ML20140B705
| ML20140B705 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 03/21/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20140B704 | List: |
| References | |
| 50-289-97-01, 50-289-97-1, 50-320-97-01, 50-320-97-1, NUDOCS 9703280089 | |
| Download: ML20140B705 (26) | |
See also: IR 05000289/1997001
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U. S. NUCLEAR REGULATORY COMMISSION l
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REGION I !
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Docket Nos. 50-289 and 50-320
j. License Nos. DPR-50 and DPR-73
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} Report Nos. 97-01, Units 1 & 2
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, Licensee: GPU Nuclear Corporation
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Facility: Three Mile Island Station, Units 1 & 2 l
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l Location: P.O. Box 480
i Middletown, PA 17057
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! Dates: January 6,1997 - March 2,1997
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Inspectors: Michele G. Evans, Senior Resident inspector
Samuel L. Hansell, Resident inspector
Thomas J. Kenny, Senior Operations Engineer
Joseph L. Nick, Radiation Specialist
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j Approved by: Peter W. Eselgroth, Chief
i Reactor Projects Branch No. 7
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9703280089
970321
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ADOCK 05000289 .
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EXECUTIVE SUMMARY -
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Three Mile Island Nuclear Power Station
Report Nos. 50-289/97-01 and 50 320/97-01
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This integrated inspection included aspects of licensee operations, engineering,
maintenance, and plant support. The report covers an eight week period of resident
! inspection from January 6,1997 to March 2,1997; in addition, it includes the results of
! an engineering review of the quality classification checklist (QCL) program with Regional
j support and an announced inspection in the radiological controls area for units 1 and 2.
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Plant ODeratIOnS
- The operations crew pedormance related to the main turbine tests was a strength
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, as noted by the excellent command and control and operator response during the )
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surveillance tests. In addition the coordination between the operations crew, plant !
management and engineering was excellent (Section 01.2). I
e Management's decision to verify the emergency trip function of the combined I
intercept valve at a reduced power level to determine if the turbine overspeed
protection existed, displayed a conservative approach to safe plant operation
(Section 01.2). l
- Operations personnel did a very good job of developing and executing the plan to l
open valve WDL-V-307 to quantify reactor coolant system leakage. An on-line !
maintenance risk management document was written, and a thorough job pre-brief
was conducted. Operations and Radiological Controls personnel were ready to
support a D-ring entry to close the reactor coolant system drain header valves if
increased leakage was observed. However, the shift supervisor did not optimize the
use of the available operations staff to monitor and respond to a potential plant
problem (Section 01.3). l
- The licensed operator requalification training program performance was excellent.
The following areas were particularly strong: upper plant management involvement
in administering the operating exams, briefing simulator evaluators on the crew's i
performance in the simulator during the previous training cycle, challenging
simulator examinations, critical evaluators, excellent alarm response and exam
security (Section 05.1).
- Improvement was noted for the crew communicaticas and individual self checking
techniques during the simulator exams comparad to the previous year's
observations. The simulator computer enhancements reflected on plant
management's continued support related to improved operator training (Section
05.1 ).
Maintenance
- The surveillance test activities observed during this inspection were performed
satisfactorily and demonstrated that the associated systems could perform their
design safety functions. Mechanical maintenance performance was excellent
related to the major overhaul of the 'C' river water travelling screen (Section M1.1).
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Enaineerina
- Licensee engineering management and staff did an excellent job of evaluating the
potential significance of valve WDL-V-307 packing and body gasket leakage
(Section 01.3).
- Overall, the licensee's implementation of the component classification process was
poor. Four apparent violations of NRC requirements were identified. The inspectors
identified eight components related to the make-up, nuclear river water, and decay
river water systems, which were improperly downgraded from the nuclear safety
related classification to a lower tier classification without appropriate safety
evaluations or other supporting engineering documentation. The inspectors
identified se'Eral examples where licensee personnel had not followed engineering
procedure, EP-011. Also, EP-011 had not received the Technical Specification
required safety reviews. In addition, the licensee's previous actions to address
program problems identified by their own quality assurance activities were not
effective or timely (Sections E2.1.1, E2.1.2, E2.1.3).
- Upon identification, the regulatory affairs department promptly submitted the proper
10 CFR 70.24 exemption request for the operating license for a generic criticality
monitor issue. Plant procedures and worker knowledge provided the reassurance
that personnel would respond appropriately to a criticality monitor alarm. This issue
is being treated as a Non-Cited Violation, consistent with Section Vll.B.1 of the NRC
Enforcement Policy (Section E7.1).
Plant Support
- Overall, the licensee continued to maintain effective programs for radioactive
material and radioactive waste management and transportation. The radiological
waste management and transportation activities were in accordance with the
applicable UFSAR process descriptions. Radioactive waste handling and processing
were performed in accordance with procedures and regulatory guidance.
Radioactive waste and radioactive material shipment records were maintained well.
Program procedures appropriately incorporated the recent changes to the
Department of Transportation and NRC regulations, and training on the revised
regulations was very good. Audits and appraisals by the licensee's staff continued
to improve the quality of the radiological controls program. Facility tours indicated
good material condition and housekeeping in areas used for processing, staging, and
Doring radioactive materials and radioactive waste (Sections R1, R3, RS, R6 and
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- The hcensee's corrective actions regarding two previous violations involving access
controls to high radiation areas were comprehensive and effective (Section R8).
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i TABLE OF CONTENTS l
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EXECUTIVE SUMM ARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ............. ii ,
TA B LE O F C O NTENT S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv
1. Operations .................................................... 1
O1 Conduct of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
01.1 General Comments .................................. 1
01.2 Main Turbine and Generator Valve Testing . . . . . . . . . . . . . . . . . 1
01.3 Evaluation of WDL-V-307 Packing and Bonnet Gasket Leakage . . 2
05 Operator Training and Qualification . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 I
l 05.1 Licensed Operator Annual Examination Process Review ....... 3
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11. Maintenance ....................................... .......... 4
M1 Conduct of Maintenance .................................. 4 l
M1.1 General Comments ................................. 4 '
ill . Engine e ring . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
E2 Engineering Support of Facilities and Equipment . . . . . . . . . . . . . . . . . . 5
E.2.1 Quality Classification List and Component Downgrade
Program Review .............................. 5
E7.1 Criticality Monitor and Exemption Request . . . . . . . . . . . . . . . . . . . . . 12
IV. Plant Support ................................................ 13
R1 Radiologic al Controls . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13
R3 Radiation Protection Procedures and Documentation . . . . . . . . . . . . . . 13
R3.1 Solid Radioactive Waste Program ...................... 13
R3.2 Radioactive Waste / Radioactive Material Shipping Program .... 14
R3.3 Implementation of the Revised DOT Shipping Regulations . . . . . 15
R5 Staff Training and Qualification in Radiation Protection . . . . . . . . . . . . 16
R5.1 Training and Qualifications of Personnel ................. 16
R6 Radiation Protection Organization and Administration . . . . . . . . . . . . . 16
R6.1 Changes in the Radiological Controls Program ............. 16
R7 Quality Assurance in Radiation Protection Activities . . . . . . . . . . . . . . 17
R7.1 Audits and Appraisals .............................. 17
R8 Other Radiological Control Program items ..................... 18
R8.1 VIO 50-289/96-02-01, VIO 50-289/96-06-01 (Closed) ....... 18
R8.2 Verification of Updated Final Safety Analysis (UFSAR)
Com mitm e nt s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19
V. M a n a g e m e nt M e e ting s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19
X1 Exit Meeting Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19
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PARTI AL LIST OF PERSONS CONTACTED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20
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ITEMS OPENED, CLOSED, AND DISCUSSED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21
LIST O F ACRO NYM S U S ED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22
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Report Details
Summary of Plant Status
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l Unit 1 remained at 100% power throughout the inspection period, with the exception of a i
! two hour power reduction to 95% to perform turbine generator valve testing on February
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l. Operations
01 Conduct of Operations (71707,92901)'
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01.1 General Comments
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Using Inspection Procedure 71707, " Plant Operations," the inspectors conducted frequent '
i reviews of ongoing plant operations. In general, the conduct of operations was
! professional and safety-conscious; specific events and noteworthy observations are j
, detailed in the sections below. Plant management ensured that potential safety issues
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l OL.2_ Main Turbine and Generator Valve Testina
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l a. Insoection Scooe (71707) i
The inspectors observed the main turbine generator monthly surveillance and valve testing ;
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from the control room. The tests were performed using Operations Procedure 11061, '
- " Turbine Generator Monthly Test and Turbine Generator Valve Testing."
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- b. Observations and Findinas
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The tests were conducted by the control room operators (CROs) and observed locally by
the auxiliary operators (AOs). The CROs verified all actions prior to performing the
component manipulations. The OP was clear, concise and performed in sequence for both
tests. The shift senior reactor operators (SROs) provided excellent command and control
throughout the surveillance tests. The SROs coordinated the tests with other shift
activities and scheduled the surveillance at a time when the tests were the only plant
activity. This allowed the entire operations crew to focus on the important tests and safe
operation of the plant without other distractions.
During the turbine valve test, combined intercept valve (CIV) No. 4 did not close on the
partial close test signal. The failure was detected immediately by the CROs and reported
to plant management. The partial close stroke test is performed on a routine basis to
verify the turbine isolation valves would close on a turbine trip preventing a potential
overspeed condition. Management took a sound approach to " step back," evaluate, and
' Topical headings such as 01, M8, etc., are used in accordance with the NRC standardized
reactor inspection report outline. Individual reports are not expected to address all outline
topics.
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analyze the potential impact on plant operation and determine the probable cause(s) of the
failure. The review included excellent involvement and coordination between the
operations shift crew, engineering and plant management. An on-line risk evaluation was
performed in conjunction with a special test procedure to troubleshoot and resolve the CIV
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The special test was reviewed and approved by qualified reviewers. The test was
designed to verify if the CIV problem was with the test solenoid or the valve emergency
trip solenoid (the trip solenoid provides the turbine overspeed protection). Management
decided to reduce reactor power and turbine load to 95% of fullload as an added
precautionary measure. When the CROs inserted a close signal for the CIV, the system
engineer noticed that the valve test solenoid was sticking and was able to free the solenoid
during the test. This allowed ClV No. 4 to go full closed as planned. A subsequent partial
close test also worked as desigr.ed. Plant management intends to initially perform the
partial stroke test at an increased frequency to ensure the sticking solenoid problem was
corrected.
c. Conclusions
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The operations crew performance related to the main turbine tests was a strength as noted
by the excellent command and control and operator response during the surveillance tests.
In addition the coordination between the operations shift, plant management and
engineering was excellent.
Management's decision to verify the emergency trip function of the valve at a reduced
power level to determine if the turbine overspeed protection existed, displayed a l
conservative approach to safe plant operation.
01.3 Evaluation of WDL-V-307 Packing and Bonnet Gasket Leakage l
During a routine Reactor Building entry the week of January 13,1997, the licensee
identified that valve, WDL-V-307, had packing and bonnet gasket leakage. WDL-V-307 is 1
a normally closed, manually operated gate valve on a 4" drain line inside the containment. l
The drain sources include the reactor coolant pump suction lines, the once-through-steam-
generator (OTSG) bottom end bells, and the pressurizer surge line. Further licensee
analysis revealed that there was likely leakage past one or more reactor coolant system
(RCS) drain line valves, allowing the drain line upstream of the normally closed WDL-V-307
valve to pressurize.
The licensee's engineering staff evaluated this condition and took actions to quantify the
leakage past the RCS drain line valves. Since Technical Specification (TS) 3.1.6.2 limits
unidentified leakage to a maximum of one gpm, the licensee wanted to quantify the
leakage. Also, the licensee wanted to route this leakage to the reactor coolant (RC) drain
tank. Per TS 3.1.6.9, ioss of reactor coolant through system valves to connecting
systems, such as the RC drain tank, is not considered as leakage. These losses when
added to leakage shali not exceed 30 gpm. Therefore, the licensee wrote and approved
temporary change notice (TCN) No. 1-97-0002, to establish a pathway for RCS losses to
be returned to the RC drain tar k.
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On January 17,1997, the cperations staff implemented the TCN. WDL-V-307 was ;
opened and the RCS losses were routed to the RC drain tank. The licensee determined '
that the leakage past the RCS drain line valves was minimal, since there was no noticeable
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change in RCS leak rate upon opening of the valve. However, the licensee decided to
continue to operate for the remainder of the cycle with WDL-V-307 open allowing any
) losses past the drain lines to be captured in the RCS leak rate calculation and to avoid any
j further pressurization of the piping upstream of WDL-V-307. Engineering initiated Material ,
- Noncomformance Condition Report (MNCR) #97-001, to document and evaluate the ;
i original over-pressurization of the piping upstream of WDL-V-307. )
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l The inspectors observed activities associated with this issue including engineering's
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j evaluation of what the leakage being seen at the valve meant with regard to the technical
l specifications for unidentified RCS leakage and operations handling of the TCN l
l implementation on January 17,1997. The inspectors found that the engineers involved l
{ did an excellent job of evaluating the potential significance of the WDL-V-307 leakage. in j
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l addition, operations did a very good job of developing and executing the plan to open the
i valve to quantify the leakage. For example, an on-line maintenance risk management
i document was written, and a thorough job pre-brief was conducted. Operations and
i Radiological Controls personnel were ready to support a D-ring entry to close the RCS drain
i header valves if increased leakage wt s observed. However, the Shift Supervisor did not
] optimize the use of the available operations staff to monitor and respond to a potential
j plant problem. Specifically, several crew personnel (the Shift Engineer, the Shift Foreman,
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and two control room operators) were in the Shift Supervisor's office receiving Emergency
- Planning training at the time the evolution occurred and the Shift Supervisor old not deem
j it necessary to bring them back to the control room for the evolution. This issue has been ;
a discussed with Operations Management who acknowledged that this situation did not meet j
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management expectations and took prompt action to address the issue with the individual
, involved.
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05 Operator Training and Qualification i
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05.1 Licensed Ooerator Annual Examination Process Review j
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a. Inspection Scoce l
The inspectors observed TMI's annual licensed operator requalification examination
process. The review included the administration of operating exams, exam security, exam
administration by the evaluators, simulator response, and crew performance.
b. Observations and Findinas
The inspectors determined that TMI's administration of the annual operating exams was
excellent as evidenced by the following:
a. Involvement of plant management in the administration of both the job performance
measures (JPMs) and simulator scenarios.
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b. Briefing the evaluators, prior to the crew simulator examinations, about the crew's
previous operating test performance.
c. Simulator scenarios that challenged both the crew and individual proficiency,
d. Conducting thorough critiques related to each crew member's performance.
e. Excellent alarm response and procedure implementation.
f. Establishment of excellent exam security.
In addition the crew performance was very good. In particular, an improvement was noted
for the crew communications and individual self checking techniques compared to the
previous year's simulator observations. Also, the simulator computer upgrade was a good
enhancement to provide improved simulator training.
c. Conclusions
The licensed operator requalification training program performance was excellent. The
following areas were particularly strong: upper plant management involvement in
administering the operating exams, briefing simulator evaluators on the crew's performance
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in the simulator during the previous training cycle, challenging simulator examinations,
critical evaluators, excellent alarm response and exam security.
Improvement was noted for the crew communications and individual self checking
techniques during the simulator exams compared to the previous year's observations. The
simulator computer enhancements reflected on plant management's continued support
related to improved operator training.
11. Maintenance
M1 Conduct of Maintenance (62707,61726)
M 1.1 General Comments
a. Inspection Scope
The inspectors observed all or portions of the following maintenance and surveillance work
activities:
- Job Order No. 131870, " Troubleshoot and Repair Auxiliary Building
Ventilation Exhaust Fan AH-E-14C."
- Job Order No. 120362, " Replace Fan Motor for Control Rod Drive Group 7."
Inspect."
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e Job Order No. 121608, "'C' Travelling Water Screen Repairs."
e Surveillance Procedure 1301-1, " Shift and Daily Checks."
e Operations Procedure 1106-1, " Turbine Generator Monthly Test and Turbine
, Generator Valve Testing."
e Job Order No. 114470, " Perform Diagnostic Testing of GL 89-10 Valve for
MU-V-148, MU Pump Suction 'B' from BWST."
e Surveillance Procedure 1303-5.2, Emergency Loading Sequence and HPI
Logic Channel / Component Test."
b. Observations and Findinos
The surveillance test activities observed during this inspection were performed
satisfactorily and demonstrated that the associated systems could perform their design
safety functions. The details of the Operations Procedure 1106-1, " Turbine Generator
1 Monthly Test and Turbine Generator Valve Testing," are included in report section 01.2.
The inspectors noted excellent work performance related to Job Order No. 121608, "'C'
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Travelling Water Screen Repairs." The major overhaul of the 'C' travelling screen was
performed without problems by experienced mechanical maintenance personnel under
difficult working conditions.
Observations and findings related to Job Order 131870, " Troubleshoot and Repair Auxiliary
Building Ventilation Exhaust Fan for AH-E-14C," are discussed in Section E2.
c. Conclusions
The surveillance test activities observed during this inspection were performed
satisfactorily and demonstrated that the associated systems could perform their design
safety functions. Mechanical maintenance performance was excellent related to the major
overhaul of the 'C' river water travelling screen.
111. Enaineerina
E2 Engineering Support of Facilities and Equipment (37751,40500,92903)
E.2.1 Quality Classification List and Component Downarade Proaram Review
The inspectors reviewed the quality classification list (QCL) and component downgrade
program using NRC inspection procedures 37751, "Onsite Engineering," and 40500,
" Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing Problems."
The inspection focused on GPU's implementation of the quality classification and
component downgrade of safety related items and included a review of the corporate
nuclear safety assessment (NSA) audit findings for corporate engineering. The process
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was performed primarily at the corporate engineering office in Parsippanny, New Jersey,
and included the downgrade of approximately 700 components from nuclear safety related l
(NSR) to "Other" and approximately 6,000 components from regulatory required (RR) to
"Other." In addition, the inspectors reviewed TMI's conformance with regulatory
requirements, licensee commitments, and industry standards related to the downgrade and
classification processes.
E.2.1.1 Enaineerina Comoonent Downarade/ Classification Proaram imolementation
Auxiliary Building Ventilation Exhaust System
The inspectors' observation of the auxiliary building ventilation exhaust system (ABVS)
maintenance work, for fan AH-E-14C, resulted in the followup of the system QCL status.
The ABVS was considered "Other" on the fan job order, (JO) No. 131870. Since the
ABVS is a system listed in the plant Technical Specifications (TSs), updated final safety
j analysis report (UFSAR), and Regulatory Guide (RG) 1.29 for seismic considerations, the
NSR or RR classification appeared to be the more appropriate system status. The
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engineering procedure EP-011, " Methodology for Preparing the Quality Classification List,"
revision 4, provides the written criteria to determine the type of system / components that
would be classified as NSR or RR. The ABVS written documentation and safety function-
coincided with the EP-011 requirements of an NSR system.
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After reviewing the issue with the system engineer, it was noted that the four ABVS fans,
fan motors, four high efficiency particulate an (HEPA) filters, system flow transmitters,
3 dampers and other components were downgraded from RR to "Other" on December 22,
1994. The components were included in the engineering program that changed the QCL ,
classification of approximately 6,000 components at TMI from RR to "Other." The
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components were evaluated using the written requirements of EP-011. Exhibit 2 from EP-
011 provided a checklist to capture the equipment downgrade /re-classification. The ABVS
checklist did not reference all of the applicable sections of the UFSAR or operating
procedures. In particular, UFSAR sections 14.2.2.5, " Maximum Hypothetical Accident,"
and 14.2.2.6, " Waste Gas Tank Rupture," were not referenced on the checklist. The
omissions excluded two of the most important design functions of the ABVS to filter and
minimize the radioactive release to the environment. The radiation doses for the postulated
gas tank rupture at the exclusion boundary were projected to be approximately 6.13 Rem
thyroid two hour dose without ABVS filtration compared to approximately .62 Rem with
- .e ABVS filters in service.
The inspectors questioned why the QCL checklist item 2.F., " Prevention of Releases to the
Environment," was checked "NO" for the ABVS re-classification. The ABVS checklist did
not contain or reference any supporting engineering documentation or evaluation to
support the reason for the missing information or to explain why it was acceptable to
downgrade /re-classify the system components. Also, the ABVS design bases functions
containad in UFSAR section 9.8.3.1, matched the EP-011 criteria a., c.. and o. for a
"!uactional class 3" system. Engineering considers a functional cbss 3 system to be NSR.
The ABVS QCL checklist documented the system components as a " functional class 6"
which is considered non-safety related.
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- Engineering reviewed the ABVS documentation and classification and determined that the
existing written information should have resulted in an ABVS classification of NSR using
the EP-011 classification process. Because of the ABVS classification as "Other" and the
fact that the system was not maintained to the NSR standards, the plant review group
(PRG) convened a meeting on February 28,1997 to determine the ABVS status. The PRG I
determined that the ABVS was inoperable and the plant operators entered the ABVS TS i
limiting condition for operation (LCO) 3.15.3. The TS LCO requires TMI to prepare and
submit a special report to the NRC within 30 days outlining the actions taken to restore
operability and the plans and schedules for restoring the system to an operable status. A
material nonconformance report (MNCR) No.97-006 was written to address the proper
classification of the ABVS components based on the current documentation and the design
requirements of the system.
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In summary, the ABVS components were downgraded to the "other," non-safety related !
quality level. However, engineering determined that these components should have been I
safety related consistent with EP-011 requirements. The ABVS is a system described in ;
the UFSAR. With regard to nuclear safety related component quality, section 12.3.1.2 of
the UFSAR states, " Materials and parts utilized in the repair and maintenance of the
nuclear-related portions of the unit will be of the same quality as, or better than, the
original materials." The actions taken by the licensee resulted in quality level reductions
for components of the ABVS, which is described in the UFSAR, without performing and
documenting a safety evaluation. This is an apparent violation of 10 CFR 50.59, i
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Nuclear River Water Discharge Valve Motor Operator
Components for the nuclear river (NR) water pump discharge valves NR-V-1 A,18,1C
motor operators were downgraded on July 18,1994, from NSR to "other", non-safety
related. The valve motor operator is required to maintain NR system operability for
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postulated accident conditions as described in TS section 3.3.1.4 and UFSAR section
9.6.2.3. With regard to nuclear safety related component quality, section 12.3.1.2 of the
UFSAR states, " Materials and parts utilized in the repair and maintenance of the nuclear-
related portions of the unit will be of the same quality as, or better than, the original
materials." The actions taken by the licensee resulted in quality level reductions for
components of the NR system, which is described in the UFSAR, without performing and
documenting a safety evaluation. This is the second example of an apparent violation of
In response to the improper downgrade, the quality verification (OV) department initiated
an MNCR No.97-005. The MNCR was written to evaluate, resolve and document the
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actions taken to return the components back to the appropriate NSR classification. in
addition, engineering conducted a detailed review of all NR discharge valve maintenance
activities and replacement parts to verify that they were treated as NSR even though the
QCL was changed. The review concluded that all NR component activities were conducted
as NSR functions and remained operable.
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Decay River Water Strainer Motor
Components for the decay river (DR) water strainer motors DR-S-1 A and 1B were revised
on July 18,1994, from NSR to "other." The valve strainer is designed to automatically
operate to maintain DR operability as described in TS section 3.3.1.4 and UFSAR section -
- 9.6.2. With regard to nuclear safety related component quality, section 12.3.1.2 of the
UFSAR states, " Materials and parts utilized in the repair and maintenance of the nuclear-
I related portions of the unit will be of the same quality as, or better than, the original ;
materials." The actions taken by the licensee resulted in quality level reductions for
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components of the DR system, which is described in the UFSAR, without performing and
documenting a safety evaluation. This is the third example of an apparent violation of 10
CFR 50.59.
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In response to the improper downgrade, the QV department initiated MNCR No.97-004.
The MNCR was written to evaluate, resolve and document the actions taken to return the i
components back to the appropriate NSR classification. In addition, engineering conducted >
a detailed review of all DR strainer motor maintenance activities and replacement parts to
determine if any components were outside of the NSR scope. The review discovered that
a motor pinion gear was not an NSR part. The review of the pinion gear and the other
component parts concluded that the DR strainer motors remained operable.
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Makeup Valve MU-V-17 Downgrade
Three components for makeup (MU) valve MU-V-17, " Normal makeup to the Reactor
Coolant System Control Valve," were downgraded from NSR to "Other," non-safety {
related. The downgrades from NSR to RR/RR to Other were performed for the valve ;
operator on March 18,1993/ November 9,1994; valve positioner on March 18, l
1993/ November 9,1994; and valve regulator on September 1992/ February 1,1994. The ;
EP-011 checklist was used to document the changes and list the applicable references. !
The checklist contained errors similar to the ABVS checklist and did not provide a technical
written justification for the downgrade.
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The most significant error was the failure to reference and recognize the MU-V-17 design '
description contained in UFSAR section 6.1.3.1, " Design Evaluation for the Makeup and
Purification System." The UFSAR description noted that on a postulated failure of valve
MU-V-18 (in series with MU-V-17) to close, MU V-17 was needed to mitigate the
consequences of a small break loss of coolant accident (SBLOCA) in the MU system. A
UFSAR revision dated July 1991, described the postulated SBLOCA that could jeopardize
the proper flow to the reactor coolant system if MU-V 17 was not closed to ensure that
adequate core cooling would be maintained. In response to an issue raised by the NRC
design inspection team (Inspection Report 50-289/96-201), on January 9,1997, the
licensee modified the controls for positioning the MU system valves and the MU-V-18
control power supply to negate the MU-V-17 safety function. The MU system changes
have resulted in removing the MU-V-17 controls from the NSR classification. However,
prior to January 9,1997, MU-V-17 served an NSR function, but was not classified as
such. In response to the improper documentation, the classification issue will be included
with an existing quality deficiency report (ODR), No. 96-2021, to resolve the MU-V-17
issue. .
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- Components for makeup (MU) valve MU-V-17, " Normal makeup to the Reactor Coolant
- System Control Valve," were downgraded from NSR to "Other," non-safety related. The
MU-V-17 design description is contained in UFSAR section 6.1.3.1, " Design Evaluation for
, the Makeup and Purification System." The UFSAR description noted that on a postulated
failure of valve MU-V-18 (in series with MU-V-17) to close, MU-V-17 was needed to
j mitigate the consequences of a small break loss of coolant accident (SBLOCA) in the MU
- system. With regard to nuclear safety related component quality, section 12.3.1.2 of the
. UFSAR states, ." Materials and parts utilized in the repair and maintenance of the nuclear-
j related portions of the unit will be of the same quality as, or better than, the original
materials." The actions taken by the licensee resulted in quality level reductions for
j
components of MU-V-17, which is described in the UFSAR, without performing and
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documenting a safety evaluation. This is the fourth example of an apparent violation of 10
CFR 50.59.
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l E.2.1.2 Enaineerina Procedure for Preparina the Quality Classification List
The inspectors reviewed the content and implementation of engineering procedure EP-011,
, " Methodology for Preparing the Quality Classification List." The procedure was originated
! as a corporate Technical Functions procedure and applied to both TMl and Oyster Creek
j sites. EP-011 Exhibit 2, provided a checklist to document the equipment downgrade /re-
j classification. The checklist was a stand alone dot ment to evaluate and downgrade /re-
, classify the approximately 700 NSR to "Other" and approximately 6,000 RR to "Other"
components at TMI. The inspectors identified several examples of licensee engineering
personnel failure to follow EP-011 which are detailed below.
EP-011 section 2.2 states that the procedure only applies to component hardware NOT
activity changes as detailed in the operational QA plan. The corporate engineering
personnel failed to follow the procedure criteria by applying and using EP-011 guidance to
change " activities" associated with the NSR and RR downgrades /re-classifications. A
majority of the QA revisions were changes in " Activities" associated with the components
referenced on the checklists. Examples of the activities discontinued were the receipt
inspection of parts, procurement engineering reviews, testing, calibration, and maintenance
activities,
in addit:ca, the inspectors discussed the use of EP-011 with engineering management,
who indicated that the procedure had not been followed for all of the component
evaluations. For example, the functional classes identified in Exhibit 3, line 14 and 14a (of
Exhibit 2) which were taken from an ANSI Standard were not reflective of the criteria used
to evaluate the actual plant. Thus, for the ABVS fan motors, the licensee did not use the
written criteria from EP-011, but rather used another criteria to downgrade the fan motors.
Engineering management also indicated that the engineers were highly trained to know
when to deviate from the procedure in order to make the downgrade decisions.
The inspectors concluded that in several instances EP-011 was not followed as required by
TMl administrative procedure AP-1001G, " Procedure Utilization," during the performance
of the downgrades of structures, systems or components. AP-1001G requires that
personnel perform activities within the scope of the OA plan for components at TMI in
accordance with written approved procedures. The failure to follow procedure EP-011 is
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an apparent violation of 10 CFR 50, Appendix 'B', Criterion V, " Instructions, Procedures,
and Drawings."
In addition, the inspectors noted that procedure EP-011, revision 4, as well as several
other engineering administrative procedures did not receive the safety reviews described by ;
TS 6.5.1.12. GPU Nuclear Corporate Policy and Procedure Manual number 1000-ADM- ;
1291.01, " Safety Review Process," implements the requirements contained in TS Section 4
6.5.1, " Technical Review and Control." The inspectors, based on review of the TS and
procedure 1000-ADM-1291.01, and discussion with NRC staff in the Office of Nuclear
Reactor Regulation, questioned engineering management regarding the basis for the
engineering administrative procedures not being included in the safety review process. In ,
addition to EP-011, these included but where not limited to EP-006, " Calculations,"
revision 4; EP-009, " Design Verification," revision 7; and EP-031, Equipment
Environmental Qualification, revision 5. Licensee management stated that the next revision ,
of EP-011 would receive the appropriate reviews per the safety review process, in !
addition, they would evaluate the need to include other engineering administrative
procedures in the safety review process. Also, per Confirmatory Action Letter 1-97-008
dated March 4,1997, the licensee agreed to take immediate steps to assure that
appropriate procedural requirements are in place prior to proceeding with plant equipment
work associated with reclassification to assure that parts of the correct quality
classification are used.
The failure to include EP-011, revision 4 it. the safety review process is considered an
apparent violation of TS 6.5.1.12.
E.2.1.3 Corrective Actions involvino Previousiv identified OCL issues ,
During review of the Quality Classification List, the inspectors found that there were
numerous deficiencies identified by the licensee's oversight organizations related to the
OCL since 1991. These findings are documented in the following reports:
- Audit 0-COM-91-13, " Technical Support," conducted November 22,1991 to June
18,1993.
- Audit O-COM-95-09, " Technical Support and Design Control," conducted June 27, ;
1995 to December 29,1995. :
- QA Monitoring Report #9100178, " Accuracy of Repetitive Task Entries and QCL
Maintenance," conducted June 10,1991 to
June 28,1991.
- QA Monitoring Report #9100178A, " Accuracy of Repetitive Task Entries and OCL
Maintenance," conducted February 24,1992 to
March 12,1992.
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in audit O-COM-91-13, finding #1, the licensee identified concerns related to the lack of
sufficient detail on the Quality Classification List work sheet to support the item !
classification. The inspector reviewed the documentation associated with the closure of !
this audit finding. The audit finding was acknowledged by the Manager, OCL/ Engineering I
Configuration on June 1,1992. During the period June 15,1992 through October 30, '
1992, documentation indicated that the responses of the Manager OCL/ Engineering
Configuration were not acceptable to the Plant QA Audit Manager, however, ;
documentation to escalate the finding to a higher level of management in the engineering I
organization, although initiated on October 30,1992, was never formally completed and
processed. The audit finding was closed by the Plant QA Audit Manager on May 25,
1993, based on neveral meetings which were held between the Manager QCL/ Engineering
Configuration and the QA Audit Group and on a procedure revision of EP-011 and sampling
,
to verify implementation of corrective actions, in audit 0-COM-95-09, finding #12, dated
, December 21,1995, the licensee again identified issues related to the lack of sufficient
- detail on the Quality Classification List work sheet to support the item classification. On
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December 23,1996, the Manager, Configuration Maintenance, TMI (responsibility for the !
QCL program was transferred to the TMI site in August 1996), requested an extension of
the due date for corrective actions for this audit finding until June 30,1997, to allow
revision to EP-011.
In addition, an Independent Safety Review (ISR) group review of audit 0-COM-95-09 was
initiated on May 16,1996. The ISR Reviewer identified the concern that changing the
quality classification of components without documented or approved basis, independent
'
review, and without a written safety evaluation documenting the basis of the change may
be a violation of 10 CFR50.59. Although there was a request that the ISR reviewer's
concern be resolved within 60 days, the reviewer's concerns were not resolved as of
February 27,1997.
The inspectors concluded based on the review of these documents and the findings of the
independent inspection effort documented in section E.2.1.1 of this report, that the
licensee's previous corrective actions with relation to QCL deficiencies had not been
effective or timely. This is an apparent violation of 10 CFR 50, Appendix B, Criterion 16,
" Corrective Action." 1
E.2.1.4 Conclusions
Overall, the licensee's implementation of the component classification process was poor.
The inspectors identified eight components, related to the make-up, nuclear river water,
and decay river water systems, which were improperly downgraded from the nuclear
safety related classification to a lower tier classification without appropriate safety
evaluations or other supporting engineering documentation. The inspectors identified
several examples where licensee personnel had not followed engineering procedure, EP-
011. Also, EP-011 had not received the Technical Specification required safety reviews.
In addition, the licensee's previous actions with relation to program problems identified by
their own quality assurance activities were not effective or timely.
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Following discussions between NRC and licensee rnanagement, Confirmatory Action Letter
(CAL) 1-97-008, dated March 4,1997, was issued to document the licensee's immediate
and long term corrective actions with regard to the identified issues.
E7.1 Criticality Monitor and Exemotion Reauest
a. insoection Scone
The inspectors reviewed a generic issue related to the criticality monitor (CM) requiremen+s
of 10 CFR 70.24. The issue applied to all operating commercial power reactors and was
focused on the status of an operating license exemption from the requirements of the
regulation. The review included the design of the two criticality monitors, associated
procedures, physical location, alarm response and operating license exemption status.
b. Observations and Findinas l
The inspectors reviewed the criticality monitor issue at TMl to determine the applicability of
the generic concern. The key issue for each site, including TMl, was whether an
exemption from the 10 CFR 70.24 requirements was submitted for the plant operating
license. TMI did have an approved exemption for the initial construction special nuclear
materiallicense. However, the exemption was not updated to cover the current operating
license. To correct the problem, TMI submitted an exemption request to the NRC on
February 7,1997.
The inspectors also reviewed the hardware related aspects of the CM issue. TMl has two
criticality monitors as required by 10 CFR 70.24. One is located on the refuel bridge
platform in the reactor cavity area and the other in the spent fuel pool (SFP)/new fuel
storage vault area. Because of the location of the monitors they are very sensitive to
background radiation levels and are located at a position to provide the earliest warning of
a postulated inadvertent criticality. The CM on the refuel bridge platform in the Reactor
Building (RB), normally reads 100mR/hr on the control room radiation monitor panel due to
its physical location, the plant being at full power, and the age of the plant. The SFP
monitor reads approximately .5 mR/hr with the plant at full power. The RB CM location
results in an alarm setpoint that is higher than the required 5 to 20 mR/hr band as noted in
10 CFR 70.24(2). However, the RB monitor is installed at the most sensitive location to
allow early detection of a potential inadvertent criticality.
The inspectors also reviewed the associated abnormal procedure 1203-10, " Unanticipated
Criticality." The procedure contained clear detailed direction to address an inadvertent
criticality event. In particular, the immediate and follow-up actions provided concise
actions to resolve an inadvertent criticality. The control room and plant operators were
familiar with the cms, the procedure guidance, and the appropriate response to a CM
alarm, The monitors do have clear and audible alarms. See Section X1 for licensee
commitments relative to handling new fuel.
c. Conclusions
Once discovered, the regulatory affairs department promptly submitted the proper 10 CFR
70.24 exemption request for the operating license. The two cms were installed and are
maintained to alert plant personnel of a potential inadvertent criticality. Plant procedures
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and worker knowledge provided the reassurance that personnel would respond
appropriately to a CM alarm. This issue is being treated as a Non-Cited Violation,
consistent with Section Vll.B.1 of the NRC Enforcement Policy.
IV. Plant SuDDort
R1 Radiological Controls (71750)
The licensee's program for radioactive materials and radioactive waste management and
transportation was reviewed. Specific areas reviewed included: audits and appraisals of
the program; changes in personnel, procedures, and equipment; facility condition versus
the UFSAR: training and qualifications of personnel; the solid radioactive waste
management program; the radioactive materials and waste shipping program; and
implementation of the new Department of Transportation (DOT) regulations. The
inspection also included a review of other items such as the implementation of corrective
actions for two previously identified violations regarding access controls to high radiation.
R3 Radiation Protection Procedures and Documentation
R3.1 Solid Radioactive Waste Proaram
a. Insoection Scope (86750)
The inspectors reviewed the solid radioactive waste program through a review of licensee
procedures, interviews with licensee personnel, and review of licensee records.
b. Observations and Findinas
The licensee used a combination of direct isotopic sampling, scaling factors, and dose-to-
curie conversions to determine the isotopic and curie content of radioactive waste
containers. Waste streams were sampled and sent to an offsite laboratory on a periodic
basis to determine the radioisotopic content. Hard to measure radionuclides (beta and
alpha emitters) were related to the gamma emitting isotopes through scaling factors.
Subsequent samples were not always sent off site, but were counted in the onsite detector
for gamma isotopic analysis. Routine packages were also analyzed and dose-to-curie
conversions were developed based on the waste stream. The inspectors reviewed the use
of the scaling factors, dose-to-curie conversion factors, and laboratory sampling and
determined that waste characterization activities were appropriate and in accordance with
NRC and industry guidance.
The licensee used a customized software program to determine the radioactive waste
classification, fissile class, and shipping categories. The inspectors reviewed various
computer records and determined that the licensee had used the software appropriately.
The licensee maintained current copies of licenses for facilities that received radioactive
materials and radioactive waste from their facility. Certificates of Compliance were
maintained for high integrity containers (HICs).
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The procedures for transferring and packaging radioactive waste and radioactive materials
were good. The Process Control Program (PCP) was maintained current; however, the
licensee was not currently solidifying or encapsulating any liquid waste. Some dry wastes
were being sorted prior to shipment, but volume reduction was performed by a vendor.
The licensee was using a vendor-supplied system to dewater liquid waste (powdex resins)
before shipment to a disposal facility. Other used resins were shipped after removing
excess water for complete dewatering and eventual disposal by vendor f acilities.
Radioactive waste was stored on the site until the volume was large enough to
economically ship it for processing or disposal. The storage f acilities had some containers
filled with waste, but the areas were not used for long term storage.
c. Conclusions
The licensee maintained a very good program for packaging, storing, and handling
radioactive waste and radioactive materials, including the appropriate use of waste stream l
analysis, isotopic composition, waste classification, and radioactivity calculations.
Program procedures were good, including the Process Control Program. No violations of
NRC regulations or significant safety concerns were identified.
R3.2 Radioactive Waste / Radioactive Material Shionina Proaram
a. Inspection Scope (86750)
The inspectors reviewed the radioactive waste and radioactive material shipping program
through a review of licensee records, interviews with licensee personnel, and review of
licensee procedures.
b. Observations and Findinas
The licensee was using various vendors for processing of radioactive waste and had
current access to the burial site in South Carolina. Some containers were staged in
designated storage buildings or in cargo vans until they were completely filled or a
container was ready, then the waste was shipped to the vendor facility for processing
(volume reduction, incineration, or decontamination). The inspectors toured all areas used
to store radioactive waste and determined that the areas were in good physical condition.
One area, the Unit 2 diesel generator (non-safety related and non-functional) room, was
currently being used to store slightly radioactive oilin 55-gallon drums. The areas had
been designated as a temporary storage area and a safety evaluation was in the process of
being written to determine the allowable total radioactivity in the area so that a
catastrophic accident (i.e., fire) did not release radioactive material to the environment in
excess of regulatory guidelines. Temporary storage areas were allowed by the licensee's
procedure so that a formal safety analysis could be documented. Preliminary conclusion by
the licensee's staff indicated that the storage did not involve an unreviewed safety
question. The inspectors noted that the oil drums were not leaking, the temporary storage
time period was less than the procedural requirement, and the oil was stored in a safe
manner. Based on the above information, the inspectors concluded that the licensee was
in compliance with their procedures and NRC regulations.
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The inspectors reviewed completed shipping records and noted the records were complete
with the appropriate information, reviewed and certified by qualified individuals, and were l
maintained in good condition. Attention to detail was very good. No violations of
regulatory requirements were noted.
The inspectors also reviewed the emergency response information provided as guidance to
emergency responders in the case of a transportation accident during transit. The
information was appropriate and was easy to locate. When a shipment was in transit, the
licensee provided a copy of the emergency response information to the control room. The
control room operators were listed as the emergency response contact after normal
operating hours, and would contact the radiation protection staff if required for other
shipment information. Because there was no shipment in transit during the period of this
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inspection, the emergency contact number and information were not directly verified by l
the inspectors.
c. Conclusions
The licensee maintained storage areas for radioactive waste and radioactive materials in I
good condition. A safety evaluation was not yet complete for storage of radioactive waste
oilin the Unit 2 diesel room. Shipping records were maintained in good condition and
emergency response information was provided as required. No violations of NRC
regulations or significant safety concerns were identified.
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R3.3 implementation of the Revised DO'I Shionino Reaulations l
a. Insoection Scooe (Tl 2515/133)
The inspectors reviewed the training of pasonnel and implementation of the revised
regulations for radioactive materials and radioactive waste through a random selection of
training records and interviews with various licensee personnel,
b. Observations and Findinos
The licensee had provided training on the revised DOT regulations to appropriate personnel
involved with the radioactive waste management program (see Section R5.1 of this
Report). The training was very thorough and covered the new regulations in good detail.
Licensee procedures also incorporated the new requirements including the revised Ai /A,
(allowed activities) values, the low specific activity (LSA) and surf ace contaminated object
(SCO) definitions, and the classification of fissile materials. The licensee had not fully
implemented the use of SI units in all procedures and software; however, this regulation is
not a requirement until mid-year 1997,
c. Conclusions
The licensee appropriately implemented the revised DOT regulations for shipping
radioactive waste and radioactive materials. No violations of NRC regulations or safety
concerns were identified.
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R5 Staff Training and Qualification in Radiation Protection
,
R 5.1 Trainina and Qualifications of Personnel
a. Inspection Scoce (86750)
The inspectors reviewed the training and qualifications of personnel involved with the
radioactive waste management program through a random selection and review of training
records and interviews with various licensee personnel.
4
b. Observations and Findinas
Training records of attendance and copies of examinations showed that all selected
individuals completed the required training for hazardous material handling, processing, and
shipping. The individuals designated as certifiers of radioactive waste shipments
completed vendor training on the revised DOT and NRC radwaste packaging and
transportation regulations. The inspectors noted that the appropriate individuals had been
trained relative to the new DOT regulations, including the technicians, principle auditor, and
radiological engineers involved with the radwasta shipping program.
The qualification records for radwaste personnel were very well documented. All records
,
checked by the inspectors indicated that personnel who performed radwaste processing or
shipping activities maintained the appropriate qualifications and attended refresher training
as required. Only one radwaste technician had not attended the annual training due to a
recent injury, but he was not currently assigned to radwaste activities. Licensee
representatives stated that the technician would receive the training prior to resuming
,
radwaste assignments. A review of the course content and examinations for the courses
revealed that they were technically accurate and contained very relevant information
including lessons learned from problems at the site and other facilities.
c. Conclusions
The licensee appropriately trained staff members on the procedures and regulations
pertaining to radioactive waste handling, processing, packaging, and shipping. The staff
was highly qualified for the positions and tasks associated with the radioactive waste
.
management program. No violations of NRC regulations or significant safety issues were
identified.
R6 Radiation Protection Organization and Administration
R6.1 Chanaes in the Radioloaical Controls Proaram
a. Inspection Scope (86750)
Changes to the radioactive waste management program were reviewed by the inspectors
through interviews with licensee personnel.
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b. Observations and Findinas
The licensee implemented minor staffing changes in the radioactive waste management
program since the last inspection. Specifically, one radwaste supervisor position had been
eliminated when an individual left the department and the licensee did not fill the position.
In addition, several radwaste technicians were assigned to the systems removal project at
Unit 2.
c. Conclusions
The changes to the licensee's radioactive waste management program were minor
personnel changes. No violations of regulatory requirements or significant safety issues
were identified by the inspectors.
R7 Quality Assurance in Radiation Protection Activities
R7.1 Audits and Aooraisals
a. Insnection Scope (86750)
Audits, surveillance reports, internal assessments, and deviation reports of the radiological
controls program documented since the last NRC inspection were reviewed by the
inspectors.
b. Observations and Findinas
The last Nuclear Safety Assessment (NSA) audit of the radioactive waste management
program was conducted in 1995 and was reviewed during the last NRC inspection of this
area. The next audit of the program was scheduled to begin in February 1997 to meet the
licensee's commitment for a program audit every 2 years. Planned members of the audit
team included NSA staff members and a technical specialist.
The NSA staff had performed some surveillance of radioactive waste activities in the last
12 months. The activities were documented as monitoring reports and included
observance of radioactive waste shipments, storage facilities, and radioactive waste
system operation. The inspectors noted that the last monitoring report for radwaste
shipment activities was dated in February 1996. Although there were three surveillance
activities in January through February of 1996, there were none since that time. The
inspectors reviewed the licensee's procedure and noted that only shipments with a greater
than Type A description were required to be observed by NSA personnel. The inspectors
determined that the NSA staff was meeting the procedural requirement, but should
consider observing other shipments because the number of shipments with high
radioactivity were decreasing over time. The licensee's NSA representatives stated that
the decrease in the number of higher activity shipments was reviewed, but the previous
good performance by the radwaste shipping personnel was also used as a reason to reduce
the number of surveillance activities. The NSA staff had identified minor deficiencies in
past surveillance activities. The inspectors reviewed the surveillance reports and noted the
corrective actions to prevent recurrence of the deficiencies. The corrective actions
implemented by the licensee's staff were appropriate and timely.
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Reports of problems or deficiencies, written by the radiation protection staff during 1996
identified and documented minor deficiencies within the radioactive waste management
program. The inspectors reviewed the reports and noted that the licensee's staff
implemented timely and technically acceptable corrective actions to prevent recurrence.
c. Conclusions
The inspectors concluded that the licensee continued to improve the quality of the
radioactive waste management program through the self-identification and correction of
minor deficiencies. No violations of NRC regulations or safety concerns were identified.
R8 Other Radiological Control Program items
R8.1 VIO 50-289/96-02-01, VIO 50-289/96-06-01 (Closed)
a. Inspection Scope (92904)
The inspectors reviewed the corrective actions implemented as a result of two previously
identified violations regarding access controls for high radiation areas,
b. Observations and Findinas
The licensee had responded to the Notice of Violation (NOV), 50-289/96-02-01, in a letter
dated June 11,1996. The licensee's response provided background information regarding
the high radiation area barrier that was found out of position by a licensee radiation
protection technician at the entrance to the decant slurry pump room. The root causes, as
determined by the licensee, were the failure to use standards, policies, and administrative
controls; lack of attention to detail by the operations personnel working in the area; less
than adequate personnel training and understanding; and the human machine interface
arrangement / placement of the barrier. Short term corrective actions had been previously
reviewed by the inspectors and were found to be adequate to prevent recurrence of the
same problem. Many of the longer term corrective actions had not been implemented at
the time of the next NOV (96-06-01). These longer term corrective actions included
incorporating the lessons learned from the event in the operators' training classes,
scheduling personnel for self-checking, effective observation and coaching techniques
training, and management support and leadership to foster an environment that encourages
attention to detailin this area. The inspectors reviewed the implementation of the longer
term corrective actions and found that they were adequate to prevent similar events. This
item is closed.
The licensee had responded to NOV, 50-289/96-06-01, in a letter dated December 11,
1996. The response outlined the licensee's corrective actions to prevent recurrence of the
event, including; a full scale investigation to determine who had propped open the high
radiation area barrier, a walk-down of all high radiation barriers to evaluate other j
opportunities for improvement, a review of previous violations and corrective actions to '
ensure continuation, institution of a verification program of all high radiation area barriers
and postings on a shiftly basis, continued efforts to reduce the total number of high ,
radiation area in the plant through shielding or flushing, and various attempts to increase
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worker sensitivity and awareness. These occurrences were discussed in meetings with all
shift workers, and GPU management stressed accountability and other expectations
regarding high radiation area barriers.
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The inspector verified that the licensee had implemented the corrective actions described in '
the response letter to the NOV. The inspectors had observed a decrease in the number of
high radiation areas in various areas. As a result of these corrective actions, the licensee
has not had a recurrence in the period since the violation. This item is closed.
c. Conclusions
The licensee's corrective actions were appropriate and timely to prevent recurrence of
violations regarding access to high radiation areas. i
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R8.2 Verification of Updated Final Safety Analysis (UFSAR) Commitments !
A recent discovery of a licensee operating its facility in a manner contrary to the UFSAR
description highlighted the need for a special focused review that compares plant practices, 1
procedures and/or parameters to the UFSAR description. While performing the radiological
controls inspections discussed in this report, the inspectors reviewed the applicable
portions of the UFSAR that related to the areas inspected.
The inspector reviewed the processing (as appropriate) and storage of radioactive waste i
and material at the Three Mile Island Station relative to descriptions and commitments
provided in the UFSAR. There were no inconsistencies identified between the UFSAR and
current practices relative to processing or storage of radioactive waste. The areas
designated for storage of radioactive waste were listed, and one area could not be ,
identified (Solid Waste Storage Building). Licensee representatives concluded that the
building no longer existed, and could be removed from the list. The overall adequacy of
the program for onsite storage of radioactive material and updating of the UFSAR, in
accordance with the 10 CFR 50.71(e) was considered very good.
V. Manaaement Meetinas
X1 Exit Meeting Summary
At the conclusion of the reporting period, the resident inspector staff conducted an exit
meeting with TMI management on February 28,1997, summarizing Unit 1 inspection
activities and findings for this report period. On January 10,1997, a Regional inspector
conducted an exit meeting with licensee management summarizing unit 2 inspection
activities in the area of radiological controls. TMI staff comments concerning the issues in
this report were documented in the applicable report section. No proprietary information
was identified as being included in the report.
With regard to the criticality monitor and exemption request discussions contained in
Section E7.1 of this report, the licensee committed in a telephone conversation between
NRC Region I (P. Eselgroth) and GPUN (M. Ross), that new fuel would not be moved in a
manner contrary to existing criticality monitoring requirements until any associated
exemption requests are approved by the NRC.
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PARTIAL LIST OF PERSONS CONTACTEC
Licensee
J. Curry, Acting Director, Nuclear Safety Assessment
D. Etheridge, Acting Radiological Controls / Occupational Safety Director
R. Goodrich, Site Security Manager
D. Hosking, NSA Manager
C.'incorvati, QV Manager
P. Karish, Procurement Engineering
R. Keaton, Vice President GPUN Engineering
J. Marsden, Procurement Engineering
L. Noll, Plant Operations Director
"M. Ross, Director, Operations and Maintenance, TMI
J. Schork, Regulatory Affairs
J. Wetmore, Manager, Regulatory Affairs
- senior licensee site manager present at exit meeting on February 28,1997.
NRC
B. Buckley, TMI Project Manager, NRR
E. Kelly, Chief, Systems Engineering Branch, Region 1
J. Wiggins, Director, Division of Reactor Safety, Region i
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INSPECTION PROCEDURES USED
IP 37551: Onsite Engineering
IP 40500: Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing !
Problems
IP 62707: Maintenance Observation
- IP 61726
- Maintenance Surveillance Observation
j IP 71707: Plant Operations
IP 71750: Plant Support Activities
IP 86750: Solid Radioactive Waste Management and Transportation of Radioactive
j Materials
- IP 92700
- Onsite Follow-up of Written Reports of Non-routine Events at Power Reactor
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Facilities
IP 92904: Followup-Plant Support 1
- Tl 2515/133: Implementation of Revised 49 CFR Parts 100-179 and 10 CFR Part 71
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4
, ITEMS OPENED, CLOSED, AND DISCUSSED
4 Opened l
None I
Closed
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50-289/96-02-01 " Decant slurry pump room high radiation barrier." (VIO)
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50-289/96-06-01 " Auxiliary building floor preservation work and moved high
j radiation barrier posting." (VIO)
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4
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4
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LIST OF ACRONYMS USED
ABVS Auxiliary Building Ventilation Exhaust System
AO Auxiliary Operator
BWST Borated Water Storage Tank
CAL Confirmatory Action Letter
ClV Combined Intercept Valve
CM Criticality Monitor
CRO Control Room Operator
CFR Code of Federal Regulations
DOT Department of Transportation
DR Decay River
HICs High Integrity Containers
ISR Independent Safety Review
JO Job Order
LCO Limiting Condition of Operation
MNCR Material Nonconformance Report
MOV Motor Operated Valve
MU Make-Up
NCV Non-Cited Violation
NR Nuclear River
NRC Nuclear Regulatory Commission
NSA Nuclear Safety Assessment
NSR Nuclear Safety Related
OTSG Once-Through-Steam-Generator
PDR Public Document Room I
PRG Plant Review Group
QCL Quality Classification List
ODR Quality Deficiency Report
QV Quality Verification
RG Regulatory Guida ....
RR Regulatory Required
RP Radiation Protection
SBLOCA Small Break Loss of Coolant Accis'nt
SRO Senior Reactor Operator
SS Shift Supervisor
TFAAl Technical Functio is Assigned Action Item
TS Technical Specifi::ation
UFSAR Updated Final S;!ety Ana'ysis Report
USO Unreviewed SafMy Quer, tion
VIO Violation
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