IR 05000289/1990015
| ML20058A617 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 10/18/1990 |
| From: | Ruland W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20058A612 | List: |
| References | |
| 50-289-90-15, 50-320-90-08, 50-320-90-8, NUDOCS 9010290036 | |
| Download: ML20058A617 (22) | |
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U. S. NUCLEAR REGULATORY COMMISSION I RTGTON I l ! Docket / Report Nos. 50-289/90-15 License Nos.: DPR-50 - 50-320/90-08.
DPR-73 Licensee: GPU Nuclear Corporation P. O. Box 480 . Middletown, Pennsylvania 17057 Facility: Three Mile Island Nuclear Station, Units-1 and 2
Location: Middletown, Pennsylvania i Dates: August 7, 1990 September 21, 1990 i - Inspectors: F. Young, Senior Resident Inspectot D. Johnson, Resident Inspector i D. Beaulieu, Resident Inspector ! i Other Personnel: R. Hernan, Project Manager j > /o//M/90 Approved by: , .. /, . ~ W. Ruland, Chief Tate' ~' ~
Reactor Projects Section No. 48 . ! Inspection Summary: Combined Inspection Report Nos. 50-289/90-15 and ' 50-320/90-08 for August 7, 1990 - September 21,-1990 Areas Inspected: The NRC staff conducted routine and reactive safety i inspections of Unit 1 power operations and Unit 2 cleanup activities.
The inspectors reviewed plant operations, maintenance and' surveillance. radiological practices, security measures and engineering support- , activities as they related to, plant safety.. Licensee _ action on previous-inspection findings was also reviewed.
' ' Results: An overview of inspection-findings.are summarized in the executive summary of.this report.
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9010290036 901019 PDR ADOCK G5000289 Q PDC _' ..
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< . . . ? l l TABLE OF CONTENTS , Page 1.0 Summa ry o f Fa c i l i ty Ac t i v i t i e s...................
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- 1.1 Licensee Activities..................
NRC Activities....,,..................
1.2 Persons Contacted........................ I .... 1.3 . . 1-l 2.0 Plant Operations (71707, 71710, 93702)*........'......
. i 2.1 Operational Safety Verification................ .-2 ' 2.2 Engineered Safety Features System Walkdown........... 3 2.3 Followup of Events Occurring During the Inspection Period...........................
3.0 Radiological Control s (71707)....................., 3 3.1 Routine Radiological Controls.................
3.2 Receipt of Antimony Source at TMI-2... ..........
, 3.3 Elevated Levels of Tritium in TMI Groundwater.........
> . , 4.0 Maintenance and Surveillance (61726, 62703, 71707)...,.....
4.1 Routine Maintenance Observations.............-...
, 4.2 Specific Checklist to Ensure Operability of. Redundant Equipment 6
4.3 Performing Work with Handwritten Step-by-Step Instructions.. 7-4.4 Routine Surveillance Observations .......,.....,. 7-5.0 Security (71707)..........................
. . . 5.1 Routine Security Evaluations.............-....
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6.0 Emergency Prepa redness Exercise...... -..............
, 6.1 Annual Emergency Preparedness Exercise....... .?. .....
' , 7.0 Safety Assessment and Quality Verification.(40500, 71707, 90712, 92700)...................--.... t.....
. . , 7.1 Special 10 CF'R 50.59 Report Review,..,,....._....
! . . t 8.0 Followup of Previous Inspection Findings ~(71707,'92702,'92701) ' 8.1 (Closed) Unresolved Item (50-289/87-09-06)........,.-. 13 -
8.2 (Closed) Unresolved Item-(50-289/88-28-01).., ....... 13-
' 8,3 (Closed) Violation NC4 (50-289/87-08-01)...... :. ,,... 13 - . 8.4 (Closed) Unresolved Item (50-289/87-08-05)L...........,.14
8.5 (Closed) Unresolved Item (50-289/88-16-01).........
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- 8.6 (Closed) Unresolved Item '(50-289/89-80-04).,......... 15 ~
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2 8.7 (Closed) Unresolved Item (50-289/89-80-03)....-.,.......:.15 8.8 (Closed) Unres:1ved Item (50-289/89-24-02)-..........=... ;15 8.9 (Closed) SIMS Items (I.C.1.2. A and il.C.1.3. A) --.' .16- . ........ 8.10 ( Cl o sed) SIMS I tem (II. F.2. 4)...... ;.. ;. ".... '>..... 16 - .. 8.11 (Closed) SIMS Item (II.D.3.L4.3)... .L...-..-.-.... ;. -.26 _... :. . 8.12 (Closed) SIMS Item (II.E.1.2.1.b)...-, ..... s.
. -- . 17.L . -... 9.0 E x i t Mee ti n g ( 30703) ............. '.. i..... :... -.. >.. ' :.17 '. > . .
- Denotes inspection modules performed
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.. l ' Executive Summary { I.
PLANT OPERATIONS i Overall, Unit 1 plant operations were conducted in a safe manner. No ' noteworthy observations were made.
II.
RADIOLOGICAL CONTROLS ! Above normal levels of tritium (53,000 pCi/ liter) was measured in the j water sampled from the TMI-2 groundwater monitoring wells.
The source
of the tritium has not been found.
Potential sources of this water . include the two Process Water Storage Tanks and the TMI-2 Borated l Water Storage Tank which contain Accident Generated Water.
The.
licensee is still investigating.
The licensee received an improperly packaged antimony source at TMI-2 > that exceeded transportation radiation limits.. This concern has been . : referred to the NRC Region III office for further investigation.
! \\ Radiological controls at TMI-l continue to be conducted properly and r no noteworthy observations were made.
III. MAINTENANCE AND SVRVEILLANCE ] Two procedures deficiencies were noted and a violation was' issued.
f regarding this matter.
The procedure for testing of the Reactor.- Protection System (RPS) failed.to provide adequate instructions to , i establish initial conditions to test 'an RPS function which could have l resulted in this function not-.being fully tested.- Also, a specific ! checklist was not performed on redundant safety. equipment' prior to maintenance as required.
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A technician was observed performing a maintenance procedure using { step-by-step handwritten instructions to supplement.the approveo
procedure. The issue is an unresolved item requirinf further review by the insepctor. IV.
ENGINEERING AND TECHNICAL SUPPORT Engineering support to plant activities was appropriate-to resolve l specific plant problems.
In general, good' engineering interface with the 'f plant staff continues to be noted, l
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EMERGENCY PREPAREDNESS (EP) ! . . -! The annual emergency preparedness exercise was conducted.. Inspection
results are documented in Inspection Report 50-289/90-80, i VI. SECVRITY ! ! Routine review of this area identified no noteworthy observations, j VII. SAFETY ASSESSMENT AND QUALITY-VERIFICATION _ .I ' A number of 10 CFR 50.59 evaluations were reviewed from the 1989 j annual report. Although the quality'of'10 CFR 50.59 ' evaluations is~. -! 1mproving, additional' improvement _is needed to' meet the intent'of this
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' . . , DETAILS t
1.0 Summary of Facility Activities i i 1.1 Licensee Activities The licensee began the inspection period operating at 97 percent ' power.
Reactor power slowly decreased over the inspection period
to 95 percent due to the gradual fouling of the secondary side of
the steam generators which reduced heat transfer.
1.2 NRC Staff Activities ' t This inspection assessed the adequacy of licensee activities
for reactor safety, safeguards and radiation protection.
The inspectors made this assessment by reviewing information on a sampling basis, through actual observation.of licensee activities, interviews with licensee personnel, or independent: calculation and selective review of applicable documents, inspections were acenmplished on both normal and back shift > hours.
NRC staff inspections were generally conducted in accordance with NRC Inspection Procedures (NIPS),. These NIPS are noted under the-
appropriate section in the Table of Contents to this report.
1.3 Persons Contacted l D. Atherholt, Operations Engineer-G. Broughton, Operations / Maintenance Director
- J. Byrne, Manager, TMI-2 Licensing G. Giangi, Manager, Corp. Emergency Preparedness
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- R. Harper, Manager, Plant Material
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- C, Hartman, Manager, Plant Engineering D. Hassler, Licensing Engineer
- H Hukill, Vice President and Director
- G. Kuehn, Site Operations Director,-TMI-2
- R. Knight, Licensing Engineer
- M. Nelson, Manager, Safety Review J. Paules, Senior Operations Engineer
- R. Rogan, Director, Licensing.and Nuclear Safety
"M. Ross, Plant Operations. Director.
- + E. Schrull, Licensing Engineer . - G. Simonetti, Manager Emergency Preparedness' R. Skillman, Director, Plant-Engineering P. Snyder, Manager, Plant _ Materiel Assessment.
! C. Smyth, Licensing Manager, TMI-1
' , J. Stacy, Manager, Security- .i R. Wells, Licensing Engineer
- K. Harkless, Nuclear Safety Manager
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- H. Wilson, ISS/ISI Coordinator
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- V. Orlandi, lead Instrumentation and. Control Engineer
- A. Paynter, Rad Con Field Ops Manager.. TMI-2
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- D. Ethridge, Rad Engineering Manager
- Denotes attendance at final exit meeting (see Section 9.0)
! 2.0 Plaat Operations j 2.1 Operational Safety Verification .' 'he inspectors observed plant operation-and verified that the p' ant was operated safely and in accordance with licensee i procedures and regulatory requirements. - Regular tours were conducted in the following plant areas: --Control Room --Control Building --Auxiliary Building --Diesel Generator Building --Switchgear Area --Yard Areas ' --Access Control Points --Containment Penetration ! --Protected Area Fence Line.
Area --Fuel Handling --Turbine' Building _ ' During the inspection, operators were interviewed concerning knowledge.of recent changes to procedures, facility configuration- - and plant conditions. The inspector verified adherence to.
I approved procedures for observed activities.
Shift turnovers-were witnessed and staffing requirements confirmed.
The .
inspectors found that ' control room access was properly controlled ! and a professional atmosphere was. maintained, inspector comments or questions resulting from these reviews were resolvs-d by ! licensee personnel.
i Contro! room instruments and plant computer' indications were [; observed for correlation between channels and for conformance i with technical specification (TS) requirements. Operability.of engineered safety features, other safety related systems and onsite and offsite power sources were' verified. The inspectors - ' . observed various alarm conditions and confirmed that. operator
response was in accordance with plant operating procedures..
Compliance with TS and. implementation of appropriate action ! statements for equipment-out of service was inspected.- Logs and records were reviewed to determine if entries were accurate and' , identified equipment status or deficiencies.
These. records 1 l included operating logs, turnover sheets,: system. safety tags, and the ' jumper and lif ted leads controlJ1og. The inspector also examined the condition of various fire protection, meteorological, and seismic monitoring. systems. ' ' > , ! , . .
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! I ' Plant housekeeping controls were monitored, including control and storage of flammable material and other potential safety hazards.
) ' 2.2 Engineered Safety Features System Walkdown 't The operability of selected engineered safety feature systems was
verified by performing detailed walkdownt of the accessible ! portions of the systems.
The inspectors confirmed that system ' components wsre in the required alignments, instrumentation was valved-in with appropriate calibration dates, as-built prints reflected the as-installed systems and the overall conditions-observed were satisfactory. The system inspected during this inspection period was the cmergency Feedwater System.
No concerns were identified.
3.
Radiological Controls 3.1 Routine Radiological Controls Posting and control of radiation and high radiation areas were inspected.
Radiation Work Permit compliance and use of personnel monitoring devices were checked.
Conditions of step-off pads, disposal of protective clothing, radiation control--job coverage, area . monitor operability and calibration (portable-and permanent) and..
personnel frisking were observed on a sampling basis.. There were no r noteworthy observations in this area.
3.2 Receipt of Antimony Source at TMI-2-On August 10, 1990, an antimony source was received at TMI-2 with a hot spot of 1800 nr/hr on contact-and a transportation index (mr/hr at 1 meter) of 22.
The limit for this mixed lading shipment was 2001 mr/hr on contact and a transportation.index of 10. -The shipping . company was Associated Couriers and the shipping invoice stated that the source cask had a Transportation Index of 18', The shipment originated from the University of Missouri.
The shipment was received by the TMI-2 staff using proper radiological. controls.
This concern has been transferred to the Region III Division'of Radiation
Safety and Safeguards for follow'up inspection.
3.3 Elevated Levels of Tritium in TM1 Groundwater , , On August 29, 1990 the licensee notified the resident inspectors that ! 29,000 pCi/ liter of tritium was-found in the August _3, 1990 sample-from ground water monitoring station MS-2.
Tritium was the only f isotope detected and MS-2 was the only one.of 10 normally sampled
stations that showed. elevated levels of tritium.
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The licensee indicated that tritium-levels at MS-2 have been. '
increasing over the last several months.
The-increasing trend. started in May,-1990 with a reading of:1200 pCi/ liter,.followed by- ! r < , %
. , .. , . - . . . ! 3800 pCi/ liter in June, 6700 pC1/ liter in July and 29,000 pC1/11ter on August 3, 1990.
The August 27 sample showed a declining trend with a concentration of 15,000 pCi/ liter._ This decrease was probably-due to increased rainfall during that week. The September 4, 1990 reading at MS-2 was back up to 29,000 pC1/ liter, the September 11, 1990 reading 46,000 pCi/ liter and the September 14, 1990 reading was 53,000 pC1/ liter.
For comparison, normal " background" levels at MS-2 for the past two years have been in the range of 500-1000 pC1/ liter, the 10 CFR 20 limit for discharges to unrestricted areas is 3 E -3-uC1/ml (3,000,000 pCi/ liter), the Environmental Protection Agency safe drinking water standard is 20,000 pC1/ liter and ambient tritium levels in rainwater are approximately 100-150 pC1/ liter.
In response to the current elevated tritium levels the licensee has initiated a sampling program to determine the source of the tritium.
The licensee has also increased the sample frequency of_the moititor-ing stations.
The source of the tritium is still unknown at this-time.
Potential sources of the tritium are the two Process Water Storage Tanks (PWST) or the TMI-2 Borated Water Storage Tank, all of which contain Accident Generated Water (AGW).
The monitoring stations and observation stations are a series of shallow wells surrounding the TMI-2 reactor building. MS-2 is located between the borated water storage tank (BWST) and the more distant processed water storage tank number 1 (PWST-1).' The ground-water monitoring program originated in 1980 to detect leakage of the 600,000 gallons of water which flooded the TMI-2 reactor building basement after the March 1979 accident.
' Though no leakage from the reactor building was ever observed, elevated tritium levels in the monitoring stations ha've been-periodically observed due to leakage from the BWST and its' associated equipment.
The highest level of tritium measured since the ground-water monitoring program was established was 1.100,000 (1.1 E 6) pCi/ liter.
This sample was obtained on March 23, 1982 from-observation station OS-17,_which is near the BWST. :During early 1982, between 2500 and 3000 gallons of water leaked from piping associated _with the BWST.
The-leaks.were repaired and'other corrective measures were implemented, including installation of'an - enclosure and catch basin around the BWST recirculation pumps and valves.
The aquifer containing.the_ tritiated water is intercepted by the-Susquehanna river, preventing the tritium from migrating to offsite.
wells. The estimated groundwater transport time to.the Susquehanna river is between 1.25 and 10 years.LThe Commonwealth of Pennsylvania monitors water in the Susquehanna river at two locations ~which are- -
.- . , ' . , 5-downstream from TMI.and upstream from Peach Bottom.
One is at York Haven, upstream from the first drinking water intake at Brunner Island.
The other is at the City. of Lancaster's water intake. None of the monthly composite samples taken at York Haven during the years 1982-1985 (when TM1. groundwater tritium levels were high) were above the lower limit of detection (LLD) of approximately 300 pCi/ liter.
Only two of the daily samples taken at Lancaster during 1982-1984 showed tritium above the LLD and both were less than 500 pCi/ liter.
The licersee is actively investigating various-possibilities to-- determine the source of the tritium and the inspector had no concerns with the investigation.
Licensee activities associated with this will be reviewed in a subsequent inspection.
4.
Maintenance and Surveillance Observations 4.1 Routine Maintenance Observations The inspector reviewed selected maintenance activities to assure that: The activity did not violate Technical Specification -- Limiting Conditions for Operation and that redundant components were operable; required approvals and releases had been obtained prior to -- commencing work; procedures used for the task were adequate and work was -- within the skills of the, trade; activities were accomplished by qualified sonnel; -- where necessary, radiological and fire ' preventive -- controls were adequate and implemented;- QC hold points were established where required and M served; -- functional testing was performed prior to declaring the -- particular component (s) operable.
equipment was verified to be properly returned to -service ( -- Maintenance activities reviewed included:' Preventative Maintenance Procedure E-18, Rev.9, Battery Chargers -- Annual Inspection. The inspector observed this procedure for j
Battery Charger "F" on September 10, 1990 and for. Battery Charger "E" on Ser.tember,'ll',.1990.
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Several weaknesses were identified'and are described in-the following-sections.
4.2 Specific Checklist to Ensure Operability of Redundant E.uipm:nt . On September 10, 1990 the inspector observed Preventati e Maintenance.
t Procedure E-18, Rev. 9, " Battery Chargers - Annual Ins,ection" for the "F" battery charger.
Step 4.4.1 of this proc c e is a prerequisite that states, "Using a specific chtcklist, an inspection ., has been performed to verify switches,. breakers, valves etc., for redundant strings of safety related equipment are in proper position to provide a fully operable redundant string " When the inspector asked the technician to see'the' checklist required by this, step, the technician indicated that a specific checklist had not been- _ ' performed.
The electrical maintenance manager stated that this was'a standard step in many-of their procedures and it was an operations department.
i responsibility to ensure that this.was done. Operations department.
i personnel caid that they do not perform a specific -checklist for -any l of the procedures. Operations department personnel-indicated.that even though a specific checklist was not performed, the: intent'of the step was accomplished by Administrative Procedure 2002, " Rules for- ! , Protection of Employees Working on Electrical and Mechanical Appara-
tus."
Also, they said that'the shift-supervisor is responsible to ensure that safety related equipment is.-taken out of service
properly.
The procedure upgrade program that the. licensee has in j place rewords this prerequisite and therefore. deletes the' requirement - [ for this specific checklist. After the inspector identified-the' concern, the-licensee issued a memorandum to maintenance department j [ personnel stating that until the time-when all procedures get;up- = graded, the step will be accomplished by. operations department' personnel using the tagout procedure instead of the specific check-list.
The inspector reviewed the tagout and the work package' associated
j with this maintenance to verify that-redundant _ equipment was oper- . l able. The inspector determined that required _ equipment was; operable-
and agreed with the licensee that even though'a specific checklist g was not performed, the general purpose of the step was still met, ' The inspector concluded that it'_ was inappropriate to' allow thist to remain in many licensee procedures for several years.
The pre- ' ! step - requisite does not accurately reflect the way the. licensee'is verify-
ing operability'of redundant equipment.
j* Inthis. case,thelicenseenotperformihg.thisstephasilittle! safety-
significance since the ganeral purpose _ of the step has.beenLmet.-. = However, this philosophy of, not. performing steps as written could.
have significant safety implicati' ns when : performing other proce-o dures.
This is an example in which the fail _ure to adequately address-
procedure compliance in' Administrative Procedure:10010 has - s
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resulted in the extension of the compliance flexibiltiy beyond its intended scope.
Failure to provide the specific checklist required by step 4.4.1 of Preventative Maintenance Procedure E-18 is a-violation cf the licensee's Technical Specification 6.8.1=. This item is characterized in Appendix ~ A Notice of Violation item "b" of this report.
(50-289/90-15-02)~ 4.3 Performing Work with Handwritten Step-by-Step Instructions On September 11, 1990 the inspector observed Preventativ' Maintenance e Procedure E-18, Rev. 9, " Battery Chargers - Annual Inspection" for-the "E" battery charger. The inspector observed that the technician-performing the procedure had a page.-of handwritten step-by-step, instructions to supplement the approved procedure. These' handwritten instructions appear to conflict-with the technical manual and/or.. operating procedures. The inspector has not had ~ sufficient' time to evaluate this issue and discuss-his findings with the licensee.
This item is unresolved pending further evaluation.of Pre'ventative Maintenance Procedure E-18 (50-289/90-15-01).
j 4.4 Routine Surveillance Observations The inspectors witnessed / reviewed. selected surveillance tests to determine whether properly-approved procedures were -in use, details - were adequate, test instrumentation was properly calibrated and used, . i Technical Specifications were satisfied,- testing was performed by' l qualified personnel and test results satisfied acceptance criteria or ' were properly dispositioned.
The following-surveillance testing activities were reviewed: Surveillance Procedure 1301-5.8, Station Battery Monthly, h -- Inspected on September 9, 1990.
Surveillance Procedure 1301-4.6,- Station Battery: Weekly.- -- Inspected on September 9, 1990.
' i . Surveillance Procedure 1303-4.1, Reactor Protection System.
-- Inspected on August 22, 1990.
Surveillance Procedure 1303-5.1, Reactor Building Cooling and i ' -- ' Isolation-System Logic Channel'and Component Test.
Inspected ons September 17, 1990.
, . Surveillance Procedure 1301-5','3, Incore Neutron Detectors-Monthly -- Check. Inspected September 9;:1990.
One concern was identified and=is described below, t > .,
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4.5 Inadequate Procedure to Test Reactor Protection System On August 22, 1990 the licensee was performing _ Surveillance Procedure.: (SP) 1303-4.1 " Reactor Protection System" for. channel C.
Step 8.6- - tests the Reactor Protection System (RPS) response to a simulated reactor coolant pump trip at a given reactor power level. Proper RPS response is verified by observing several bistable _ lights changing from dim to bright, indicating the bistables'have tripped. When testing the RPS response to tripping _ reactor pumps when simulating 55-percent power, it was. noticed that some of the bistables that were.
required to be checked to ensure proper RPS response _ were already tripped.
Further investigation indic'ated that there was no specific' step in the procedure directing the technician to reset the bistcbles following the previous test. Prior to this, the procedne was written such that after completion of testing of one.RPS funtion, the initial conditions were establishedJfor testingcof the next function.- In step 8.6 the procedure did 'not establish the initial conditions and therefore required the technicians to: establish initial condi-tions based on their-training.
If the. technician does not notice this, then this function of the RPS:goes. untested.
When the problem was questioned by the inspector,'the technician - contacted his supervisor. The supervisor referred the technician:to . step 5.15.1 in the limits and precautions section-of the procedure which states: At a number of locations in this test-procedure,.the Reactor Trip Module may already be tripped from a trip feature other.than the one being tested. This can occur from test module. switching transients or from off-normal test voltages.
Should'this occur, reset-the Trip Module before verifying the: trip' feature'being tested. -To do.this the bistable which caused the trip mustlalso be reset.
The inspector concluded that'the. procedure was~ inadequate'for the i .. following reasons- ' l In all fifteen previous instances whe're resetting..of bistables- -- was required, there'was a. specific' step in'the' procedure stating-which bistables to reset and'thereby established the initial _ - q conditions for testing of-the next function.
If the. technician
i _ does not notice that the; procedure-does not:do-this in step'8.6',. ! then that function of the RPS.goes; untested.
} - The technicians' performing this test have performed-1t manyLtimes ] -- and they:were unaware;that step?S.15.1' existed.
l'l . Step 5.15.1 has.not always~been in'the-procedure. 'The step was; > -- - added in January,1986 when technicians-found _that removing. the ~ , test equipment;from the' modules?sometimes caused a slight- } .s .j c
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.l Q ' voltage spike which tripped the bistables..This: step: allowed' them to reset the channel when this occurred.
Each surveillance procedure has a procedure review checklist that can be filled out after a procedure is performed.
The~ checklist asks if-
the procedure adequately covers the steps performed and if the steps
were clear.
The deficiency discovered _in step 8.6 was:never ' addressed in one of these checklists-in the past.
'? The licensee has agreed that there was a weakness in the procedure and- ! issued a change to the procedure to add a step'after.in step 8.6.that: } specifically states to reset'the bistable after completion of_the ~ ' testing of the previous function. This change'was. completed prior _to a testing RPS channel "D":on August _ 29. 1990; They also reviewed the: j rest of the procedure and found no similiar~ problem.
The-licensee' ' also agreed to consider looking for this type.of problem in other procedures in the biennial review process.
,y The inspector concluded that the licensee did not provide adequate-a guidance to establish initial conditions for-testing.RPS response-to-i tripping reactor-coolant pumps while simulating 55 percent power.
! This is a violation of the licensee's Technical Specification 6.8~1.
. This item is characterized ~in Appendix A Notice of Violation item "a" of this report. (50-289/90-15-02) 5.
Security . 5.1 Routine Security Evaluations ' Implementation of. the Physical Security Plan was observed in; the-following plant areas: .. ! Protected Area and Vital Area barriers were well maintained and --- not compromised;
Isolation zones were clear;- -- t Personnel and vehicles entering and packages!being delivered to: I -- the Protected Area were properly, searched and access control was.
' in accordance with approved; licensee-procedures; A Persons granted access to the site were badged to' indicate ~ i -- whether they have unescorted access or escorted authorization; < l t Security' access co'trols to Vital-Areas were.being maintained'and- ' -- n that persons in: Vital l Areas were:-properly authorized;: Security posts were adequately' staffed and equipped, security -- personnel were alert and'knowledgeableiregarding position'... > ' requirements, and that written procedures were(available;.and! j q.
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-- No noteable observations were made.
6.
Emergency Preparedness 6.1 Ani.ial Emergency Preparedness Exercise On September 20, 1990 the licensee conducted its. annual Emergency Preparedness exercise. An inspection of the exercise was conducted by a team of inspectors and is documented intInspection Report 50-289/90-80.
7.
Safety Assessment and Quality Verification 7.1 Special 10 CFR 50.59 Report Review , The inspector reviewed the 1989.10 CFR 50.59 Report for completeness, accuracy, understandability and timeliness.of the evaluations included in the report.
This report contained summaries of 12' permanent plant modifications,. five temporary plant' modifications, eight plant procedure revisions and'one;special temporary procedure.
A number of these 10 CFR 50.59 evaluations were selected for detailed '! - review. An additional sampling'of evaluations completed since the report cutoff 'date was reviewed. A number of reviewers involved inn i performing 10 CFR 50.59 evaluations and personnelLinvolved in.
, training the reviewers were interviewed.
Iniaddition,.a: review:was '
performed to determine the extent to:which the guidelines published in-NSAC/125, " Guidelines for 10 CFR 50.59 Safety' Evaluations,? are ' being used at TMI-1 and whether improvements lcan be made in those guidelines.
- During the review, the-inspector notediseveral weaknesses' associated
, with specific evaluations which are described as;follows.
i Safety Evaluation No. TI-135400-009 was performed to suppo.rt Cycle-8 ^ core reload, using Technical, Data Report (TOR) No'.L 989las the basis.
TOR 989 was revised on March 2,1990, to reflect a number of changes - in the core design, including -use of-a reconstituted :(or.,recaged) fuel'
' assembly as the result of two defective. fuel pins in-one of the-assemblies that was-to be' reused..Although.the TDR and SE were . technically: sound end adequate, the inspector noted that Revision 1Lof-the.SE was signed-off on February 25,;1990, severalidays before'the' supporting revision'to.the'TDR was signed o.ff. This 1s an " administrative irregularity'that'could have invalidated the SE.
Special Temporary Procrdure No. 1-90-0014 was : prepared ' for actual-recagirig of the defective' fuel assembly. 'A150.59 Safety Evaluation ai ' was also performed'for.this' procedure, - ,
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. ' a1 , ' -. . , l Change Modification Request (CMR) No. 89-036-M, covering modifica-- ' tions to valves PP-V100/103/132/135, was reviewed.
Attachment I to
the CMR was a safety determination on which the answer to questions.4 and 5 was "yes."
The answer to question 5 was later changed to "no."
, The footnote to the form states that "if-any of the answers to-questions 3,4,5 or 6 are yes, proceed :to Exhibit 8 and prepare a- , written safety evaluation." No Exhibit 8 or safety evaluation could i be found. The safety environmental determination _for CMR-89-036-M ! did reference a previous SE No. 128124-001 which addressed disabling the automatic actuation of the penetration pressurization system.
' This SE was adequate consideration of a safety concern although a
formal exhibit 8-should have been prepared to note this fact. Also, the answers to these questions contained the statement "the change'is' - ,' NITS," apparently meaning "not important to safety." This_ term is not discussed nor defined in licensee procedures and is obsolete at", ! TMI-1, Furthermore, NSAC/125 contains the-statement "Accordingly,-
the focus of the 50.59 evaluation is-to clearly define the proposed change, test, or experiment and evaluate its potential effect(s) on.-
the SAR conclusions without regard to any overall classification such-
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as ITS."
The safety evaluation in support of CMR 89-036-M is neither: ' consistent with the licensee's procedure nor with the: guidelines in' NSAC/125.
Procedure Change Requests (PCRs) 1-0S-89-0053 and 0054 were rev.iewed.
These changes were to data tables associated with emergency diesel generator procedures OP 1107-3-and SP 1303-4.16.
The associated.
safety evaluation was cursory in that.it did not state why the change has no adverse impact on nuclear safety; it merely stated that<it-didn't with rationale not given_.. _ PCR 1-0S-88-0676 was reviewed.. This PCR deletedLthe requirement to manually trip the reactor upon inadvertent-closure-of a main steam isolation valve.
The initial _ technical evaluation.in: support of this change was completed in January 1986. The 50.59 safety evaluation,of the proposed change was not prepared until September 1988' The safety evaluation discussed why this' change would:be acceptable - because the_ condition created by closure of a-main steam isolation.
valve is similar to that during main-tu'rbine:stop'and control valve testing in accordance with procedure OP,1106-1.
The evaluationdid ' not, however, specifically address' the questions in-50.59,. includingi what margin of safety was considered and.why it:was not reduced. The.
change to the procedure was finally approved in May 1989;. The safety evaluation for installation of: temporary l Jumpers.on pressurizer heater _ groups 12-and 13 was' reviewed; The _ev'aluation used'- the-rationale that an unreviewed safaty question'was:not involved because.the heater banks are "not impor e t to safety" and did:not specifically address the -50.59 questions, o t > t'- a
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The inspector also noted several. safety evaluations in which he had no concerns.
Safety evaluation No. 000224-009, "0TSG Operating Level Verification," was reviewed. This evaluation involved a special test conducted at TMI-1 in June _1990 to determine the effects'of raising steam generator level above the administrative limit established in
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TMI-1 operating procedures.
The evaluation was very thorough and ! complete and could serve as a model for future safety evaluations at TMI-1.
! Procedure Change request No.'l-MT-90-1003 was processed _to eliminate.
> hot clearance measurements on the Main Steam Relief Valve Support System.
The-Safety Evaluation of_the change was' detailed and.
. i adequate, and included a table of historical results of.these' readings since 1985.
A review was performed to determine-the extent to which the guidelines published in NSAC/125, " Guidelines for 10 CFR 50.59 Safety'Evalua-tions," are being used at TMI-1 and wnether improvements can-be made . in those guidelines. NSAC/125 is the product of an effort by NUMARC and the NRC to provide detailed guidelines to interpret the intent of 10 CFR 50.59.
Licensee Administrative Procedure -( AP) 1001A, " Procedure Review and Approval", provides. instructions for changes to procedures including how to implement 10 CFR-50.59' requirements and provides forms for documentation of 50.59 safety reviews.
Changes-were made to this procedure in ear _1y-1989, partly in response to-previous NRC inspection report comments.
These changes were'made prior to publication of NSAC/125 (June ~1989) and AP 1001A'was not subsequently revised to specifically adopt NSAC/125 guidance.
However, the retraining provided to reviewers in 1988 covered.
NSAC/125 and NRC comments to earlier draft versions of the document.. ' ~ The above observations indicate that=, although the;qualityLof!10 CFR 50.59 reviews at TMI-1 is generally improving when compared with . observations in past NRC inspection reports, additional improvement ' is needed to fully meet the intent,of this regulation. LThe staff recognized-that-some. training-has been given:at TMI-1 on.the philo-sophy of NSAC/125 and that in some areas, AP 1001A is even more-conservative than NSAC/125. ~ Vse of th~ : NSAC/125 guidelines by . e , I utilities is voluntary and the' current' version ~.may be revised before ! formal NRC endorsement occurs. However', the inspec' tor believes that ~ l increased familiarization with and use of these' guidelines, even in: ^ their present form,-will result in higher-quality 10;CFR 50.59: evaluations at TMI-1 and, therefore,-reduce;the, chance of unreviewed safety questions'being overlooked as changes are'made.
, 8.
. Followup of Previous Inspection Findings The NRC Outstanding Items (01) List; and the Safety Issues Management-System list was reviewed with cognizant-licensee personnel. -Items- , ' selected by'the inspector were subsequently reviewed through: ( s .,, i .
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> discussions with licensee personnel, documentation reviews and f.ield inspection to determine whether licensee actions specified in the Ols had been satisfactorily completed..The overall status of previously ' identified inspection findings was reviewed,~ and planned / completed . licensee actions were discussed for the-items reported below, ' 8.1. (Closed) Unresolved Item (50-289/87-09-06) Adequacy.of Calibration for HSPS and Other D/P Instrumentation Used in High Pressure Service (Reg, Guide 1.97) This issue concerned the calibration of various: instrumentation used in high pressure applications such as heat sink protection system (HSPS) OTSG level instrumentation.
The licensee had established an adequate calibration procedure but repeatability of data on succes-sive calibrations had not been obtained. -Inspector review of cali-a bration checks of this instrumentation during the' previous two operating cycles revealed that proper l calibration had:been - meintained. Additionally,~in' inspection report 50-289/90-10, as region based inspection of Reg,' Guide 1.97-instrumentation calibra-tion revealed no problems.
Based on the above,.this item-is closed, 8.2 (Closed) Unresolved Item (50-289/88-28-01)rLicensee to Document Acceptance Criteria for ABT Functional Tests-This item concerned the verification of. automatic bus: transfer (ABT) devices to properly function duringotesting of various; transfer i devices such as the vital inverters. The lic9nsee changed Survell-lance Procedure SP 1303-11,10 Engineered Safeguards 1aad System Emergency Sequence and Power Transfer Test, 20 provide. a; verification- . .. that the inverters transferred from the normal AC source to the; l batteries on loss of power.
The inspector also verified that j preventative maintenance was scheduled and accomplished on various electrical'switchgear (including trans'fer~ devices); Thisiitem is~ closed, -j ! 8.3 (Closed) Violation NC4 (50-289/87-03-01)': Adequacy offthe 10'CFR
50,59 Review Process j ~ h This unresolved item concerned the adequacy.of the procedures, train-
ing, and implementation used by the licensee'to, meet the requirements.
of.10 CFR 50.59, Specific issues-were (a') no. identification-in the procedures of.the responsible technical'and? independent reviewers ' for specific changes (b) safety evaluationsLwere notLperformed for
certain changes because the component or system' involved was not. I classified "important to safety", (c) inconsistent;terminolo'y.
g applied in the procedures used to-comply (withe 10 CFRy50.59,-(d) -) inconsistency between-the governing plant / corporate procedures-.and i the TMI-1 Technical Specifications, (e) guidancefregarding which-I procedures are "as described inxthe Safety. Analysis Report (SAR)", q (f) lack of a definition of " Licensing Basis Documentt in the' pro-
cedures, and.(g) too narrow of a scope for which; safety evaluations ! -should be performed for procedure revisions.
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During the three year period since th: 3 violation was identified,-a number of activities and procedure revisions pertaining to 10 CFR 50.59 reviews have taken place.
One of these activities has been the development of guidelines regarding 50.59 evaluations by the nuclear f industry (specifically NUMARC) and development of NRC staf f positions relative to those guidelines.
The result, to date, has been issuance of NSAC/125, " Guidelines for 10 CFR 50.59 Safety Evaluations" in June 1989.
The NRC is in the process of developing a formal endorsement of that document.
In the interim, it has been issued for voluntary use by utilities. Action by the licensee has been (a)~to revise the corporate procedure governing safety reviews, 1000-ADM-1291.01', in October 1987, (b) to meet with the NRC staf f in April 1988 to work out remaining differences in this area, and (c) to develop additional guidance regarding when a change has a potential adverse' effect on-- nuclear safety or safe plant operations.
The latter guidance:was conveyed to the NRC in a letter dated May 6, 1988.
Subsequently, both the corporate and TMI-1 procedures were revised and personnel were retrained on the procedures as well as the philosophy of NSAC/125.
Additionally, to resolve terminology inconsistencies and provide additional definitions, the NRC issued License Amendment No. 141 on , June 3, 1988, in response to a request by the licensee.
In general, ' the amendment deleted the term "important to safety".and replaced.it with "affecting nuclear safety" and defined the term " substantive".
With the actions discussed above, the staff. concerns expressed in Inspection Report 50-289/87-08 have been addressed and resolved. :The existence of clearer guidelines, both in the licensee's procedures and in NSAC/125, has provided a more uniform understanding between the staff and the licensee regarding changes under 50.59 and, in particular, revisions to plant procedures.
The staff.will continue - to monitor licensee performance in this area.
This item is closed.
8.4 (Closed) Unresolved Item (50-289/87-08-05) Licensee Disposition Independent Safety Reviews of STP's and Operation Procedures 'i This item concerned inadequate reviews of licensee Special Temporary l Procedures (STPs).
These procedures are written to cover evolutions ' that are acccmplished on a one time or infrequent basis. The staff was' concerned that.these procedures would not receive a proper-safety _ review as they were written and approved by only one department of the' licensee organization.
The-licensee evaluated this concern and revised procedure EP-027 to ensure a technical function organization review of these procedures, when required.
The inspector reviewed several recently completed STPs such as the one used-for the recently-completed 0TSG level modifications and noted a complete and comprehensive safety evaluation.
Licensee action on this' item , adequately resolved the inspectors concen.. Additionally, in section 8.3, a complete discussion of licensee action on improving the 10 CFR 50.59 safety review process is discussed.
These actions also contributed to resolution of this concern.
This item is closed. '
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8.5 (Closed) Unresolved Item (50-289/88-16-01) Licensee Evaluate' Acceptability of Feedwater Flow Delta Pressure Instrumentation This item concerned the acceptability of the electrical connection and the calibration drif t of the main feedwater pump delta pressure instrumentation.
This concern was similar to the one identified in Unresolved Item (289/87-09-06) which was closed in NRC Inspection . Report (50-289/90-12), therefore, this item is also closed.
8.6 (Closed) Unresolved Item (50-289/89-80-04) Strengthening of Biennial Review Process During the E0P team inspection, the staff commented that the' licensee-biennial procedure review process should have detected more of the types of procedure weaknesses that.the team identified. The licensee - agreed to review the applicable procedures for possible improvement.
The licensee subsequently revised Administrative Procedures AP 1001E, Writers Guide for ATP's, and AP.1001K, Biennial Procedure Review.
The inspector reviewed these procedure changes and concluded that the changes would adequately alert procedure preparers and~ reviewers to the types of problems identified by the team.
Licensee action on this item was satisfactory and this item is closed.' 8.7 (Closed) Unresolved Item (50-289/89-80-03) Licensee to Evaluate QA-Review of Emergency Operating Procedures (EOP) This item concerned an E0P team inspection finding th'at the licensee-QA organization did not have an in-line review function _for the - E0P's.
The licensee QA plan is specifically designed-to allow QA participation in the E0P process at the implementation stage.
The licensee provides this QA invohement via review of-operating activi-ties and simulator activities through monitoring actions of the Operations QA monitoring group.
The licensee took exception to the staff assertion that the QA reviews of E0Ps should haveibeen:during the procedure preparation stage. -The. inspector reviewed several QA Modifications / Operation monitoring _ reports that-documented licensee activities in this area, and concluded that the licensee had imple-mented their QA coverage of this activity as prescribed by their QA plan.
Recent review of good operator performance in using the E0Ps has also been noted in recent licensed operator requalification exams.
Based on the above, the inspector determined that licensee -QA involvement in the E0P process was acceptable and this _ item is closed.
8.8 (Closed) Unresolved Item (50-289/89-24-02) Licensee Review Procedure for Restoration of Systems Af ter ESAS Actuation This item concerned an inadvertent ESAS actuation during Surveillance testing.
Following-the event, the inspector noted that the licensee-had not reset the IC ES valves and IM DC distribution panel transfe'r switches. This was evident several days after the event when an
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' attempt was made to transfer the IC ES valves power supply.
The operator reset the lockout relays and was able to complete the transfer.
Subsequently, the licensee changed Operating Procedure . OP-1105-3 " Safeguards _ Actuation System" Revision 28 to accomplish the reset actions following an ESAS actuation.
The inspector reviewed the procedure revision and concluded that licensee action to resolve this concern was adequately addressed.
This item is closed.
8.9 (Closed) Safety Issue Management System-Item (I.C.1.2.A and I.C.1.3.A) Guidance for the Evaluation and Development of Procedures for Transients and Accidents, Implementation of NUREG 0737 item I.C.1 by the licensee was completed in several parts.
I.C.1.2. A ensured that Emergency Operating _ Pro-- cedures (EOPs) adequately addressed. concerns about adequate core cooling and I.C.1.3.A ensured E0Ps adequately addressed various transients and accidents.
The E0Ps (called Abnormal Transient - Procedures) were inspected using Temporary Instruction (TI) 2515/79 " Inspection of Emergency Operating Procedures." Results of the inspection are documented in Inspection Report (IR) 50-289/84-11 and 50-289/87-23.
In inspection report 50-269/87-23, the inspector concluded that the Abnormal Transient procedures had been extensively scrutinized by management and had undergone many changes in order to improve overall quality of procedures and TI 2515/79 was c_losed.
The E0Ps have also been extensively reviewed in a team _ inspection which is documented in inspection report 50-289/89-80.
These items are closed.
8.10 (Closed) Safety Issue Management System Item (II.F.2.4) Instrumentation for Detection of Inadequate Core Cooling.
The NRC Order for Modification of the License for TMI-1, dated- ' December'10, 1982 required the licensee to provide additional instrumentation to detect inadequate core cooling during post-accident conditions.
NUREG 0737, Clarification of TMI-Action Plan _ Requirements, Section II.F.2,-provided. specific requirements and.' guidelines for this instrumentation. This equipment is currently installed and operational.
This instrumentation has been inspected in Inspection Report (RI)' 50-289/87-01, IR 50-289/87-09,: and IR S0-289/87-13 with no outstanding concerns.
This item is closed.
8.11 (Closed) Safety Issue Management System Item (II.D.3.4.3) Control Room Habitability - Implement Modifications The Supplemental Safety Evaluation' dated August 14, 1986 completed the NRR review of this-item.
However, as part of this review, the licensee made commitments to complete modifications during the cycle 6 refueling outage.
These modifications have been inspected and are documented in Inspection Report 50-289/87-11.
There was'only one outstanding concern which related to spurious high chlorine detector response.
This concern was inspected in Inspection Report 50-289/ 88-07 and the concern was resolved. This item-is closed.
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8.12 (Closed). Safety Issue Management System Item (II,E.1.2;1.b) A'uxiliary ! Feedwater. System Automatic Initiation and Flow IndicationL '~ ] This item concerns the' upgrading of the automatic' initiation;signalsL 's and circuits for Emergency Feedwater (EFW) t'o meet safety grade L . ? ' requirements. This was inspected'and documented in Inspection' Report ' !, 50-289/88-16 Section 4.0, Emergency Feedwater. System Upgrades.
The d inspector had_one concern relating;to the_reliabi.11ty of;the!1oss of.
! main feed flow signal and this:was made open item 50-289/88-16-01.; , Open item 88-16-01 is being: closed in this inspection report. ;This:
item is closed, . g, n 9.
Exit Meeting U d . . . .- A summary of inspection findings was further-discussed with'the -...
licensee at the conclusion of.the report. period on. September.21,;1990.
Persons designated _with an asterisk inLSection 1.3.were-present. at the; ! exit meeting.- U} - , ' . -. . T , L;
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