IR 05000289/1987017
| ML20236E109 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 10/22/1987 |
| From: | Cowgill C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20236E080 | List: |
| References | |
| 50-289-87-17, NUDOCS 8710290069 | |
| Download: ML20236E109 (27) | |
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U. S. NUCLEAR REGULATORY COMMISSION i
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REGION I
Docket / Report No. 50-289/87-17 Licensee: DRP-50 Licensee: GPU Nuclear Corporation P. O. Box 480 Middletown, Pennsylvania 17057 Facility: Three Mile Island Nuclear Station, Unit 1 Location: Middletown, Pennsylvania l
l Dates: September 4 - October 2, 1987 Inspectors: R. Conte, Senior Resident Inspector (TMI-1)
D. Johnson, Resident Inspector (TMI-1)
S. Peleschak, Reactor Engineer, Region I Reporting Inspector: D. Johnson, Resident Inspector l Reviewed by: R. Conte, Senior Resident Inspector
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Approved by: /o/.u/87 C.Cogill, Chief,ReactorSectionNo.1A,DRP 'Da te' j l
Inspection Summary:
The NRC resident staff conducted safety inspections (142 hours0.00164 days <br />0.0394 hours <br />2.347884e-4 weeks <br />5.4031e-5 months <br />) of power operations, focusing on performance in operations, which included event ,
response; maintenance; and, surveillance areas. The reactor trip of i September 16, 1987, was reviv,a 1 Items reviewed in the plant operations )
area were: reactor startup, wss of ventilation testing, and a radwaste discharge problem, Other items reviewed included: licensee event reports, radiation protection program implementation, diesel generator i non-essential trip bypasses, and licensee actions on past inspection !
finding pDR710290069 871022 g ADOCM 05000289 PDR
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Inspection Results:
No violations were identified in this repor One unresolved item, concerning the discharge of a liquid waste tank with RM-L-6 in defeat, was identifie Licensee response to the turbine / reactor trip on September 16, 1987, was acceptable. Licensee management was actively involved in the post-trip review process and the plant was restarted without major problems. Late identification of a problem with one of the Once-Through Steam Generator (OTSG) level transmitters, which was evident approximately one hour after j the trip, was a weak aspect on the review of post-trip shutdown activities.
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Licensee accomplishment of routine maintenance and surveillance activities without major problems was noteworthy. The radiological control aspects of this work was also noted to be acceptabl Licensee ventilation system testing to validate assumptions for the fire hazards analysis was conducted without any interruptions of normal plant activitie Finally, licensee action on previous findings was noted to be accomplished in a timely fashio .
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' 1. 0' Introduction and Overview j
1.1 NRC Staff Activities 3 The overall purpose of this inspection was to assess licensee activities l during the power operations mode as they related to reactor safety and I radiation protection. Within each area, the inspectors documented the i specific purpose of the area under review, acceptance criteria and scope i of inspections, along with appropriate findings / conclusions, The {
inspector made this assessment by reviewing information on a sampling basis through actual observation of licensee activities, interviews with i licensee personnel, maesurement of radiation levels, or independent calculation and selective review of listed applicable document I 1.2 Licensee Activities During this period, the licensee operated the plant at essentially full I power. The plant tripped from 100 percent power on September 16, 1987, and was restored to full power on Septembe,r 17, 1987 (see Section 5).
2.0 plant Operations 2.1 Criteria / Scope of Review The resident inspectors periodically inspected the facility to determine the licensee's compliance with the general operating requirements of Section 6 of the Technical Specifications (TS) in the following areas:
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review of selected plant parameters for abnormal trends; !
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plant status from a maintenance / modification viewpoint, including plant housekeeping and fire protection measures;
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control of ongoing and special evolutions, including control room personnel awareness of these evolutions;
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control of documents, including logkeeping practices; i
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implementation of radiological controls; and,
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implementation of the security plan including access control, boundary integrity, and badging practice l
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The inspectors focused on the areas listed in Attachment I
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2.2 Findings / Conclusions 2. Liquid Radwaste Tank Release On September 26, 1987, at 6:30 p.m., the licensee initiated a routine release,to the Susquehanna River of the "A" Waste Evaporator Condensate
] Storage Tank (WECST). The WECST's are two tanks used to collect the distillate from evaporator liquid waste processing. At 6:35 p.m., a spurious " Alert" and " Alarm" condition on the tank effluent monitor (RM-L-6) was received in the control room. The alarm condition was supposed to shut a discharge valve, WDL-V-257, by interlock; but it did no Upon investigation by shift personnel, they immediately identified that the interlock was in " defeat" (inoperable) and they, then, enabled it by placing the appropriate switch into the " enable" position. The shift continued the release with RM-L-6 not being in an alert or alarm condi-tion, i
The resident inspector became aware of the problem upon log review on the morning of September 28, 1987; and, later that day, licensee representa-tives reported to the inspector that the event was reportable in accor-dance with 10 CFR 50.73. The event was reportable because the applicable l Technical Specification (TS) was not met with a release initiated and the RM-L-6 interlock defeated. The licensee initiated a review of the event in preparation for the required 30-day repor On September 29, 1987, the licensee issued an internal Plant Incident ,
Report (PIR) on the matter (No.1-87-07). By this report, the licensee i concluded that there was no adverse impact on the environment with a dose I assessment for the release at 0.1 percent of the TS quarterly limi The licensee reported that the switch was inadvertently left in defeat due l to personnel error of not following procedures. However, corrective !
actions also included procedure enhancement ,
In addition to reviewing the PIR, the inspector independently reviewed the applicable procedures for the release. The procedures reviewed were:
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Surveillance Procedure (SP) 1302-3.1, Revision 48, dated June 25, 1987, "R.M.S. Calibration;" and ,
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Operating Procedure (CP) 1104-295, Revision 32, dated April 29, 1986,
" Transfer from the Waste Evaporator Condensate Storage Tanks."
OP 1104-295 was the controlling procedure with sign-offs for the release, and it references SP 1302-3.1 for RM-L-6 alert and alarm adjustment just prior to the release. The SP is essentially a TS-required periodic cali-bration procedure for general area and liquid radiation monitor <
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Apparently, a number of sections in SP 1302-3.1, dealing with RM-L-6, do not have to be accomplished for the release evolution. Since OP 1104-29S is not specific on what sections are needed, it appeared that the tech-nician entered SP 1302-3.1 just before the step requiring interlock defeat ;
and he left the SP before the defeat function was restored to enable. At j the conclusion of the inspection, it was not clear how the rest of the i functional checks of the interlock were accomplished per OP 1104-295 with I the interlock in defeat, j
The inspector also noted that there were not any independent verification procedural steps in SP 1302-3.1 for any radiation monitor calibrations warranting interlock defeats. This will be reviewed in conjunction with an upcoming NRC review of licensee measures to verify correct operating activities (289/85-27-08).
The inspector acknowledged the licensee's conclusion of minimal impact on the environment. However, RM-L-6 event is unresolved pending: (1)
licensee final event review as embodied in the applicable licensee event report (LER); and, (2) NRC Region I review of the LER and the event for any implication to the licensee's program for verifying correct operating activities (289/87-17-01). Licensee corrective action appears to be oriented to addressing these problem . Post-Trip Plant Startup During September 17, 1987, the resident inspectors selectively reviewed licensee activities associated with plant startup following the reactor trip of September 16, 198 No additional plant trips occurred during that startup, especially with integrated centrol stations in manual. It appearea operators were attentive to their duties and the proper sequence of actions to assure a relatively smooth startup. For the activities reviewed, the inspector noted proper proce' Jure adherence to facility procedures. Overall, there was good coordination of startup activitie . Loss of Ventilation Testing The licensee has been conducting a series of tests to verify certain assumptions concerning the need for local ventilation systems (fans)
to maintain operability of the safety related equipment (pumps, motors, I switchgear), which are located in areas serviced by the fans or ver,tila-tion systems. Specifically, the necessity of the local ventilation fans in the in-take structure, the closed-cycle cooling pump room, the lower elevations of the intermediate building containing the emergency feedwater pumps, and the control building heating ventilation ar.d air conditions (HVAC) system have been evaluate The tests basically consist of securing the local fans or ventilation systems (as in the case of the control building) and monitoring ambient temperature increase for periods of one-to-two hour Relevant tempera-tures of the energized equipment in the specific areas, as well as ambient I
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temperatures, were monitored. The data were intended to be provided t NRR as information to be used in evaluating the licensee requested exemptions to Appendix R fire protection requirement The inspector witnessed portions of.the loss of ventilation test for'the control building accomplished on September 25, 1987. The licensee generated Special Temporary Procedure (STP) 1-87-0040, " Control Tower. Loss ,
of Ventilation Test," and an appropriate safety evaluation. .The inspector '!
questioned the licensee concerning the operability of both trains of.the control tower emergency recirculation fans AH-E-18A/B, which are required by technical specification (TS). TS 3.15.1.4 requires that if both trains of emergency ventilation are inoperable when containment integrity is required, then the reactor must be in cold shutdown within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. The licensee responded that since no auto start signals are used with the AH-E-18 A/B fans and that they would not be in " pull to lock," the oper- l" ability of the fans was not a concern. The inspector concluded that the licensee was not entering a TS Limiting Condition for Operations (LCO)
action statement for the conduct of this tes The inspector considered the safety evaluation as satisfactory to address the performance of the test. The test was accomplished satisfactorily with ventilation secured for approximately one and a half hours. None of the areas monitored exceeded pre-determined limits and temperatures returned to normal shortly after the normal ventilation was restore The licensee conducted a pre-test briefing and used several personnel at each level of the control building to monitor temperatures during the tes The inspector evaluated the test conduct as satisfactor The licensee intends to submit test data to NRR at a later date. 'The inspector had no other concerns on this issu .3 Plant Operations Summary In general, plant operations were conducted in a safe manner and oriented toward nuclear safet Problems were identified with the radwaste discha'rge and the identification of the cause of the emergency feedwater (EFW) auto initiation signal on low Once-Through Steam Generator (OTSG) level (discussed in Section 5).
3. Maintenance / Surveillance - Operability Review 3.1 General Criteria / Scope of Review The inspector reviewed activities to verify proper implementation of the applicable portions of the maintenance and surveillance programs. The inspector used the general criteria listed under the plant operations section of the repor Specific areas of review are listed in Attachment A more detailed review of equipment operability was also addressed belo !
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3.2, - Findings / Conclusions (Decay Heat Pump Maintenance))
During..the period of, September 22-24, 1987, the licensee removed.the ;
"18" decay heat (DH) loop =from service to accomplish minor repairs '
to the mechanical portion of the'"1B" DH pump, DH-P-1B, and for some ~i preventive: maintenance activities on the system. The inspector witnessed
- p'ortions of the' repair activities,. discussed the consuct of the mainte- l nance activities.with workers and maintenance. supervision, and also j
' witnessed.a portion of the return to service surveillance test. The 4 inspector also reviewed the radiological control aspects of this job as l'
' documented in Section One of the items that was-repaired 'was a leaking pipe connection to the?"18" pump casing drain valve. .This item had been noted during
'the Region I administrator tour and documented in NRC Inspection Report No. 50-289/87-1 Initially, the licensee had not planned to repair this item unti_1 the next outage; but, due to other minor 1 leaks also existing on the DH system piping, they accomplished the i
= repairs at this time'. l Since one train of the DH system can only be inoperable for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or reactor shutdown is required, job planning and work coordination must be. accomplished expeditiously. The inspector noted that the-system' isolation and draining, tagout and work performance, along with testing, were' completed ahead of schedule. No major problems ;
were noted during-the repair evolution. The inspector had no safety concerns on.the operability of the "18" DH loo .3 Operability Summary For the equipment maintenance and surveillance testing required above, and as noted in Attachment 1, the inspector verified operability and satisfactory completion of procedures. Maintenance and surveillance activities appear to be accomplished with the appropriate emphasis on reactor safet .0 Licensee Event Reports (LER's)
'The inspector reviewed the LER's listed below, which were submitted'
to the NRC Region I office pursuant to 10 CFR 50.73. Based on a review of these LER's, the inspector determined that corrective actions discussed'in the report were appropriate and that there were no generic issues. As noted below, some cf the LER's are considered
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licensee-identified violations; and, ia accordance with NRC enforcement policy, no Notice of Violation will be issue ,
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LER 87-002, dated April 1, 1987, for event on March 2, 1987,
" Diesel Generator Auto Start Due to Adjustment of an Auxiliary 1 Transformer Relay Cover During Maintenance" - This LER was I written due to an automatic start of the "1B" emergency diesel l generator (EDG), resulting from work by electrical maintenance j
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,g',- personnelfin the "1A" auxiliary transformer control cabinet. The
. root cause was' improperly " dressed" wiring ~'on the fault pressure:
relay which 'did not. allow a fit of the relay cover. This condition apparently has.existedisince construction. .The relay. wiring and-.
^ cover interference.for'this and other relay covers will be resolved during the 7R refueling outage. The licensee has determined that
. operation in the interim. poses no safety significance. The inspector-had'no oth'er concerns on this item. Thie event was reviewed in Inspection Report'No. 50-289/87-0 ;
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LER 87-003, dated May 1,1987, f 7r event on April 2,1987, " Fire Barrier-Penetration Seal Not Installed.Due to Personnel Error" -
Licensee personnel discovered that no seal wa's installed ~in a pene-
,tration in a fire zone boundary. A fire watch was posted within the hour;and the penetration was sealed on the day of discovery. -This sea 1Lwas required by 10 CFR 50, Appendix R, Section II The-licensee:has inspected all new seals to assure compliance with-th regulations = set forth in 10 CFR 50, Appendix R. The . root cause of
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this event'was failur'e of ~ engineering personnel to ' perform a sea inspection of the zone prior.to acceptance of the zone as complying-w'ith the Fire Hazards Analysis Report (FHAR). The.importance of physical.. observations to support conclusions has been stressed with the engineering personnel involved in the event. The inspector had no additional comment This was a licensee-identified violatio '
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LER 87-004, dated May 28, 1987, for event on May 2, 1987,
" Reactor Trip on High Pressure Due to Operator Error During Testing"'- This LER was written due to an automatic unplanned reactor protection system (RPS) actuation. During power range calibration surveillance on the RPS, a license operator acci-dentally transferred the " flow" signal instead of the " flux" signal. The root cause of the event was the failure of the operator to recognize that the wrong switch had been selected, which resulted
- in a reactor trip on high pressur The licensee improi/ed the af-fected procedures and counseled all operators on this event. The inspector had no other concerns on this item. This event was re-viewed in Inspection Report No. 50-289/87-10.
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87-005, dated June 3, 1987, for event on May 6, 1987, " Missed Sample Prior to an Industrial Waste Filter System Release" -
With' radiation monitoring liquid effluent instrumentation (RM-L-12) in bypass, activity samples for the waste treatment system were to be taken every eight hours and prior to a i release since a high radiation effluent would not be terminated i automatically. On May 6,1987, an industrial waste filter system release was initiated without analysis of a grab sample for activit This was a violation of the Technical Specifica-tion. The root cause of the event was cognitive personnel error l
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"(failure to-follow. procedures). -During the release, an operator was assigned to RM-L-12 to terminate the release upon.'a high alar Operations and chemistry personnel have been, counseled to assure
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compliance with procedures and' technical specification requirement '
The inspector!had no additional comments on this matter. .This-was'a
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licensee- identified violatio ' L87-006,-dated July 13, 1987, for event on June 12, 1987, " Reactor-Trip:on High RCS Pressure.Due to Operator' Error" -'This LER was written due to'a RPS. actuation which resulted from.a. cognitive
' operator error. (root: cause). The speed of'feedwater pump "1B"'was;
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~being reduced to prevent excessive differential pressure across.the:
-feedwater regulating valve as power was.being reduced. The speed of
' FW-P-1B was reduced below that necessary. to ' provide' adequate feed-water to the OTSG's. When 'ow feedwater flow was recognized, the; pump [s speed.was increased; but the reactor coolant system (RCS)
pressure had increased.to the.RPS trip setpoint and the OTSG's were slightly overfed. The licensee will emphasize feed pump h.and. control .,
during the 1987-88 training cycle. -Licensed operators have been "
briefed on this issue. .The inspector had- no other comments on this
.i ssue. This event was reviewed in Inspection Report No. 50-289/ 1
'87-1 LER 87-007, dated July 23, 1987, for event on June 25, 1987, .!
" Inadvertent Reactor Protection' System (RPS) Actuation on High l Pressure During Heatup Due to Cognitive Operator Error" - 1
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Licensee' was performing' a heatup and restoring the high pressure
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trip setpoint to 2300 psi The RCS heatup must be coordinated with the RPS. reset prior _to achiev'ng'1720 psig. With three RCP.'s running, the operator was moni;oring RCS pressure from the pressure indicator on the loop with two RCP's running, which was indicating a RCS pressure approximately 50 pounds lower than the loop with one RCP'
running. RCS pressure increased above 1720 psig. The RPS automati-cally actuated and dropped the only control group withdrawn. The high pressure trip setpoint was reset and RCS heatup was complete The licensee has briefed all operating crews and has initiated a procedure change. The inspector had no other comments on this matter. This event was reviewed in Inspection Report No. 50-289/
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In addition to the technical adequacy of the LER's, the compliance I with the requirements of the 10 CFR 50.73 format was reviewed. The i subject LER's were deficient in some area The LER's are listed below and the deficiencies of each LER are discusse I l
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LLER (87-03 -LTha Energy Industry Identification system (EIIS)'
component function identifier is not- included in the .LER.as-
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LER'87-04 - The EIIS identifier is not. included. in the repor The time of. the event 'is not included 'as' per 50.73(b)(2)(11)'.
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- LER 87-05.- The title of'this LER does not provide a cause/ resul link -The title-fails'to' indicate the root caus '
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.LER 87-06 - Acronyms used.in the LER arefnot defined. The 10 CFR 50.73(b)(2)(i) requires a clear. specific,-narrative descriptio !LER 87-07 - The EIIS code is.not included,in the' report. The
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(LER does'not give a complete discussion... The location of the
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pressure: sensors and.the resultant- pressure' indicator's
9 . location'onLthe control board and which sensors the operators
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were observing is not discusse The license'e. acknowledged'these finding ^ These. discrepancies were similar to.that' identified by NRC's Office of-the Analysis.and' Evaluation of Operational Data (AE00) for 1986 LER's (see prev.ious_ unresolved item 289/87-09-15). Although the ~
number of discrepancies appear to have been reduced, the in~pector s concludedithat licensee personnel still need to be'more attentive to the:10 CFR 50.73 requirements for a complete report. Also, the NRC staff will perform additional . reviews 'of this matter as a result of the upcoming Systematic Assessment of Licensee Performance (SALP)
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proces . Event. Review
' Background During this inspection period, the inspectors reviewed the reactor trip of September 16, 1987. In general, the following aspects were considered:
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details regarding the cause of the event and event chronology;
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functioning of safety systems as required by plant conditions;
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consistency of licensee actions with licensee requirements, approved procedures, and the nature of the event;
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' radiological consequences (on site or off site) and personnel ;
exposure, if any;
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proposed licensee actions to correct the causes of the event;
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verification that plant and system performance is within the'
limits of analyses described in the Final Safety Analysis-Report (FSAR); and,
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proper notification of the NRC was made in accordance with 10 CFR 50.7 The inspector provided a chronological / factual summary, specific scope of NRC staff review, licensee findings, and NRC staff finding An overall conclusion on licensee performance is also provide .2 Event Chronology At 5:04 p.m. on September 16, 1987, the reactor tripped from 100 percent powe The initiating event was a turbine trip. The turbine trip was caused by a high moisture separator level trip signal, which was generated by one of the six moisture separator high level trip switches. The six switches are a one-out-of-one logic that feed a common relay which generates a signal for the turbine trip bus. The licensee could not positively determine which of the six switches activated the trip. No alarm was received on the moisture separator drain tank high level alarms. Normally, if a high level had existed in any one of the moisture separator drain tanks, a high level alarm switch would have actuated and given an alarm in the control room. It initially appeared that the moisture separator high level trip sigra' ,us a spurious actuatio Subsequently and prior to turbine plant startup, licensee troubleshooting '
revealed that the level controller (LC-77E) for the "E" moisture separator drain tank was not functioning properly. This condition resulted in the flow control valve, MO-V-2E, only being able to open approximately 35 percent. With this condition, it is possible that the level had risen to a high enough level in the moisture separator to cause a valid signal, but the MO-T-1E drain tank high level switch would have had to fail to actuate
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in this case. In either situation (actual or spurious high level), one of ,
the magnetic switches would have had to fail to operate properly. The precise root cause remains unknown at this tim )
i The licensee recovered the plant after the trip to normal hot shutdown conditions; no major problems were note There were no engineered safe- ,'
guards actuations immediately following the trip. The post-trip plant response was normal with no significant problems identified during the post-trip review. The plant was restarted at approximately 4:30 a.m. on September 17, 1987, and full power was achieved at 9:30 !
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l 11-i However,' the emergency feedwater (EFW) low level initiation was '
defeated; and, approximately one hour after.the trip, the EFW auto I initiation on low "A" 0TSG 1evel alarm was received on the computer alarm printer. This alarm was noted by the inspector, based on a
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records review during startup activities, on September 17, 198 The inspector questioned the licensee about the cause of this alarm and if an auto initiation had occurred. As the EFW auto initiation on low OTSG-level " enable / defeat" switches were in " defeat" at_the time, no injection occurred;into the OTSG. The licensee. subsequently re-interviewed the operators who were on watch at the time of the ,
event to gain a more detailed explanation of the alar I J
The operators were allowing the "A" OTSG to " boil down" to the .
30-inch startup (SV) range level to allow automatic level control by
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the Integrated Control System (ICS) via FW-V-16A, "A" feed regulating bypass valve. There was a discrepancy between the SU level recorder at 50 inches and the digital operating range indication of 21 inches at the control board. The operator was controlling the level.by using the SU recorder, which apparently was indicating higher than actual leve The OTSG was allowed to boil down to the 10-inch auto actuation setpoint, which, then, actuated 1 the EFW auto initiation on two out of four low level instrument channels. The operators could not_ recall exactly which two channels caused the actuatio Subsequently, on September 22, 1987, during routine surveillance of the OTSG 1evel indicator, it was discovered that LT-1047, "A" OTSG l SU range level transmitter, was out of calibration. The licensee was unable to calibrate this detector and it was replaced and, then, i satisfactorily calibrate l 5.3 Specific Scope of NRC Staff Review for the Reactor Trips Specific to the reactor trip event noted above, the inspector verified the below-listed items:
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initial proper response of the plant to the post-trip window on the i pressure-temperature (P-T) plot;
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personnel properly implemented Abnormal Transient Procedures and prudently acted on unusual conditions;
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indication of the sequential proximate causes for the trip, along with a reasonable determination of the root cause;
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post-trip review was conducted in accordance with Administrative Procedure (AP) 1063, " Reactor Review Process;" and, p j
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no unreviewed safety issues identified in post-trip review dat !
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In addition to discussions with cognizant licensee personnel, the inspector:
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made an independent assessment of post-trip parameter response based on visible strip charts and indicators in the control room shortly after the events; ,
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' attended the licensee's post-trip review; .
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reviewed the complete post-trip review packages (No. 87-04);
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reviewed AP 1063, " Reactor Trip Review Process," for adequac .4 Licensee Findings j The licensee identified the following problems after the trip and provided corrective action as noted belo The turbine bypass valve's (TBV's) exhibited oscillating control several minutes after the trip. Turbine header pressure swings were observed to be approximately 25-50 psi Normally 5-10 !
psig variations are observe The licensee subsequently discovered broken linkages on two of the TBV's and these were repaired and tested during the startu Subsequently, TBV pressure control was observed to be i norma '
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Two body-to-bonnet steam leaks were observed in the reactor building in instrument root valves for steam generator level indication. One valve was re-torqued and the leak substantial-ly reduced. The other valve was evaluated as satisfactory for operation but will need to be repaired at the next available shutdow The licensee checked all the moisture separators, drain tank high levels, and trip switches, and could find no problem The level controller malfunction was discovered during the troubleshootin Radioactivity released as a result of the trip (mostly noble gases) from the stcom generators safety valves was 8.16E-6 curies, which was well below Technical Specifications (TS)
limits.
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5.5 NRC Findings The inspector independently confirmed the licensee findings / conclusions as noted abov Plant response was essentially as expected. The licensee adequately identified these problems and took/ planned appropriate and reasonable action for immediate correction to prevent recurrenc Operator responses to the planned evolutions and/or off-normal conditions were essentially consistent with facility operating and emergency procedure It appeared that they were conscious of and oriented their actions toward confirming reactor shutdown conditions and adequate decay heat removal. Management involvement in these events and post-trip reviews were noteworth AP 1063 was adequate to identify / confirm the sequential and root causes of the reactor trip and the post-trip review was reasonably thorough to identify appropriate corrective actions before startup.
l The inspector determined that licensee action to resolve these I balance of plant trips, which can be generated through a turbine l trip sequence, is presently an ongoing effort. The possible problem i
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with the trip scenario caused by the moisture separators had been previously identified by the license Corrective action efforts to resolve the reliability of the moisture separator high level trip remains a part of the larger effort on the part of the licensee with their participation in the Babcock & Wilcox Owners Group (BWOG) trip reduction program and their own program to reduce balance of plant trip The inspector was concerned that the licensee could not positively identify the root cause of the tri It is still uncertain which level switch malfunctioned, although the problem with moisture separator drain tank level control via LC-77E has tended to lead the licensee to believe that an actual high level did exist in the moisture separator and that the failure was in the drain tank high level alarm switch. In either case further evaluation of the balance of plant trip process is require The inspector was also concerned that the identification of the problem with OTSG level that resulted in the "A" OTSG low level actuation of EFW was not identified prior to restar The inspector concluded that a more detailed review of the circumstances of the level discrepancy and subsequent Heat Sink Protection System (HSPS)
actuation at the time could have resulted in earlier resolution of the problem. The licensee does have a process for this type of detailed review (i.e., the plant incident report per AP 1029), but the threshold level for the internal triggering of this type of review is apparently too hig _ - __
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.c < 14-The calibration of the' subject OTSG. level transmitters that caused 1theTinitia_1 problem.has been the subject of previous' inspections .
-(see Inspection Report'.No.. 50-2.89/87-09).
5.6 Trip' Summary Overa11', operator' response to the trip was orientedLtoward safety
- and, in general, in accordance with facility procedure Management involvement in the p'ost-trip. review and startup were
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noteworthy. . Licensee action to evaluate improvements .in the early identification and evaluation of abnormal events, such as the 0TSG 1evel discrepancy, are= encourage Plant' response was.as~ expected and allcsafety systems functioned appropriatel There were no challenges to the emergency core cooling system . -Diesel Generator Non-Essential Trip Bypass- I
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- Scope of Review-
. Region I' issued Regional Temporary Instruction No. 87-04-(289/87-TI-04) on--July 16, 1987. This instruction provided guidance to resident inspectors to-follow up on a potentially generic issue
<concerning-emergency diesel-generator protective trips that are not bypassed on. loss of power condition The inspector conducted a review of the logic' circuitry, operating. procedures,'and surveil-lance procedures for.the emergency diesel generators to determine:
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the actual = diesel engine trip scheme to ascertain which engine
, . protective trips are bypassed during Engineered Safeguards
- Actuation System (ESAS) signals and loss of off-site power signals;
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which bypass features are routinely tested; and, l
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for trips not bypassed, that a reliability scheme (two of three !
logic) is developed to prevent spurious trip actuation The following documents were reviewed:
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Operating Procedure (0P) 1107-3, Revision 40, dated June 16, 1987,
'.' Diesel LGenerator;"
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Regulatory Guide (RG) 1 9, Revision 1, dated November 1978,
" Selection Design and Qualification of Design Generators;"
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TMI Operations Plant Manual, Volume 1, Section A-4, " Emergency Diesel Generator and Auxiliaries;"
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.SP 1303-4.18, Revision 7, dated June 3, 1985, "4KV ES Bu Undervoltage Relay Test;"
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SP 1302-5.3,- Revision 12, dated December 15, 1986, " Diesel l Generator Protective Relaying;" !
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GAI Drawing C-600-702, Sheet 1, Revision 2, " Electrical Schematic Diesel Engine Control;" and,
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GAI Drawing 208-164, Revision 15, "4160 Switthgear (IE3) G11-02 i Diesel Generator 1B Breaker." 4 The inspector also conducted a discussion with licensee engineering personnel to review the actual logic setup for diesel engine trip .2 Findings / Conclusions I
The two emergency diesel generators at TMI-1 are Fairbanks-Morse l Model 3800 TDI-118, rated at 3000 Kw continuous load. The engine {
has the following trips: 1
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low lube oil pressure (idle speed), 6 psig at 250-810 rpm;
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start failure; i
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two out of three low lube oil pressure (running), 12 psig with j greater than 810 rpm;
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two out of three high crankcase pressure (running), 0.5 psi; l
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engine overspeed;
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stop pushbuttons at engine and control room;
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86/G generator differential current; and, l
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86/8 - 10 (1E) bus overloa Of the above engine trips, the following will not trip the diesel when an ESAS is actuated (i.e., they are the protective trips that are bypassed): !
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start failure - which is generated if the engine fails to start in seven seconds (250 rpm or 6 psig oil pressure);
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low lube oil pressure (idle speed) actuates if 6 psig oil i
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pressure is not attained at 250-810 rpm; and, l
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stop pushbuttons - at engine and control roo '
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f The following signals will shut down the diesel with an ESAS,'but the initiation circuit uses a two-of-three logic for reliability:
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low lube oil pressure running; and,
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high crankcase pressure, j j
The following signals will shut down the diesel regardless of l whether or not an ESA"> signal is present: '
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engine overspeed;
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(86G) generator differential overcurrent; and, ;
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(868) 4160 volt a.c. bus overloa ,
All of the diesel engine trips noted above are operable when a loss of off-site power (LOOP) is experienced (dead bus). Although, if an i ESAS is present coincident with the LOOP, then the protective trip bypass features are still operable. The specific trips are routine-ly tested (yearly basis) during the annual diesel engine inspection or during refueling outage '6.3 Diesel Generator Trip Summary At TMI-1 it appears that the diesel generator engine protective l trips are actuated in a manner that is reliable and consistent with
- i the applicable provisions of Regulatory Guide (RG) 1.9. Although the non-essential trips that are bypassed for LCCA signals are not bypassed during LOOP conditions, the effect on the plant of (1) a failure of the diesel to start or (2) a spurious non-essential trip during station blackout would not be significant. Decay heat removal could be accomplished by one steam-driven EFW pump and its i related control signals which are powered by vital station l power supplies. The diesel should be able to be restarted in a timely manner should a spurious engine trip occur or if the engine fails to star This information was gathered in response to Regional Temporary Instruction 87-04. This TI is considered close . Radiation Protection Program Implementation Review 7.1 Scope of Review The inspector conducted a limited review of the licensee radiation protection program implementation. The inspector accomplished this review during routine plant tours, by observation of ongoing mainte-nance and surveillance activities, review of Radiation Work Permit (RWP) records, and review of general radiological conditions in the
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17-7 a radiologically-controlled areas .(RCA). ~ Additional- review of the licensee radwas y
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program and:of ed~1n'Section ~ some operational problems in this
. 7.2 .-Findings / Conclusion '7. General--
-The presence of license radiological-controlled.~ personnel-is generally evident-at. job sites: during routine plant maintenance activities.: Th inspector observedfdirect rad-tech coverage a't the site of the."1B" decays
~ heat (DH) pump maintenance during the "B":. train outage. . The. inspector ?
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also' reviewed pre-job, weekly, and post-job decon radiation / contamination surveys of the "B'.' DH vault to- verify that ~up-to-date information was -
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l available to. worker The : inspector l interviewed radiological control.' supervisors to verif that they were aware of ongoing-activities. For the "B" train DH i
- system' outages, no pre-job briefing was' held, although' workers doing-
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l work 'in radiologically-controlled areas are frequently 1 observed by
- the-inspector to be in. contact.with'the radiological field operations staff and. supervisor-in order to obtain the appropriate !
- guidance.for the wor i a
Overall,-it appears that radiological controls personnel and their i i
management are adequately involved in routine radiological wor 'l Lack of involvement-by radi.ological controls personnel'was the
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subject of a previous violation (289/87-09-12). It appears that 1 performance in .this area has improve ]
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7.2.2 -Radiation' Work-Permits-(RWP's) !
The inspector reviewed the-RWP No. 33205, which was for work / repairs
- to'the "B"'DH pump in the "B" DH vault in the auxiliary buildin The RWP was' reviewed for the following elements:
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job description;
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radiation / contamination levels; !
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concentrations of airborne radioactivity - actual or anticipated; j
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' respiratory / protective clothing-equipment requirements; dosimetry;
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special precautions;
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expiration date;
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health physics coverage; and,-
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approvals / review All areas reviewed were adequately covered on the RW The inspector observed portions of the work in progress which included realignment of i the pump-to-motor coupling. No problems were encountered by the workers that were'not'previously addressed on the RWP. The inspector interviewed maintenance supervision at the job site to verify that there was adequate-understanding of the radiological requirements of the job. No problems were noted with the implementation of this particular RW . External Radiation Exposure The inspector observed proper use of dosimetry during work on the
"1B" DH pump. Personnel were required to wear a self-reading dosimeter, along with digital alarming dosimeters with pre-set alarm No problems were note The work was accomplished in a high radiation area as noted in licensee procedures. The area was properly posted and controlled with rope barriers and signs.. The general radiation levels at the worksite were approximately 10 mrem / hour. The inspector reviewed the dose accumulation records for personnel who worked this job. It appeared that for the time the personnel were signed-in on the RWP, the total accumulated dose was reasonable considering that not all of the time may have been spent at the actual job sit Personnel at the radiologically-controlled area (RCA) exit control i point were observed to have their dosimetry properly logged in and the dosimeter reading was properly recorded. No problems were noted with controlling / recording radiation exposur .2.4 Control of Radioactive Material The inspector observed licensee personnel entering and exiting radiologically-controlled areas (RCA) at various times. All personnel were observed to comply with requirements to obtain 1 appropriate dosimetry prior to working in the RCA by discussions '
with control point radiological control technicians and by review of the appropriate Radiation Work Permit (RWP). Exit procedures from
.the RCA require whole body frisking. Presently, at TMI-1, the 4 licensee utilizes the PCM-1, a computerized automated whole body !
frisking device that accomplishes the survey, Personnel were i routinely observed to be properly using the PCM-1. Additional
" friskers" (hand held personnel survey devices) are stationed throughout the RCA for use on an "as-required basis." The inspector had no concerns with control of radioactive materia )
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7.3 Radiological Controls Summary Based on the isolated jobs reviewed, the radiological controls program is being properly implemente . Licensee Action on Previous Findings 8.1 (Closed) Inspector Follow Item (289/82-BC-75): Board Imposi tion of Sanctions on License As a result of the TMI-1 restart hearing and related NRC investigations into the " cheating incident" among licensed operators in April 1981, the applicable Atomic Safety and Licensing Board issued its partial initial decision on the matter in July 27, 198 In that decision the board 1 considered a $100,000 monetary penalty for various apparent failures on the part of licensee management that led to the cheating incident, but the Board was not'sure of its jurisdiction to impose such a sanctio . _ .
In any case, by paragraph 2419(3), the Board recommended that the Commission impose that sanction if the Board lacked such jurisdic-tion. By Commission order, dated October 14, 1982 (CLI-82-31), the review and neea for such sanction was referred to the NRC staf The NRC staff completed its review; and, in a letter, dated July 22, i 1983, the staff proposed a civil penalty for two apparent violations l identified during the NRC's review of the cheating incident: (1)
material false statements in the licensee application and attached
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certification for the TMI-2 Operations supervisor (at the time of the TMI-2 accident) ($100,000); and, (2) because of known instances of cheating, the licensee failed to assure that its training program
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was capable of ensuring proper training of licensed operators or accurately evaluating trainee knowledge level in areas necessary for safe operation of the plant ($40,000).
Because applicable investigation reports were not released to the licensee, the licensee requested and obtained NRC staff approval to delay response to the above-noted Notice of Violation (NOV) until thirty days after licensee receipt of the subject report By letter dated January 31, 1983, the licensee submitted its own investigation into the matte By letter dated August 2, 1984, the licensee reiterated their response position but remitted the $40,000 uncontested portion of the civil penalty (CP).
By letter dated April 15, 1985, Metropolitan Edison (former licensee) paid the residual amount of the CP (under protest) and responded to certain requirements of the NOV (written statement or explanation including admission or denial and, if admitted, reasons for the violation). GPU Nuclear (GPUN) (current licensee) responded on the same date, providing corrective action taken/ planned along with results achieved and date of complianc .
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- GPUN!s response-reiterated and/or summari cd a number of. corrective'
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j Tactions which became.part.of the hearing record and.the licensee's i and'NRC: staff's certification process for TMI-1 restar Individual' -!
' licensee actions'were separately reviewed and/or verified by.NRC 3 staff-foryboth violations.- Subsequent NRC staff inspections have l confirmed continued implementation of the corrective action j l
This item was administrative 1y opened to assure proper resolution o l
'the Board's decision to impose sanctions'on,the licensee. Based on 1 the above,'that open issue is resolve !
8.2r (Closed). Inspector Follow Item (289/82-BC-76): Licensee to ;
Preserve the Record of False Certification of the Former TMI-2 !
Operations Superviso As.noted above on'th'e " cheating incident" among licensed operators in April 1981, the applicable Atomic: Safety and. Licensing Board i issued its partialfinitial decision on the matter on July'27,L198 ' A residual . recommendation on the. cheating issues was that (paragraph" l 2419(2)): j l
"The Commission direct' the NRC staff to conduct an investigation-into;the August 3, 1979 certification l of VV [TMI-2 Operations Supervisor at the time of <
the TMI-2 accident] to the NRC for the operator i H license renewal..." I
Further, by' paragraph'2422, the Board directed "the licensee to 1 preserve'all records pertaining to ... [that] investigation..."
This requirement became Restart Condition No. 1.m by NRC letter, dated October 2, 198 The recommended investigation was completed as indicated by NRC i letter,- dated July 22, 1983. A civil. penalty'was proposed by that I letter (see previous item), but the accompanying violations did not '
involve failure to preserve applicable records. 'The staff investi-gation was completed without an adverse finding-related to the preservation of record It is the licensee's understanding that the records are to remain preserved as long as the restart condition exist The inspector considered this particular matter of record preservation
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8.3 (Closed) Unresolved Item (289/86-09-01): Licensee Review Evolutions Involving Cross-Tie of the IN 480 Volt a.c. Bus with Turbine Plant Load During evolutions preceding the reactor trip on June 2,1986, operations personnel attempted to supply power to a turbine plant 480 volt a.c. bus from the IN 480 volt a.c. bus. The appropriate-ness of this action was questioned by the inspectors. The licensee evaluated the use of the IN bus for supplying power to selected non-safety buses as acceptable in certain situations. Guidance was added to OP 1107-1, Revision 18, " Normal Electrical System," to specify that the use of the IN bus to cross-tie to turbine plant loads should not normally be made at power. ihe plant operations director was designated as the individual whose permission is required to temporarily power certain 480 volt a.c. substations from the IN bu The inspector reviewed the change to OP 1107-1 and concluded that the guidance provided was satisfactory and that operators had been made aware of potential problems with using the IN bus to power turbine plant load This item is considered close .4 (Closed) Violation (289/86-17-01): Reactor Building Equipment Hatch Missile Shield Door Not Positioned Correctl During startup from the March 1986 outage, the missile shield door for the reactor building equipment hatch was not positioned in front of the hatch _until approximately five days after reactor startu This condition was contrary to the FSAR Figure 5.1-1, which denotes the shield door as part of the containment protection against the design basis aircraft impact accident. Additionally, no procedures existed to ensure that the door was in place prior to reactor startu The licensee disagreed with the contention that the door was required to be positioned at all times when the reactor was critica Due to the low probability of aircraft impact, approximately 10 8 per year as documented in licensee SER 15737-2-G07-111, it was permissible to have the missile shield moved away from the equipment hatch for short periods of tim Consequently, the licensee revised the startup procedure OP 1102-2 to include assurance that the door was properly positioned for startu Additionally, a sign is posted to notify personnel that the permission of the Operations and Maintenance Director is required to reposition the doo The inspector verified that the door was correctly positioned after startup from the most recent outage, March 1987. The inspector also verified that the appropriate sign had been posted, The inspector reviewed the change to OP 1107-2, Revision 79, and concluded that
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appropriate. guidance was available to ensure that the door would not
- be ' unnecessarily' removed from its position in front of the equipment-
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hatch.. This item is. considered close :)
8.5 (0 pen), Unresolved Item-(289/87-09-06): Steam Generator' Level 0 Instrument Static Shif ]
Additional.informatio'n was' addressed in Section .6- (0 pen) Violation (289/87-09-12). Failure to Survey Letdown f Pre-Filter Cubicle Additional'information was' addressed in Section .7'~(0 pen) Unresolved. Item (289/87-09-15): Licensee to Improve l Content of Licensee Event Report "
Additional.information was. addressed in Section' ~
8.8 (Closed) Unresolved Item (289/87-11-02): NRC Staff to Review Licensee Corporate Reorganizatio The. previous inspection report documented a licensee corporate level reorganization change that was not reviewed and approved by NRC _
staff. ' Technical Specification Change Request (TSCR) No. 172, dated
" Jure 18, 1987, was submitted by the licensee to address-the chang NRC staff issued Licensee Amendment No. 132, dated September 1, 1987, to approve this change. The licensee representatives acknowledged that the change should have been submitted prior to implementatio .t The NRC encourages licensee to take corrective action as a result of self-review and self-identification of problems. The reorganization-was, in part, a corrective-action to enhance overall safety' performance of the licensee's management structure. .Accordingly, no enforcement ;
action'will be taken on this matter and the inspector considers this item !
close . Exit Interview The inspectors discussed the inspection scope and findings with licensee management at a final exit interview conducted October 2,
- 198 Senior licensee personnel attending the final exit meeting included the following:
J. Colitz, Plant Engineering Director, TMI-1
' Hukill, Director, TMI-1 C. Incorvati,-Audits Supervisor, TMI-1 ]
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S. Otto, TMI-1 Licensing M. Ross, Director, Plant Operations, TMI-1 ]
q D. Shovlin, Manager, Plant Maintenance, TMI-1 E__ __ _ _- __ -
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T. Moslak, Resident Inspector, TMI-2, also. attended the meetin The inspection results as discussed at the meeting are summarized in the' cover page of the inspection report. Licensee representatives {
did not indicate that any of the subjects discussed contained' l proprietary or safeguards informatio i Unresolved Items are matters about which more-information is required in order to ascertain whether they are acceptable, violations, or deviation j Unresolved items discussed during the exit meeting are addressed in "
. Sections 2 and t Inspector Follow Items are significant open issues warranting follow-up review by the inspector to properly disposition the matter. An inspector j follow-up. item was addressed in Section j
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ATTACHMENT 1 NRC INSPECTION REPORT NO. 50-289/87-17 ACTIVITIES REVIEWED P_lant Operations Control room operations during regular and back shift hours, including i frequent observation'of activities in progress and periodic reviews of selected sections of the shift foreman's log and control room operator's log and selected sections of other control room daily logs i
Areas outside the control room '
Selected licensee planning meetings 9/16/87 - licensee post-trip review process; AP 1063 9/17/87 - selected reactor startup arid power ascension activities; OP 1102-1, " Plant Heatup to 525 F," and OP 1102-4, " Power Operation."
9/25/87 - STP 1-87-040, " Control Building Loss of Ventilation Testing" During this inspection period, the inspectors conducted direct inspec-tions during the following back shift hour Date Time i,
9/5/87 10:00 a.m. - 12:00 {
'9/16/87 6:15 p.m. - 10:15 )
9/17/87 6:00 a.m. - 7:00 /20/87 12:00 a.m. = 3:00 !
Maintenance
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9/8-10/87 - reactor building emergency cooling river water pump, RR-P-1B, j outage and maintenance
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9/22-25/87 - decay heat (DH) pump "B" and associated piping mainter:ance activities Surveillance 9/21-25/87 - HSPS surveillance activities per Surveillance Procedure (SP)
1303-11.37 B/C/0, "HSPS Channel II, III, IV Level and Pressure Checks" l 9/24/87 - DH pump "1B" inservice testing (IST) surveillance, SP 1300-38 Reactor Coolant System (RCS) Leak Rate l
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The inspector selectively reviewed RCS leak rate data for the past
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inspection period. The inspector independently calculated certain RCS leak rate data reviewed using licensee input data and a generic NRC
" BASIC" computer program "RCSLK9" as specified in NUREG 110 Licensee (L) and NRC (N) data are tabulated belo TABLE ,
RCS LEAK RATE DATA All Values GPM DATE/ TIME (NUREG 1107) CORRECTED DURATION- L N Ng L G "G U U 9/8/87 .2934 .29 .02 .0844 .0849
'09:11:06 2 Hours -
9/8/87 .2234 .22 .02 .0844 .0790 15:09:34 2 Hours 9/15/87 .7218 .72 09 .0144 .0174 17:07:07 2 Hours 9/16/87 .5812 .59 .05 .0544 .0552 01:01:08 2 Hours 9/21/87 .0146 .02 .16 .0556 .0594 02:16:56 2 Hours
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9/26/87 .0424 .04 .12 .0156 .0190 01:40:53 2 Hours 9/26/87 .0515 .05 .13 .0256 .0271 09:02:50 2 Hours 9/30/87 .1732 .17 .15 .0456 .0416 00:19:00 2 Hours 9/30/8 .0883 .09 .11 .0056 .0028 09:03:12 2 Hours G = Identified gross leakage U = Unidentified leakage L = Licensee calculated N - NRC calculated I
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i Columns 2 and 3, 5 and 6 correlate + 0.2 gpm in accordance with NUREG 1107. (N is corrected by adding OT1044 gpm. to the NUREG 1107 N due to u u total purge flow through the No. 3 seal from RCP' .
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