify areas warranting follow-up from past evaluations. On a sampling basis, the inspector determined that the manual system was effective in identification of recurrent problem No safety issue has needed to be identified by NSCC to the S of Appropriate compliance prc51 ems were identified. Generic examples of issues brought directly to the B of D were: procedure adherence problem; adequacy of event logging for self review and effective corrective action; adequacy of a program to identify shelf life of parts internal to components stored in the warehouses; and, need for better training in the operator licensing area in light of failures on internal examination Since the President of GPUN is a member of the B of D, he is respon-sive to the NSCC concerns identified at the monthly B of D meeting There is a system for GPUN management to formally respond to the semi-annual report of the NSCC. Further, GPUN Administrative Division recently issued an internal memorandum for all divisions to update the status of previous issues identified by the NSC Division status reports are being prepare Observation meetings are informal discussions between GPUN management and the NSCC/NSCC staff. These meetings are vehicles for a two-way flow of information. The GPUN representatives have the opportunity to update the NSCC on various special topics of interest at TMI-1 and Oyster Creek. The other purpose is for the NSCC/NSCC staff to make recommendations for improvement in performance. These meetings are occurring at 6-month intervals and they appear to be effective as a communication tool for these two organization Based on the review and applicable sections of the GPUN B of D meet-ing minutes, it appears that the NSCC is reasonably effective in implementing its charter and in meeting applicable regulatory

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requirements. These activities reflect their independence of GPUN management. There appears to be frank and open discussion between NSCC and B of 0 with issues appropriately being identified to the B of D. That gives a unique aspect to the licensee's B of D in that they independently get, first-hand, compliance or potential safety issues. The President of GPUN indicated to the inspector that he does not rely on the work of the NSCC and that he expects his own required and initiative review groups to similarly identify those issues. There apparently have been few surprises at B of D meetings on the NSCC reports. Overall, the GPUN President finds the NSCC beneficia During the interview with NSCC members, the NSCC expressed its concern on how it was complying with regulatory requirements, its charter, and licensee commitments. As a self-review initiative, the NSCC had its staff review their own activities to assure proper implementation. No violations or deviations were identifie Independently, the inspector concluded that the NSCC and its staff were properly implementing requirements, its charter, and licensee commitment .2 (Open) Unresolved (289/85-25-04): Region I to Review Adequacy of Hydrogen /0xygen Storage in Emergency Feedwater Area. The licensee responded to NRC staff questions on this item in which a summary of the issue and response was provided in paragraph 6. The response was accepted; and, as noted in paragraph 6, this area remained unresolved pending the NRC staff fire protection inspection to assure proper implementation of 10 CFR 50 Appendix .3 (0 pen) Inspector Follow Item (289/86-10-02): Significant Damage to the Diesel-Driven Fire Pump Building. The licensee issued a Plant Incident Report (PIR), dated July 18, 1986. Actions identified in the PIR were:

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maintenance / engineering to determine the measure to prevent recurrence of the failure to FS-V-27;

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engineering to determine the function of FS-V-18 (check valve that permits screenhouse water to feed FS-P3 sump area) and determine if this valve is necessary for system operation; and,

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installation of an open grating at the suction pit access instead of a solid manway to provide a vent path for the pi Review of maintenance records associated with FS-V-27 found that the valve was not within the licensee's periodic maintenance program; however, the licensee has decided to add the valve to the fire system surveillance program / maintenance progra In addition, the inspector raised a concern about inadvertent flooding of the auxiliary building and Borated Water Storage Tank (BWST) tunnel via

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a common cable conduit run between the tunnel and fire pump building. This specific action is still under engineering evaluatio Also, FS-V-18 was initially installed to ensure a leak of diesel fuel from the fire pump diesel could not be inadvertently spilled I into the Susquehanna River. Subsequent modifications to the building has eliminated the initial design intent and the flapper associated with the valve has been remove Engineering evaluations determined that an open grating at the suction pit access was acceptable. In addition, however, a solid manway must be available to be installed in the event of severe flooding by the rive With respect to inadvertent flooding via cable runs, the licensee is still in the process of evaluating the long-term solution. Walkdown inspection of the cable run by the licensee indicated no damage had occurred to the cables in the cable ru The inspector reviewed the licensee's actions associated work packages and found their actions acceptable. The inspector will review the licensee action on final resolution of flooding via cable run and FS-V-27 final surveillance to be conducted in a follow-up inspection. This item will remain ope .4 (Closed) Inspector Follow Item (289/84-10-03). Review of Changes to the Containment Integrated Leak Rate Procedure SP 1303-6.1. The inspector reviewed Revision 20 to the subject procedure to verify that certain changes had been made as noted in Inspection Report No. 84-10. The changes involved adding 10 CFR 50 Appendix J and The Volumetrics Test Manual as references. The inspector verified that these changes had been made as part of Revision 14. Also, a requirement to verify that " lay flat" tubing or other obstructions near open ended vent piping are removed. The final change involved adding a precaution to specify operational requirements for the reactor building industrial cooling syste Appropriate precaution steps were added. All requirement changes to SP 1303-6.1 were added. This item is close .5 (Closed) Unresolved Item (289/86-05-01). Degas Methodolog Licensee action on this item was to review the RCS degas procedures and flow paths to resolve the problem that occurred during degas operations for the SM outage. During this evolution, a release of noble gas occurred from the auxiliary building, due to the lift of a relief valve in the degas flow path. The licensee acknowledged that this was a somewhat difficult evolution to control and, subsequently, revised the controlling procedure to describe these problems and to allow for tighter control of the evolution. The licensee, subse-quently, decided that no major changes would be made to this flow path and that degas evolutions would be conducted in the future in accordance with present procedure .. _ _

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Additionally, the licensee has conducted degas evolutions in preparation for the upcoming 6R outage with no significant problem It appears that the additional procedure restraints and personnel awareness have alleviated this problem. The inspector had no other concerns in this area. This item is close .6 (Closed) Unresolved Item (289/86-02-01): Licensee Corrective Action on Personnel Contamination Due to Valve Maintenance Activities. Review of Radiological Investigative Report (RIR) N indicated that closecut corrective actions to preclude similar events in the future were tasked to two groups: Operations and Training. Operations has completed their corrective actions, which included clearly labelling valve position indicators for the involved valves. The Radiological Engineering section verified that the training section emphasized the need for proper taping of anti-contamination clothing hoods to respirators in radiation worker trainin .7 (Open) Inspector Follow Item (289/85-30-05): NRC staff will review adequacy of the records for on-the-job training. The licensee had revised their Standing Order No. 17 to require the evaluator to be more detciled when signing off a trainee on the basis of previous work experience. This requirement however was not implemented. The requirement will be part of the training programs for maintenance personnel that are being evaluated by INPO for accreditation in November and December, 1986. In the interim the Training Department will ensure that all on-the-job (0JT) is detailed for sign of .8 Conclusion The review of licensee's follow-up on outstanding items indicated the licensee was taking timely and proper corrective actions associated with each item. For the items that remain open, the licenses is pursuing resolutions of the issues. The inspector concluded that the licensee's technical approach to the different issues addressed above were adequate. In general, the licensee's system for resolving and completing outstanding items was found to be acceptabl a

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1 Exit Interview The inspectors discussed the inspection scope and findings with licensee management at a final exit interview conducted October 30, 1986. Interim exit interviews were also conducted on October 23, 1986, concerning engi-neer training; October 24, 1986, concerning confirmatory effluents measurements; and, October 31, 1986, concerning radwaste/ effluents programs. Senior licensee personnel attending the final exit meeting included the following:

J. Colitz, Manager, Plant Engineering, TMI-1 K. Harkless, Plant Chemistry, TMI-1 M. Press, QA Auditor, TMI -1 P. Sinegar, Plant Maintenance, TMI-1 C. Smythe, Manager, Licensing, TMI-1 H. Wilson, Supervisor, Preventive Maintenance, TMI-1 The inspection results, as discussed at the meeting, are summarized in the

! cover page of the inspection report. Licensee representatives indicated i

that none of the subjects discussed contained proprietary or safeguard information.

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Unresolved Items are matters about which information is required in order to ascertain whether they are acceptable items, violations, or deviation Unresolved item (s) discussed during the exit meeting are documented in paragraphs 5.3, 6, and 7.5.

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Inspector Follow Items are significant open issues warranting followup by the inspector at a later time to determine if it is acceptable, unresolved, a violation, or a deviatio Inspector follow items discussed during the exit meeting are addressed in paragraphs 8.1.2, 8.1.3, and 8.2.3.

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n ATTACHMENT 1 CRITERIA FOR COMPARING ANALYTICAL MEASUREMENTS This attachment provides criteria for comparing results of capability tests and verification measurements. The criteria are based on an empirical relationship which combines prior experience and the accuracy needs of this progra In these criteria, the judgement limits are variable in relation to the comparison of the NRC Reference Laboratory's value to its associated uncertainty. As that ratio, referred to in this program a> " Resolution,"

increases the acceptability of a licensee's measurement should be more selective. Conversely, poorer agreement must be considered acceptable as the resolution decrease Resolution = NRC REFERENCE VALUE RATIO = LICENSEE VALUE REFERENCE VALUE UNCERTAINTY NRC REFERENCE VALUE Resolution Agreement

<3 0.4 - .5 - .6 - 1,66 16 - 50 0.75 - 1.33 51 - 200 0.80 - 1.25 >200 0.85 - 1.18

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TABLE 1 TMI-1 VERIFICATION TEST RESULTS RESULTS IN MICR0 CURIES PER MILLILITER SAMPLE ISOTOPE NRC VALUE LICENSEE VALUE COMPARIS0N Simulated Cr-51 (4.4 1 0.4) E-5 (5.5 1 0.7) E-5 Agreement liquid waste storage tank Mn-54 (2.7 1 0.1) E-5 (2.75 1 0.14) E-5 Agreement 10/23/86 12:00:00 Co-58 (9.2 1 0.6) E-6 (1.0 1 0.1) E-5 Agreement Fe-59 (1.25 1 0.13) E-5 (1.17 1 0.17) E-5 Agreement Co-60 (3.4 1 0.1) E-5 (3.33 1 0.17) E-5 Agreement Zn-65 (2.72 1 0.14) E-5 (2.8 1 0.2) E-5 Agreement Cs-134 (3.81 1 0.10) E-5 (4.03 1 0.16) E-5 Agreement Cs-137 (1.87 1 0.08) E-5 (1.65 1 0.12) E-5 Agreement Ce-141 (7.3 1 0.6) E-6 (7.7 1 1.2) E-6 Agreement Ce-144 (5.3 1 0.3) E-5 (5.1 1 0.6) E-5 Agreement Reactor Xe-133 (8.39 1 0.05) E-1 (9.8 1 0.4) E-1 Agreement Coolant Letdown Xe-135 (2.89 1 0.04) E-1 (2.91 1 0.06) E-1 Agreement 10/21/86 00:40:00 I-131 (3.86 1 0.10) E-2 (3.74 1 0.11) E-2 Agreement I-133 (1.72 1 0.03) E-1 (1.72 1 0.04) E-1 Agreement I-135 (2.89 1 0.18) E-1 (2.74 1 0.14) E-1 Agreement Cs-137 (4.0 1 0.5) E-3 (2.8 1 0.4) E-3 Agreement RCLD Co-58 (5.65 1 0.12) E-4 (6.25 1 0.17) E-4 Agreement Crud Filter 10/22/86 Cs-137 (4.1 1 0.4) E-5 (3.5 1 0.7) E-5 Agreement 00:55:00 I-131 (7.2 1 0.5) E-5 (8.1 1 0.5) E-5 Agreement I-132 (5.5 1 0.5) E-4 (3.4 1 0.4) E-4 Agreement I-133 (2.94 1 0.09) E-4 (3.15 1 0.12) E-4 Agreement I-135 (4.3 1 0.4) E-4 (5.1 1 0.4) E-4 Agreement

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TABLE 1 (continued)

TMI-1 VERIFICATION TEST RESULTS RESULTS IN MICR0 CURIES PER CC SAMPLE ISOTOPE NRC VALUE LICENSEE VALUE COMPARISON WGDT-B Xe-133m (1.64 1 0.05) E-3 (2.4 1 0.2) E-3 Disagreement *

10/21/86 10:45:00 Xe-133 (1.889 1 0.002) E-1 (2.52 1 0.11) E-1 Disagreement *

Xe-135 (8.1 1 0.4) E-5 (1.12 1 0.17) E-4 Disagreement *

WGDT-B Xe-133m (8.6 1 2.5) E-4 (1.24 1 0.15) E-3 Agreement 10/23/86 11:10:00 Xe-133 (1.431 1 0.007) E-1 (1.77 1 0.08) E-1 Disagreement *

RESULTS IN TOTAL MICR0 CURIES NRC Cd-109 2.37 1 0.12 2.19 1 0.07 Agreement Simulated Charcoal Co-57 (4.04 1 0.20) E-2 (3.9 1 0.2) E-2 Agreement Cartridge ( Face- Ce-139 (7.88 1 0.39) E-2 (7.4 1 0.6) E-2 Agreement Loaded)

4/01/85 Sn-113 (1.72 1 0.09) E-1 (1.60 1 0.16) E-1 Agreement 12:00:00 Cs-137 (1.02 1 0.05) E-1 (9.2 1 0.2) E-2 Agreement Y-88 (2.66 1 0.13) E-1 (2.5 1 0.3) E-1 Agreement Co-60 (1.06 1 0.05) E-1 (1.02 1 0.03) E-1 Agreement NRC Cd-109 2.29 1 0.11 1.63 1 0.05 Disagreement *

Simulated Charcoal Co-57 (3.9 1 0.2) E-2 (2.80 1 0.17) E-2 Disagreement *

Cartridge (Uniformly Ce-139 (7.6 1 0.4) E-2 (5.5 1 0.5) E-2 Disagreement *

Loaded)

4/01/85 Sn-113 (1.66 1 0.08) E-1 (1.25 1 0.14) E-1 Disagreement *

1200 hours0.0139 days <br />0.333 hours <br />0.00198 weeks <br />4.566e-4 months <br /> Cs-137 (9.8 1 0.5) E-2 (6.98 1 0.18) E-2 Disagreement *

Co-60 (1.02 1 0.05) E-1 (7.8 1 0.2) E-2 Disagreement *

Y-88 (2.57 1 0.13) E-1 (2.1 1 0.2) E-1 Disagreement *

• See text, paragraph 8. o

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3 TABLE 1 (continued)

TMI-I VERIFICATION TEST RESULTS RESULTS IN TOTAL MICR0 CURIES SAMPLE ISOTOPE NRC VALUE LICENSEE VALUE COMPARISON NRC Cd-109 2.29 + 0.11

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1.69 + 0.08

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Agreement Simulated Charcoal Co-57 (3.9 1 0.2) E-2 (2.9 1 0.2) E-2 Agreement Cartridge (Uniformly Ce-139 (7.6 1 0.4) E-2 (6.0 1 0.7) E-2 Agreement Loaded)

4/01/85 Sn-113 (1.66 1 0.08) E-1 (1.4 1 0.2) E-1 Agreement 12:00:00 Re-analysis Cs-137 (9.8 1 0.5) E-2 (7.8 1 0.3) E-2 Agreement Co-60 (1.02 1 0.05) E-1 (8.5 1 0.3) E-2 Agreement Y-88 (2.57 1 0.13) E-1 (2.2 1 0.4) E-1 Agreement RESULTS IN MICR0 CURIES PER CC Reactor I-131 (6.4 1 0.3) E-10 (5 1 1) E-10 Agreement Building >

Charcoal I-133 (1,52 1 0.05) E-9 (1.5 1 0.2) E-9 Agreement Cartridge 10/22/86 I-135 (1.7 1 0.2) E-9 No Comparison 09:50:00 Analysis by HP

'..o l

GPU Nuclear Corporation' '2 cc w/encis:

R. J..Toole, Operations and Maintenance Director, TMI-1 C. W. Smyth, TMI-1 Licensing Manager

- R. J. McGoey, Manager, PWR Licensing E. L. Blake, J TMI-1 OTSG Hearing Service List Public Document Room (POR)

local Public Document Room (LPDR)

Nuclear Safety Information Center (NSIC)

NRC Resident Inspector Commonwealth of Pennsylvania