IR 05000289/1987013
| ML20235W400 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 09/24/1987 |
| From: | Blough A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20235W398 | List: |
| References | |
| REF-GTECI-A-47, REF-GTECI-SY, RTR-NUREG-0737, RTR-NUREG-737, TASK-2.E.1.2, TASK-A-47, TASK-OR, TASK-TM 50-289-87-13, IEB-79-02, IEB-79-2, IEIN-86-053, IEIN-86-53, NUDOCS 8710160204 | |
| Download: ML20235W400 (36) | |
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U. S. NUCLEAR REGULATORY COMMISSION j
REGION I
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Docket / Report No. 50-289/87-13 License:
DRP-50 Licensee:
GPU Nuclear Corporation i
P. O. Box 480 Middletown, Pennsylvania 17057 Facility:
Three Mile Island Nuclear Station, Unit 1 Location:
Middletown, Pennsylvania i
Dates:
July 9 - September 4, 1987 Inspectors:
R. Conte, Senior Resident Inspector (TMI-1)
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D. Johnson, Resident Inspector (TMI-1)
C. Myers, Resident Inspector (Rancho Seco), Region V S. Peleschak, Reactor Engineer, Region I (RI)
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f-W. Russell, Region I Administrator, RI D. Wallace, Reactor Engineer, RI C. Woodard, Reactor Engineer, RI Approved by:
N')U i3 h 4/
f7 A11en' R. Brou)h, ChidY Date r
Reactor Sec'tidn No. 1A Division of Reactor Projects Inspection Summary:
The NRC resident staff and Region-based inspectors conducted safety inspections (253 hours0.00293 days <br />0.0703 hours <br />4.183201e-4 weeks <br />9.62665e-5 months <br />) of power operations, focusing on performance in operations, which included maintenance and surveillance areas.
Items reviewed in the plant operations area were plant material condition, reactor building bulk tempera-ture, and intermediate building high temperature.
Other items reviewed included nuclear services closed cooling water system operability, support / base plate anchor bolt installation, letdown cooler outage work (MS-V-4A and NI-I repair), overall maintenance program, reactor coolant inventory trending system installation, worker attentiveness, and licensee action on previous inspection findings.
8710160204 871006 PDR ADOCK 05000289 G
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Inspection Results:
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Licensee management continued their. detailed attention and involvement in the functional areas reviewed.
Response to the housekeeping and plant material condition issues identified during the regional administrator inspection was thorough and corrective action was completed in a timely manner. Also, licensee corrective action taken in response to the licensee-identified improperly installed anchor bolts in the chilled water system was adequate.
Final administrative closecut of several work items from the 6R refueling outage were noted to be incomplete and/or delayed for a long period of time.
The licensee' acknowledged this fact and agreed to review the final processing of work items to assure more expeditious processing.
Licensee response to the worker attentiveness to duty issue was positive.
The licensee was adequately prepared to deal with this issue.
There were no major concerns with worker / operator alertness and attentiveness to duty.
A review of the maintenance program revealed that maintenance activities were being accomplished in a satisfactory manner. A separate review of the installation of the Reactor Coolant Inventory Trending System (RCITS)
also' indicated satisfactory installation, testing, and operation of this newly implemented portion of the inadequate core cooling monitoring system.
Unresolved items in the areas of anchor bolt installation for chilled water system Heating, Ventilating and Air Conditioning (HVAC) supports and adequacy of review for replacement-in-kind parts were identified.
No violations of regulatory requirements were observed.
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I DETAIf.S
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1, Introduction and Overview
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1.1 NRC Staff Activities
The overall purpose of.this inspection w4s to assess licensie activities during the power operations mode as they related to reactor safdy and radiation protection. Within each area, the inspectors. documented the speed fic purpose of the area under review, acceptance criteria and scope of npections, along with appropriate findings / conclusions. On;a sampling basis, the inspector made this assessment by reviewing informa-tion through actual observation of licensee activities, interviews with licensee personnel, measurement < f radiation Tsveh, or independent calculation and selective review of listed applicable documents.
Also, during this' peri /d, a resident inspector participated in a manage-ment meeting between R('itt I and the licensee on July 13y 1987, concern-ing a licensed operator C,teping allegation at TMir2 (see Section 7). A
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resident inspector also participated'in an emergency planning inspection (NRC Inspection Report No. 50-289/ 87-14), a. security inspection (NRC Inspection Report No. 50-289/ 87-15), and an inspection at the Saxton Nuclear Experimental Station (NRC Inspection Report No. 50-140/87-01).
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1.2 Licensee Activities
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During this period, the licensee operated the plant at fu?1 -power,
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except for a period of about eight hours on September 2, 1937, at 76 r-percent power.
This power reduction was necessary becwse e. main feedwater pump was removed from service to correct a high vibration problem at the turbine-to pump shaft coupling.
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2, 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:
review of selected plant parameters for abnormal trends;
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plant status from a maintenance / modification viewpoint, including
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plant housekeeping and fire protection measures;
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control of ongoing and special evolutions, including control room personnel awareness of these evolutions; control of documents, including logkeeping practices;
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implementation of.the radiological controls program on area /
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-boundary control; and,
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L implementation of the security plan, including access control, l '
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boundary integrity, and badging practices.
The inspectors focused on the specific areas listed in Attachment 1.
As a result of this review, the inspectors reviewed specific evolutions in more detail as noted below.
2,2 Findings / Conclusions:
y 2.2.1 Material Condition of the Plant n
y On ' July 16. 1987, the Region I Administrator conducted an inspection on a. sampling basis of the material conditions of plant spaces and l equipment. Attachment 2 is a list of discrepancies identified. At the conclusion of the inspection, the Region I Administrator exited
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with the Director of TMI-1.
Housekeeping was considered abo,'e average,'but material deficiencies need to be corrected.
System / component labelling could be more formal.
There appears to be a good valve maintenance program in place in safety-related areas. A few leaks were noted in the secondary plant with some packing glands not having much room for tightening. Many operating linkages on components were in good shape, free of dirt to prevent malfuncth,ning.
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On a sampling basis, the resident inspector verified the corrections of the discrepancies or that the licensee had taken action to assure q
required repairs would be accomplished. All areas where housekeep-J ing discrepancies were noted were observed by the inspector during
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routine. tours of the auxiliary bui'if ng, fuel handling building, and control tower. The specific housekeep'ng item: had been corrected to a satisfactory condition.
In addition, the inspector noted that, in general, other spaces had been :aaintained at an acceptable level of cleanliness.
The inspector also witnessed the accomplishnsent of the recair of the
"B" emergency diesel generator (EDG) speed changer cable, which was observed to have insulation that was frayed.
The licensee installed new cabling with a new flexible cloth type conduit ar.d used plastic wire ties to position the cable out of the path of the fuel rack mechanism.
The completed repair was examined by the inspector and its. appeared that the new installation was satisfactory.
In addi-tion,, the same cable on the "A" EDG was tied back from the fuel rack l
leu rs to prevent contact. The "A" EDG cable was evaluated by the
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licensee as not needing repair.
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1 The ' inspector also observed that the air in-take louvers for. various L
pumps in the closed cooling system pump room on the 305 foot elevation of the auxiliary building had been cleaned.
The inspector also observed other pumps in the plant with electric motor _ drives and verified accep-table levels of cleanliness.
The inspector questioned licensee personnel on the acceptability of-the placing of plywood spacers between individual battery cells for the vital batteries.
The licensee contacted their vendor and had
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-them-provide justification on the acceptability of existing spacing, i
As documented by the licensee, the vender stated that the maximum electrolyte temperature rise was estimau n to be 5-6 F during any normal and load discharge testing and this temperature rise was too small to cause significant internal or external thermal expansion.
A tight fit between the battery cell and the racks was recommended by the vendor.
The leak in the "B" decay heat (DH) vault on the vent valve for the pump casing was evaluated by the licensee as needing repair. A job ticket was written and the repair will be accomplished during the next outage when the DH system is down for repairs.
2.2.2 Reactor Building Bulk Temperatures Based 'on a problem identified at Arkansas Nuclear One, Unit 1 (ANO-1), the inspector reviewed licensee activities to assure that actual reactor building (RB) ambient temperatures were within the design basis as described in the Final Safety Analysis Report (FSAR).
In add 4 tion to discussions with cognizant licensee person-nel, the inspector monitored actual RB ambient temperatures and
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reviewed the following documents:
FSAR, Sections 5.2.1.2.9, 5.2.1.2.10, and Table 6.6-8;
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Safety Evaluation Report (SER), dated May 28, 1978, for License
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Amendment No. 41; Surveillance Procedure (SP) 1301-1, Revision 68, " Shift and
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Daily Checks," Enclosure 6, Section 6.7, Selected Data for July and August 1987; and, Technical Specification (TS) 3.17 and 4.20;
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i The inspector found that licensee monitoring and data collection is in accordance with applicable TS conditions and frequency require-
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ments. Daily temperatures are taken; and, if any temperature point is greater than 130 F at locations greater than 320 foot elevation, the average is computed to assure that bulk temperature is less than 130 F.
This is similarly done below the 320 foot elevation profile, except a 120 F limit applies.
The inspector noted that the worst case recorded temperature on the SP data (highest temperature l
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profile) for July and August 1987 occurred on July 26, 1987, when the southeast wall elevation (352 foot) was 133 F.
The average bulk temperature for elevations above 320 feet (17 temperature monitor points) was 120.94 F and the average bulk temperature for all 24 monitor points was 114.7 F.
The inspector noted that these average temperatures were in accor-
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dance with TS and FSAR, Section 5, limits; but they exceeded the assumed bulk initial temperature for a loss of coolant accident (110
F). The NRR staff's SER for Technical Specification Amendment (TSA)
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No. 41 addressed this situation, when the then new TS were issued, and imposed the 120 F and 130 F limits.
The SER stated that the staff did a sensitivity study on post-LOCA conditions for 90, 130, 150, and 170 F initial bulk RB temperature..The study found that none of the calculated peak accident pressures exceeded the contain-
.i ment design pressure of 55 psig.
In addition, the study indicated that calculated peak accident temperature does not exceed design temperature of 281 F for those cases where the initial temperature in the RB is less than 150 F.
Based on the above, the inspector j
concluded that'the licensee was operating within the design basis for the plant with respect to normal operating RB bulk temperatures
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related to post-LOCA design basis parameters.
Also, the basis for the TS limit appears to be related to structural integrity of the RB (FSAR, Section 5.2.1.2.9 and 10).
The inspector also questioned the licensee representatives on whether or not the existing RB temperature profile data for the
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summer months was consistent with temperatures assumed for aging
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analysis-in the environmental qualification (EQ) area. A licensee representative reported that a 100% re-review was conducted for EQ
equipment in the RB with no problems identified with respect to
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being outside (nonconservatively) temperature assumptions for aging analysis.
The inspector had no further questions in this area.
2.2.3 Intermediate Building High Temperature During this period, on July 21, 1987, a security guard collapsed
'from heat stress when making his round on the top elevation (355 foot) of the intermediate building (IB). The IB houses, in part, the main steam system where hot and humid conditions normally exist
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due to a significant heat load inside the building.
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temperature when the security guard succumbed was slightly in excess
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of 150 F.
Fortunately, the guard was able to contact other person-nel for help.
The licensee subsequently restricted access to this area when temperatures are in excess of 130 F.
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Since_ security procedures require this area to be toured, security personnel documented the failure to comply in an internal security incident report on a shift basis.
In consultation with Region I security inspectors, the inspector concurred with the licensee's
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action in the interest of personnel safety.
The root cause of the problem appears to be poor ventilation for that area of the IB.-
The licensee reports that no ventilation design changes are planned, but they acknowledged that other mea-sures such as main steam line re-insulation to reduce heat load in the building are being examined.
Restrictions will continue to be
imposed if temperatures on that elevation go above 130 F.
Security will document the situations and temperatures when the restrictions
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are imposed.
The inspector had no additional questions on this matter.
2.3 Plant Operations Summary The material condition.of the plant remains quite good. No major equipment was out of service at the close of the inspection.
Licensee corrective actions for problems in the plant appear to be adequate.
No procedure adherence problems with respect to operating procedures were noted.
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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 this report.
Specific areas of review are listed in Attachment 1.
A more detailed review of equipment operability was also addressed
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3.2 Findings / Conclusions 3.2.1 Support / Base Plate Anchor Bolt Installation On July 9,1987, the licensee verbally reported the improper installation of two support / base plates on the piping of the chilled water (CW) system for the control building (and control room) ventilation system.
The supports are located at the 330 foot elevation of the control building (CB) on adjacent supply and return lines along a vertical run of piping.
On each support, one of four bolts was not anchored to the wall.
Instead, the bolt heads were welded at the back of the base plate, giving the appearance of proper installation.
For each of the support plates, only three of the four required holes were drilled in the wal v
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The licensee 1found,this during inservice inspection of the supports.
The' licensee ve~rified proper installation of the remaining three bolts.on each support'by torque checks and. calculated a' safety.
factor:of'.approximately.2.4 for the-existing installation.
The licensee wrote a " material nonconformance: report"'on the matter and documenteditheir'engi.neering' evaluation to. resolve the issue.
In accordance with NRC staff guidance.on the-matter, they had planned
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to make repairs'during"the.next'. plant shutdown but, during the week of August 3, 1987, they decided.to make repairs.
They traced the faulty workmanship on 'one support to a welder who was at the site for a: short time in the early.1980's.
The other support apparently dates back-to new construction.
The licensee begancto. evaluate any additional corrective actions with respect-to.
NRC Bulletin 79-02. requirements.
On. August ~ 10, 1987, the licensee verbally reported additional information on the remaining anchors-for each of the supports, which they: determined during the above-noted repair for the subject y
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supports. - Apparently,.- because of "rebar" interference, the cement anchors'(shells) were cut from their normal lengths.
For one-support,:the: licensee found the shells cut.short in the remaining three anchor holes.
For the other. support, the licensee found one shell cut short.in the remaining three anchor holes.
Torque checks taken previous to the repair work were satisfactory for the remain -
ing three anchorfsites in each support.
The' licensee issued another internal material nonconforma_nce. report.
Based on their data, they have tentatively. concluded that, even with the cut shells, the supports would have functioned during a safe.
shutdown earthquake. The licensee reviewed this matter for report-ability and concluded that it was not reportable for the above reason.
Additional measures planned by the licensee included:
(1) identify-
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apparently.had only the CB' Heating, Ventilating and Air Conditioning
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(HVAC) contract (apparently Category I supports for HVAC ducting)
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and (2) reviewing IE Bulletin 79-02 records on the subject CW system for indications of similar deficient anchors.
L In a conference call with Region I on August 17, 1987, the licensee
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E reported that one contractor was responsible for CB CW piping installation and records were not clear, as yet, on who installed
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l-Seismic Category I ducting for the CB HVAC.
Bulletin 79-02 records
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revealed the existence of short anchor shells based on sampling mea-i l
surements. The records also indicated acceptable testing at 40
percent of ultimate yield strength.
For the CW system's 400 bolts, i
200 bolts were inspected with tension testing on 100 bolts (at least l-one bolt per sampled base plate) (with shells short and long). Of l
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the 100, sixteen had shortened shells and were satisfactorily
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tested. The licensee reported completing the records review by September'1; 1987, and closing out the latest material nonconfor-
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mance report (MNCR.No. 129-87) on the CW system. The licensee reported that the HVAC. ducting issue would be resolved separately.
This is unresolved pending completion of ifcensee action as noted above and subsequent NRC Region I review (289/87-13-01).
3.2.2 Letdown Cooler Outage Work The inspector reviewed the documentation of various job tickets (JT's) for work completed during the letdown cooler replacement
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outage of June 1987.
These JT's are listed in Attachment 1.
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particular. Jobs were of interest.
One was the replacement of a
. faulty relay in the Integrated Control System (ICS) circuitry to
. operate MS-V4-A, atmospheric dump valve for the "A" steam generator.
The relay failure caused the valve to fail partially open (see Inspection Report No. 50-289/87-10).
The other job was the replace-ment of cables for NI-1, one of two source range channels for nuclear instrumentation.
The inspector referred the following comments to licensee management.
MS-V-4A (JT CN 221)
The JT was not written until July 1, 1987, but it did indicate
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work was accomplished on June 27, 1987. Apparent controls in place for the repair was the use of the lifted leads and jumper l
administrative procedure of which the records were attached to I
the JT.
There was sufficient evidence of proper interface I
between maintenance and operations personnel for the work, i
However, it appeared that the documented release and restora-i tion to operable status of the equipment was not completed as
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provided for by the maintenance control system through the use of the JT.
NI-1(JTCM914,ChangeModificationNo.0821M1
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As of September 2, 1987, the operability section was not signed
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off for this JT. Other evidence, such as surveillance tests and outage prerequisite lists, document a determination that
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NI-1 was operable at the time of startup in late June 1987; however, the JT record was obviously unclear on this matter.
i As documented in the JT, a reflectometry trace was not obtained
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or listed on the change modification package attached to the
JT. A note on the documentation indicated that it would be
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l accomplished at a later date. The trace is used to obtain baseline electronic characteristics for the cable for future troubleshooting activities.
It was also unclear how that outstanding action was being tracked for follow-up.
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9 The cable used was not exactly the same as that originally
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installed.
The change modification package indicated that the new cable was a replacement-in-kind.
Based on a verbal discus-sion.with the cognizant engineer, the inspector concluded that the engineer did a mental evaluation of cable electronic characteristic similarity, considering original design basis documentation and GPUN specification for the purchase from the original vendor.
The. licensee committed to clarify the change modification justification to at least identify applicable design / specification documents for the NI-2 cable replacement,-
which is expected to be completed in the next refueling outage.
This item is unresolved pending review of the NI-2 cable replacement (289/87-13-02).
This review would be applicable to the NI-I cable replacement.
The inspector also noted that these and other JT's were in the
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' instrument shop waiting for Nuclear Power Plant Reliability Data System (NPRDS) reporting.
The inspector questioned licensee management on the adequacy of this practice of keeping
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completed original JT records stored in apparently unprotected
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files.
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Licensee management agreed to review the above-noted items with
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respect to proper implementation of related administrative
controls. The inspector had no additional comments on these matters.
3.3 Operability Summary
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For the equipment reviewed, the inspector verified operability and/or satisfactory completion of test procedures.
Apparent job ticket documentation problems were addressed in Section 5 of this report.
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Nuclear Services Closed Cycle Cooling System Review
4.1 Review
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The inspector reviewed Operating Procedure (0P) 1104-11, Revision
J 25, dated December 3, 1986, " Nuclear Service Closed Cooling Water System," to ascertain whether it is in accordance with regulatory requirements and whether its technical adequacy is consistent with desired actions and modes of operation. As part of this review, the inspector also examined the following documents:
Surveillance Procedure (SP) 1300-4E, Revision 8, dated May 13, j
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1987, "NSCCW Pump and Valve Functional Test During Refueling;"
j SP 1300-3J, Revision 15, dated April 9, 1987, "NSCCW Pump and
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Valve Functional Test;"
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GAI Drawing'C-302-610, Revision 31, " Nuclear Service Closed
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Cycle Cooling Water;" and, I
ANSI N18.7-1976, " Administrative Controls and Quality Assurance
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for the Operational Phase of Nuclear Power Plants."
In addition to the above, the inspector toured accessible ~ areas of the system. A valve lineup of approxin,ately 45 valves was ccmpleted during the week of September 10, 1987.
4.2 Finding The inspector' determined that the procedure's stepwise instructions I
were compatible with checklist information and provisions for signoffs were evident.
Various precautions and. notes concerning equipment.and adininistrative operability requirements and appropri-ate technical specification requirements were also incorporated into
'l the procedure.. The valve checklist and the Piping and Instrument Diagram (P&ID) were compatible and agreed with each other.
During the tour of. the "C" and "B" NSCCW pung motor breaker's interiors, the inspector found them to have acceptable cleani'iness.
The inspector found that the correlation of instrumentation to
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calibration records was cumbersome but appropriate records were found to support calibration of the selected instrumentation.
This item was noted to the licensee.
4.3 Conclusion The inspector determined that OP 1104-11 is adequate to control safety-related operations within applicable regulatory requirements.
If called upon, this procedure can be used by operations to remove decay heat from the core after the plant is shut down.
No adverse conditions were found that would affect plant safety.
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Maintenance Program Review 5.1 Acceptance Criteria / Scope of Review An inspection was conducted to verify that the licensee has estab-lished a program to ensure corrective and preventive maintenance is being properly performed on plant components and systems.
The following references constituted the basic requirements:
10 CFR 50, Appendix B, Quality Assurance Criteria for Nuclear
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Power Plants; Regulatory (RG) 1.33-1978, Quality Assurance Program Require-
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.for the Operational' Phase of Nuclear Power Plants.
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5.2 Programmatic Review i
The inspector reviewed applicable administrative procedures and held discussions with Maintenance and Quality Control personnel in order
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to verify that the following attributes have been adequately incor-porated into the maintenance program; written procedures have been established for initiating requests for
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corrective maintenance; s
critaria and responsibilities have been established for review
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and approval of maintenance requests; criteria and respon:ibilities have been established in order to
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categorize activities as safety related or non-safety related; criteria and responsibilities have been designated for' performance
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of inspections.of maintenance activities; j
methods and responsibilities have been designated for performing
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functional testing following maintenance work; procedures have been established to specify that the operations
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staff will release equipment and systems to maintenance; l
equipment that is environmentally qualified is identified as
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such prior to maintenance and controls exist to re-establish j
the environmental qualification of these components;
1 procedures and responsibilities have been established for j
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determining when independent verification of component / system j
is required; j
a written preventive maintenance program for safety-related i
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structures, systems, and components has been established;
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an overall schedule for preventive maintenance has been estab-lished; l
responsibilities and methods have been established for determining
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preventive maintenance frequencies; j
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and systems;
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cleanliness classification for plant systems have been established;
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administrative controls and responsibilities for general housekeeping
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have been established.
The inspector also reviewed approximately fifteen completed corrective maintenance job tickets and fifteen completed preventive maintenance job
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tickets $n order to verify that maintenance program requirements were L
being implemented correctly.
In particular, the following attributes were verified on each job ticket as applicable.
j Job scope section of job ticket is adequately completed, j
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Materials used in the performance of the work are recorded as
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required, Necessary prerequisites have been. completed prior to commencing
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the actual work,
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Worg performed section of the job ticket is completed.
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Completion of job has been verified by qualified personnel,
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i Appropriate acceptance criteria is specified for post-maintenance
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testing.
Environmental Qualification status of equipment is indicated
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and controlled in accordance with station procedures.
Applicable maintenance procedures were attached to job tickets when
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required.
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~The inspector also reviewed selected minor maintenance and blanket job ticket activities -- approximately 34 and 10, respectively.
l These activities are typically utilized for maintenance of equipmtnt
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that is not of a complex nature and can usually be performed by i
qualified personnel without a detailed procedure.
In particular, l
the inspector verified the following criteria:
q description of the problem was adequately described on the
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minor maintenance form or blanket job ticket work form;
resolution or work performed sections were adequately completed;
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post-maintenance testing has been adequately addressed and
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documented; i
permission to start work has been granted by qualified personnel; j
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and,
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l maintenance activities are reviewed by maintenance foreman and
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shift foreman,
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'13 5.3 Findings The inspector determined that the requirements referenced in paragraph 5.1 and the attributes listed in paragraph 5.2 of this report have been ade-quately incorporated into the maintenance program. During his review of maintenance records, the inspector generally found all required informa-tion to be adequately documented and that records of post-maintenance activities were retrievable as part 'of plant history.
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In one instance, the inspector questioned preventive maintenance personnel concerning the adequacy of the description of maintenance I
performed on EF-V-308, Emergency Feed Control Valve. The inspector's inquiry was later resolved by the licensee, who clarified the description of the work performed on the valve in question. The inspector noted to the licensee the importance of documenting exactly what work has been performed. The licensee agreed with the inspector's concern and initiated additional documentation for the work in question.
The inspector had no further questions or concerns in this area, since this was the only item of its type identified.
The inspector's review of blanket job tickets and minor maintenance work indicated that these activities were controlled in accordance with administrative procedures and regulatory requirements.
In
.several instances, the inspector observed cases where work being performed under a blanket job ticket was cancelled and reinitiated under. a regular job ticket because the scope of work was perceived as exceeding the intent of a blanket job ticket.
The inspector verified-that the activities reviewed did not constitute work activities that would be considered beyond the scope of minor maintenance or blanket job tickets.
During the review of minor maintenance post-maintenance testing, the inspector observed that leak tightness appeared to be the only acceptance criteria for minor maintenance performed to stop or reduce vai n packing leakage.
The licensee was able to demonstrate that the motor-operated valves in question (DH-V-5A/B) had been cycled since their packing had been tightened, but maintenance documentation did not indicate that proper functioning of the valves
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was necessarily designated as an acceptance criteria.
The inspector discussed this issue with management, stressing the importance of evaluating whether cycling or stroke time testing is necessary even for valves being worked under " minor maintenance" controls.
Manage-i ment agreed with the inspector's concern and indicated that the issue would be further reviewed.
The inspector had no further questions.
No violations or deviations were identified.
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The. inspector reviewed Quality Assurance Audit No. S-TMI-86-01, TMI
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Maintenance, March.1986, and conducted interviews with Quality
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Assurance (QA);and Quality Control (QC) personnel to ascertain the role and effectiveness of QA/QC in maintenance.
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Audit S-TMI-86-01 closed previous QA findings in the area of mainte-nance and initiated corrective action for several new findings.
The inspector verified-the adequacy of corrective actions for those items that were closed by S-TMI-86-01.
The inspector also verified that QA was tracking the corrective action for new items identified in S-TMI-86-01.
The inspector determined through discussions with QA/QC personnel and
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procedure reviews that QC hold points are adequately incorporated into l
maintenance procedures using standarf'/.ed methodologies where possible, a
In addition, the inspector verified that the QC group rely on their own
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weekly review of the plant's work schedule to ensure adequate QC oversight
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in the area of maintenance.
j 5.5 Maintenance Program Summary
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The maintenance program is well established; and, in general, it is properly implemented.
Licensee personnel were responsive to inspector concerns.
In light of the reviews noted in thir section and Section 4 of this report, there appeared to be a problem with the adequacy and timeliness with respect to maintenance documentation of the testing and restoration to service phases for certain equipment, including safety-related equip-ment.
In all cases, the licensee was able to demonstrate satisfactory results for these particular phases for the selected equipment, primarily because other management control systems apparently assured proper testing and restoration to service, such as the start-up from outage prerequisite list.
The inspectors' concern narrowed down to that of an administrative l
nature on how the licensee assures proper implementation of related administrative controls for the proper documentation of testing and
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restoration to service with respect to specific job tickets.
The timeliness of the records getting into proper storage facilities also appears to be a problem. The resident inspectors will routinely
follow-up on this area.
No violations or deviations were identified.
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6.
Instrumentation for Detection of Inadequate Core Cooling Modifications -
6.1 Background The NRC Order for Modification of License for TMI-1, dated December 10, 1982, required that the licensee provide additional instrumentation to detect inadequate core cooling during post-accident conditions.
NUREG 0737, Clarification of TMI Action Plan Requirements,Section II.F.2,
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provided specific requirements and guidelines for this instrumentation.
l Accordingly, the licensee has designated, installed, and made operational
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Inadequate Core Cooling Instrumentation (ICCI), which is composed of three basic sub-systems as follows.
Subcooling Margin Monitor - Indicates the approach to ICC by
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showing both saturation and superheat conditions when the reactor coolant pumps are on or when natural circulation can be verified.
Core Exit Thermocouple - For determining both the existence of
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ICC and the trends of recovery action.
Reactor Coolant Inventory Trending System (RCITS) = Indicates
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the trend of water inventory in the Reactor Coolant System (RCS) above the core in the quiescent state and the trend in void fractions with the reactor coolant pumps (RCP's).
An inspection was made during this period of the licensee's plant modifications to include the RCITS.
6.2 Documents Reviewed l
The inspector reviewed pertinent work and quality assurance records for l
the design, procurement, qualification, construction, installation, and tests of the RCITS to assure that the records reflect the implementation of systems consistent with NRC requirements and licensee commitments.
Documents reviewed included the following.
Speci fications
SDD-662C Division I System Description for TMI-1 RCITS, l
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Revision 11, dated June 23, 1987
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SDD-TI-662C Division II System Description for TMI-1 RC1
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Trending System, Revision 2, dated March 17, 1987
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TI-IS-412023-Installation Specification for TMI-1 RCITS,
001 Revision 0, dated March 16, 1984 l
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TI-IS-412023-Installation Specification for TMI-1 RCITS
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006 Impulse Line Modification, Revision 0, dated December 18, 1986 Drawings
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M0002 Flow Diagram - Reactor Coolant System Vent to
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Containment, Revision 1, dated December 22, 1986
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M0025 RCITS - Isometric, Hot Leg "B" and Reactor Vessel, Reactor Building, Revision 2, dated December 21, 1986
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M0026 RCITS - Isometric, Decay Heat and Drain Down Reactor Building, Revision 3, dated July 23, 1984 M0027 RCITS - Isometric, Hot Leg "A" and Reactor
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Vessel, Reactor Building, Revision 2, dated December 24, 1986 M0003 RCITS Flow Diagram, Decay Heat Removal, Revision 0,
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dated March 13, 1984 M0021 RCITS - Isometric, Reactor Vessel Vent Piping at
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5ervice Structure, Revision 1, dated December 24, 1986 M0035 RCITS - Reactor Vessel Impulse Line
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Modification, Revision 1, dated January 18, 1987 l
QA/QC Inspection Reports CS-33361-84 Final Acceptance Inspection of the RCITS
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Modification (Civil / Structural / Welding), December 26, 1986 EL-13273-84 Final Acceptance of the RCITS Modification (Electrical), December 27, 1984 IC-23209-84 Final Acceptance Inspection of the RCITS Modification (Instrumentation), December 27, 1984 ME-03123-87 Hydro of RCITS Tubing and Welds on Reactor Head, January 27, 1987 ME-03214-87 Turnover Inspection for RCITS Reactor Vessel Impulse Line Modification, February 18, 1987
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CS-33338-87 Civil, Structural Work Items Turnover RCITS,
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CS-33360-84 Civil, Structural Work Items Turnover RCITS,
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December 14, 1984 Turnover Documents A25F-30023-S RCITS Release to Startup and Test Notification, February 24, 1987 A25F-30023-T RCITS Turnover Notification to Plant / Station, April 14, 1987
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6.3 RCITS Installation Modification l
6.3.1 Reactor and Hot Leg Water Level Measurements e
The reactor / hot leg water level portion of the RCITS consists of two identical independent instrument loops to measure hot leg water
inventory and two identical instrument loops to measure reactor
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vessel water inventory. The "A" instrument loops measure the l
primary coolant water inventory in the "A" hot leg and the reactor
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vessel.
Similarly the "B" instrument loops measure the primary
coolant water inventory in the "B" hot leg and the reactor vessel.
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Inspection of this system was made by a review of the procurement and I
installation documents listed in paragraph 6.2 and by physical inspection.
The water level instruments are Foxboro Model N-E13DH differential pressure units that were environmentally qualified in accordance with 10 CFR 50.49, paragraph (f) (2), which permits qualification by type testing a similar unit with supporting analysis to show that the qualification covers the actual units installed. A review of the vendor and licensee j
environmental and seismic qualification test data and documentation did j
not disclose any reasons to question the qualifications.
However:
periodic preventive maintenance of the instruments is required to maintain
this qualified status.
The inspector confirmed that this instrumentation
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is included in the licensee's preventive maintenance program.
Criterion 3 of Appendix B to NUREG 0737 states that "the instrumentation should be energized from Station Class IE power sources." A review by the inspector disclosed that Loop "A" and Loop "B" are powered from separate Class 1E instrumentation vital buses.
Inspection within signal conditioning cabinets "A" and "B" for these
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two trains of water level instrumentation signals disclosed that
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proper electrical separation is maintained between cable trays, cables, and wiring in accordance with licensee commitments contained in electrical and instrumentation specifications, SP-5615 and i
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l The RCITS water level signals.from cabinets "A" and "B" are routed l
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The signals are l
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multiplexed and all four water levels can be viewed on a computer i
screen trend plot at one time.
The reactor control room operator was familiar with the instrumentation the water level trends displays and printouts.
i Since the only visual readout / printout of the water level signals f
for this system is on a computer (through a single multiplex unit)
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in the control room, the NRC staff had questioned single failure
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backup capability.
The inspector found that the licensee has f
developed and implemented Operating Procedure (0P) 1103-1, Enclosure III, which provides instructions for obtaining water level instru-mentation voltage measurements at terminals in the signal condition-
ing cabinets "A" and "B", which can be equated to reactor vessel and
ho leg water levels.
Inspection within these cabinets disclosed
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that the operator could take these measurements with greater ease and wie less chance of shorting or grounding by taking the measure-ments from the signal module output measuring jacks, rather than from terminals on the terminal blocks.
The licensee plans to amend OP 1103-1 accordingly.
i 6.3.2 Re ttor Coolant Void Fraction Measurements The reactor coolant void fraction portion of the RCITS consists of four instrument loops to measure reactor coolant pump input electrical power.
The void fractions instrumentation system relates the electrical power required to drive these pumps to the amount of coolant being pumped i
and, thus, to the amount of coolant in the primary system and the amount j
of voids and, thereby, provides the void fraction calculation. The cal-
culation (void fraction) for each loop is displayed on a computer trend graphic screen with update signals each 30 seconds.
Printouts of the trends are made as required.
This system is described in detail by the system description listed in paragraph 6.2.
.l Inspection of this system was made by a review of the procurement and installation documents listed in paragraph 6.2 and by a physical inspection of portions of the system.
The four pump power signals for void fraction measurements are derived from existing watts f
transducers mounted in the RCP power monitor racks in the control
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building. These watts transducers currently feed RCP information into the RPS and are Class IE safety relatec Since the transducer l
circuits are safety related, it was necessary for the licensee to install qualified signal isolation devices in the pump monitor racks to provide appropriate signal isolations between the reactor protection circuits and the void fraction circuits.
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I Inspection review of licensee / vendor qualification test reports l
confirmed that the Rochester Model SC-1302 signal isolators used are properly qualified.
The Model SC-1302 has a 500 millisecond re-sponse line, which is considered satisfactory for the void fraction trend monitor with its 30-second periodic reacout scen but is not satisfactory for the RPS.
The' licensee had specified the Model SC-1300 with 50 millisecond response time for the RPS, However, the
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test report to the t!RC included in GPU letter 5211-87-2051,. dated March 9, 1987, shows that the isolation devices used to provide pump
information to both the void fraction and RPS instrumentation'are Model SC-1302. Due to the fact that the plant was at power and these signal isolators were not easily accessible, the inspector could not confirm what isolators are installed in both of the circuits.
Subsequently, the licensee reported that they verified that the proper isolation devices are installed in each of the circuits and that the sketch shown in their March 9,1987,. letter is in error.
7.
Operator / Worker Alertness and Attentiveness 7.1 Background On July 9,1987, the NRC staff at TMI-2 received an anonymous allegation that a licensed senior reactor operator (SRO) slept on watch and that apparently licensee management did not take appropri-ate corrective action.
That same week, licensee (TMI-2) management also reported their discovery of made-up sleeping area " nests" and/or card game playing areas primarily in the Unit 2 turbine building, an area of limited.use with respect to TMI-2 defueling activities.
The NRC staff, concurrent with the licensee, immediate-
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ly initiated a review of the matter; and, during that review, the potential applicability of the allegation to TMI-1 was considered.
As a result, the below-noted review, along with findings, was conducted by the NRC (TMI-1) resident office.
7.2 Acceptance Criteria / Scope of Review As a result of the above-noted allegation, the TMI-1 NRC resident inspectors rescheduled their July back shift and wakend coverage for July 10-12, 1987.
The purpose of the review was to verify the below-listed aspects.
Assess overall safety of operation during back shift hours
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cation (TS) Section 6.
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Shift manning was in accordance with 10 CFR 50.54(m).
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Licensed operators were mentally alert and attentive to duties i
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Other operators and workers were mentally alert and attentive to
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their. duties in accordance with Administrative Control Procedure
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1029, " Conduct of Operations."
Identify any (out-of-the-way) area of " sleeping nests" already
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'made up within the plant proper and in other remote areas such as the circulating water house, screenhouse, or TMI-1 Operations Support Facility (new administrative building).
The results of this inspection were discussed at a licensee management meeting' on July 13, 1987, for both units.
As a result of that meeting, an investigation was initiated at TMI-2 (licensee had initiated an investi-gation on July 10,1987).
The NRC staff decided that a programmatic inspection would be conducted at TMI-1, since there was no direct evidence
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of potential wrongdoing at TMI-1.
I By direction of regional management, the resident inspectors were to take a " fresh look" at the licensee program and practices related to operator / worker mental alertness and attentiveness to duties.
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program review examined licensee policy / procedures, procedural
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implementation, monitoring, and practices for disciplinary action for transgressions of the related policy. Applicable NRC rules referenced above were used. The following licensee documents and l
records were reviewed.
Administrative Procedure (AP) 1029, Revision 25, effective May 20,
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1987, " Conduct of Operations" Annual Memorandum (Serial No. 3000-86-14), dated June 4, 1986,
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" Command Responsibilities," from Director, TMI-1, and President, GPUN, to Shift Supervisors NRC letter, dated July 9,1987, from W. Russell, NRC, to P. Clark,
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GPUN Various licensee internal memoranda from 1981 to 1987 from H. Hukill
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on the "Of f-Shif t Tours by Management" Licensee internal memorandum, dated April 3, 1987, from H. Hukill,
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Director, GPUN, to P. Clark, President, GPM, "TMI-1 Status Report for the Period March 1, 1987, through March 31, 1987 l
Various off-site tour reports (approximately 21) for the period
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January 8, 1987, to July 2, 1987
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Various Quality Assurance Monitor Reports (approximately 45) fo ' the l
period March 14, 1987, to June 30, 1987 Selected records of disciplinary action on individuals for alleged
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sleeping 'or dereliction of duty (misconduct)
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7.3-Findings / Conclusions l
7.3.1 Weekend Coverage of July 10-12, 1987 Shift manning was in accordance with 10 CFR 50.54(m). At least one reactor operator (RO) and an SRO were in the control room with the plant above 200 F and at least two RO's and two SR0's were assigned
to the-shift. All, operators in the control room were awake and
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appeared to be mentally alert and attentive to their duties.
No unsafe practices were noted.
Similarly, during the tour of out-of-the-way areas, non-licensed l
operators and workers observed were awake and they apoeared to be i
performing their assigned duties.
No " sleeping nests' were found.
I Service Building offices, which have licensee management authorized
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" cots," were locked to prevent access. The " cots" are to be used j
when key managers work long hours in which it is inconvenient or j
inefficient to return home for sleep.
Licensee management reported I
these " cots" are infrequently used.
The licensee also reported that one of the cots was found in the overhead ceiling in the Service Building with no evidence of being used.
It was previously taken out of one of the manager's offices for space consideration, It was removed from the Service Building.
No unsafe practices were noted during the inspection of the area spaces.
7.3.2 Policy / Procedures The licensee's administrative controls for the conduct of operations (AP 1029) provides acceptable measures to assure operator and worker mental alertness and attentiveness to duties. These measures date back to 1981 and were, therefore, in place prior to the Peach Bottom
" sleeping operator" situation, which was made public in the spring of 1987. The AP applies to all who work at TMI-1, including those not under direct control of the Operations and Maintenance Director.
Several sections of the AP allude to worker responsibilities to be
" alert," " attentive," " vigilant," and avoiding unnecessary distrac-tions such as with televisions, radios, and unauthorized reading material.
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In particular, the policy is stated quite clearly as noted below:
"All on-duty personnel shall be physically fit and mentally alert. The Shift Supervisor / Foreman is responsible to ensure that all personnel, both licensed and non-licensed, meet this criterion and that person who does not exhibit physical fitness
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and mental alertness is immediately relieved of all duties l
associated with the operation of the plant."
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L-i This AP also mandates the annual memerandum to shift supervisors i
l-from the Director of TMI-1 and President of GPUN reminding them of
.their'overall command and control responsibilities.. Although sleeping is not especially addressed, there is a mandate to assure
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their people are " constantly vigilant."
Based on interviews with licensee personnel during this inspection
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and past inspections, the inspector noted no misunderstanding.on the applicability of AP 1029 to workers at TMI-1 and no misunderstanding
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that sleeping would be contrary to the policy.
i 7.3.3 Policy Implementation and Licensee Manitoring Past NRC staff inspections and this inspection have confirmed proper implementation of this policy or administrative procedures.
Also,' the licensee uses two primary means to assure proper implementation of the subject AP.
Records of the " Management Off-Shift Tour" program I
indicated attention to propar implementation of AP 1029 and, in some
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instances, they focused on operator / worker alertness and attentiveness.
These records predate the Peach Bottom public announcement.
Similarly, the other means is the Quality Assurance Monitor Reports, which also have verified proper implementation of this AP and, in particular, worker alertness / attentiveness. All records reviewed confirmed no violation of alertness / attentiveness requirements.
Subsequent to the Peach Bottom announcement, the off-shift tours again focused particularly on worker alertness / attentiveness with no discrepancies identified.
I The' inspector noted that the. Management Off-Shift Tour program was a l
substantial initiative on the licensee's part.
It forced managers in all disciplines to be in the plant during various back shift
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hours, thereby being an extension of the eyes and cars of the
Director of TMI-1.
To the operator / worker, it appears to be a reflection of mcnagement's attention to activities on the back shift.
To the managers conducting the tours, it appears to make them more aware of problems in the plant under their cognizance.
l For discrepancies identified, corrective actions are initiated.
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At a result of the Region I Administrator's tour, the Director of TMI-1 was receptive to having the off-shift tours focus on specific areas for material condition examination.
7.3.4 Disciplinary Records
The inspector also reviewed the records (made available to the NRC's Office of Investigation) of disciplinary actions on workers who were found or appeared to be asleep and for other adverse performances or misconducts.
The records, dating'back to"1980, indicate four
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' technician, and auxiliary operators.
Licensee representatives reported that no licensed operators were disciplined for sleeping since 1980.
The licensee took disciplinary action on each of these cases; the most severe being a termination of employment.
The inspector concluded that there were a few incidents of workers
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caught sleeping on the job and it appeared that licensee management actions were appropriate for the circumstances.
The' inspector discussed plausible disciplinary action with the key managers in the security, radiological controls, and operations areas for people found asleep. All managers were very vocal in expressing that sleeping on the job was a serious violation of company policy and indicated.that situation to be intolerable for which severe sanctions would be facing an individual if' caught and proven to be asleep. As an example, an automatic termination of employment would be imposed' on a security guard asleep. The inspector had no further comments in this area.
7.3.5 Work Management i
During the initial review of the TMI-2 allegation, the NRC staff questioned the use of the interface door (Door 7) between Unit I and Unit 2 at the south end of the Unit 1 Turbine Building.
The staff explored the possibility of TMI-1 personnel, while on duty, sneaking down to the TMI-2 Turbine Building -- the apparently unmonitored areas where sleeping was either consciously or inadvertently toler-ated.
The security department researched one day's access records for Door 7.
The computer search took two hours and it identified approxi-mately fifty transits through the door.
This seemed relatively high, so the inspector discussed with key managers the routine activities requiring personnel to use Door 7.
The transits were primarily security guards, radiological control technicians, and
auxiliary operators.
Typical activities are listed below:
security door checks;
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Unit 1 auxiliary operators (AO's) monitoring of the million
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gallon tank used for Unit 1, but it is located in Unit 2; and,
Unit 2 radiological control technicians use Door 7 as a path to
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the fuel handling truck loading bay when the Unit 2 "model room" is locked.
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These activities appear to be reasonable justifications for using the dcor.- Access is permitted by the security request procedures.
No-violation of. regulatory requirements were noted.
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review of the Door 7 records, in general, when someone left a unit,
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j frame.
Certainly, the implication of wrongdoing is unjustified based solely on the fact that a Unit I assigned individual transited the door.
These transits appeared to be for legitimate reasons.
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The inspector then quer,tioned key managers (security, radcon, operations)
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on accountability of personnel on the back shift who are assigned Unit 1 duties that may lead to Unit 2.
The general response of these managers was the work is intentionally managed to keep people working and/or prevent boredom. Ar, an example in the security area, a particularly dull post - is r.ot manned for more than two hours at a time.
Backshift
personnel have rou*,ine tours to perform.
Likewise, in the radcon area, work is also intentionally scheduled on the back shift so that a radcon
technician missing for two hours would be noticeable.
In operations, the swing shift is used to catch up on vlork from the day shift.
The mid-shif t usually has the responsibility for less complex surveillance in addition to routine tours and log readings.
The inspector could not rule out the possibility of a Unit 1 indi-vidual entering Unit 2 areas to engage in an unauthorized activity, such as sleeping.
However, because of lack of evidence based on past observations and based on this review, it appeared to the inspector that reasorable supervisory measures exist that preclude i
such an event without it being icentified by shift supervision.
i 7.3.6 Actions as a result of Peach Bottom /TMI-2 Events From the above review, the inspector noted that the licensee's policy, implementation, and monitoring were in place before the Peach Bottom
" sleeping operator" event was made public.
The inspector discussed additional actions that the licensee took subsequent to the Peach Bottom announcement.
They were:
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personnel discussion by the TMI-1 Director with all shift supervisors, reminding them of their responsibilities in this regard; re-emphasis in the "Off-Shift Management Tours;"
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NRC order to shutdown Peach Bottom and related documents;
other department briefings on the matter, such as the "Radcon" i
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manager with the Group Radiological Control Supervisors; and, re review of company policy on the matter to assure clear i
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guidance that the Peach Bottot situation would not occur at
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TMI-1.
In a status report to the GPUN President, the Director of TMI-1 stated:
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" appropriate written directions, high standards of performance, and j
numerous cross-checks and balances exist to prevent a problem of this t
magnitude from occurring or going unnoticed without appropriate corrective action at TMI-1...."
Also', as a result of the TMI-2 allegation, TMI-1 licensee management enhanced coverage on back shif ts for the weekend of July 10-12, 1987, and j
they reviewed out-of-the-way areas of the plant for " sleeping nests." No i
significant discrepancies were noted.
l 7.4 Worker Alertness / Attentiveness Summary The licensee's program to aseure worker alertness and attentiveness is a strength.
There was evidence that the policy was established i
well in advance of recent events on the issue.
The policy appears I
to be nroperly implemented and substantially monitored to assure i
adherence. Although sleeping incidents occurred, none involved j
licensed operators at TMI-1 since 1980.
The number are few and they could be expected considering the number of people to be managed and the unusual work hour conditions at the facility.
It appears the j
licensee appropriately dealt with past transgressions of the policy.
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i No conclusion can be drawn about the frequent use of the interface door between units.
It appears that the shift work load is spread out among J
twenty-four hours in which supervision and work accountability would reasonably lead to the detection of unauthorized activities by a Unit 1
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inoividual in TMI-2.
Unless the TMI-2 investigation uncovers evidence of
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wrongdoing by Unit 1 personnel, this matter will not be further reviewed by the NRC staff.
8.
Response to Licensee Performance Report j
On April 1, 1987, the licensee responded to the latest Systematic Assessment of Licensee Performance (SALP - NRC Inspection Report No.
50-289/86-99).
Region I acknowledged the licensee's response letter and incorporated it into the final SALP report by letter dated May 6, 1987, to the licensee.
The purpose of this review was to follow up on certain outstanding actions to which the licensee committed by l
their April 1, 1987, letter.
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1 In the functio ~nal areas of maintenance, emergency preparedness, and training,'.the licensee has no specific outstanding actions or
. comments in response-to the NRC staff's position on those areas.
There were no outstanding SALP board recommendations for these-
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areas.
In the Assurance of Quality area, the licensee deferred any specific comments to a meeting (a.SALP board recommendation) conducted February 12, 1987, on the Licensee Technical and Safety Review
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Process (see NRC Inspection Report 50-289/87-04).
This area was l
further reviewed in NRC Inspection Report No. 50-289/87-08.
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Also, in this area, the SALP board recommended that the licensee continue efforts in correcting procedure adherence and adequacy problems. The licensee acknowledged this in their response to the plant operations area.
The inspectors are continuing to routinely review this for proper procedure adherence and adequacy.
Outstanding actions in the radiological control, surveillance, l
security, and technical support areas were not within the scope of this review.
During this period, the licensee reported that the self-assessment in the technical support area has begun and is expected to be completed by the fall of 1987.
In the licensing area, the licensee's response to timely licensing submittals oriented toward improvements in the technical support area; i.e., incorporation of NRC submittals into the lung range plan.
The licensee is in the process of adopting an integrated living schedule. The licensee also e ntly reorganized.and has a corporate vice president assigned to the planning and safety review area for both TMI-1 and Cyster Creek.
Improvements in the technical support area will be reviewed in a future inspection.
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Also, in the licensing area, the licensee responded to inservice testing
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issues for Cycle 6 startup.
This was reviewed in NRC Inspection Report i
Nos. 50-289/87-08 and 87-09.
During this inspection, the inspector focused on the licensee's response in the plant operations area and discussed planned actions with licensee management. The SALP board recommendations in this area were the same for the Assuran<.e of Quality area.
The licensee response is summarized below.
The licensee disagreed with the staff's comment that, at times,
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middle management may have been a negative influence in the procedure adherence and adequacy area with respect to schedule pressures and shortsighted review of events.
Additional steps were taken to strengthen the Independent
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Verification Program (previous Unresolved Item No.
289/85-27-08, which is still open) is to be reviewed in a future inspection.
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implementation (see NRC Inspection Report No. 50289/87-08).
l The latter two items were addressed by NRC staff as referenced above.
Consequently, the middle management issue was discussed with the licensee site upper management.
They reiterated their position that middle management is a positive influence in this area.
The inspector stated l
.that, in general, the SALP report acknowledges that; but the purpose of staff comments was for licensee upper management to re-examine those i
events where negative performance resulted.
The re-examination should have identified the factor (s) identified by the staff and other factors l
identified by the licensee that may have adversely affected performance such that upper management could remove or alleviate such factors to a
enhance performance.
Site upper management agreed to keep that in mind, especially during the non-routine or off-normal events, to assure such things as proper procedure adherence / adequacy, proper pace of activities commensurate with procedure adherence, and adequacy of review of events.
In this area, overall positive performance results occurred for the i
unexpected recent letdown cooler replacement outage (see NRC Inspection
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Report No. 50-289/87-11).
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The inspector had no additional comments in this area, 9.
Licensee Actions on Previous Inspection Findings
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9.1 (Closed) Unresolved Item (289/87-09-07):
Restoration of the Main Feedwater (MFW) Isolation Function i
.During the previous inspection, the licensee revealed a concern j
about placing the Heat Sink Protection System (HSPS) MFW isolation function in " enable." Apparently, as a result of the licensee's review of the Davis-Besse loss of the feedwater event of June 1985, the licensee became concerned that steam pressure oscillation may
also calse significant oscillation in the HSPS steam generator (SG)
l operating range (0P) level indication system.
This could result in an inadvertent MFW isolation on a post-trip situation, thereby compounding the operator post-trip response actions.
Electronic filters for the startup (50) and OP ranges were ordered but not received prior to Cycle 6 startup.
The electronic filters were designed to dampen these oscillations.
The electronic filters were received and installed and the isolation function was enabled as documented in NRC Inspection Report No.
50-289/87-10.
Plant experiences on two reactor trips since that installation showed no problem with level oscillations causing inadvertent isolation of the MFW system.
This item is considered closed.
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I 9.2 (Closed) Unresolved Item (289/87-09-08):
Adequacy of Licensee
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Safety Evaluation for the HSPS MFW Isolation Function in " Defeat" Related to the above item (289/87-09-07), the licensee comple'ted two j
safety evaluations for permitting power. operation with the HSPS MFW
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isolation function in defeat until the electronic filter was in-stalled and until the isolation setting was raised from 94 percent j
to 97.5 percent on the operating range. With the assistance of.the
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Office of Nu. ar Reactor Regulation (NRR), Region I reviewed these
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safety evaluations and related licensee internal correspondence.
NRR provided Region I with the following information.
l There is presently no safety design basis or staff requirement for the High Level Isolation of MFW (HLIMFW) system.
However, the proposed resolution of Unresolved Safety Issue (USI) A-47, " Safety Implications of Control Systems," includes requirements for all-pressurized water reactor (PWR) plants to have a single-failure-
proof commercial grade (or better) automatic steam generator over-fill protection system that isolates the main feedwater system on
high level..The. proposed resolution also would require all PWR licensees to include the overfill protection system in their plant i
technical. specifications.
The staff has no current requirements for j
steam generator overfill protection to be in'cluded in technical
specifications.
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NRR also concluded that the evaluations were thorough and adequate.
However, the licensee committed to the installation and operability i
of the HLIMFW during the TMI-2 Restart Hearing and in correspondence l
related to TMI TAP II.E.1.2.
Although no TS exist for this
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function, facility procedures require its operability to fu' fill
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commitments made by the licensee.
This item is considered closed.
9.3 (Closed) Unresolved Item (289/87-09-09):
Deportability of Engineered Safety Feature (ESF) Actuations During the past inspection, the inspector questioned why inadvertent actuation of the HSPS was not reported in accordance with 10 CFR j
50.72 or 50.73.
The licensee recently installed HSPS as the safety grade initiation and control system for the emergency feedwater (EFW) system.
The licensee initially responded that the HSPS/EFW were not considered ESF systems.
NRC Region I disagreed; l
and, in the cover letter, dated June 23, 1987, to NRC Inspection
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Report No. 50-289/87-09, Region I acknowledged the following action
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to be taken by the licensee.
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"(1) consider future actuations of EFW/HSPS as ESF actuations for reporting purposes -- implementation of this commitment will be n
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proceduralized through revisions to appropriate station procedures;
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+ snd, (2) conduct a raview to determine if actuations of any other systems not listed as ESF's in your FSAR should be reported in accordance with the intent of 10 CFR 50.72 and 10 CFR 50.73 - you will advise the NRC resident inspectors of the results of this review."'
Further, NRC Region I did not expect Licensee Event Reports (i.e.,
10 CFR 50.73 reports) for past EFW/HSPS actuations. With respect to item (1) above, the idcensee revised (Revision 17, dated July 17,
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1987) Administrative Procedure (AP) 1044, " Event Review and Report-
,ing Requirements." The applicab7e sections of the procedure which
' implement 10 CFR 50.72/50.73 requirements now have a note indicat-
. fog: "At TMI-1, ESF includes RPS [ Reactor Protection System] and
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.ESAS [ Engineered Safety Actuation System]. Additionally, TMI-1 has agreed to report HSPS actuations." This licensee action resolved the deportability of the HSPS actuations issue.
With_ respect to item (2), the licensee reported that the review of
the above noted revision to AP 1044 also constituted their determi-nation that, since no ESF definition occurs in the Final Safety Analysis Report (FSAR), no other system actuations need be reported.
[The ESAS systems are tabulated and listed in AP 1002, Enclosure 9, and they do= include HPI (high pressure injection), LPI (low pressure injection), and containment isolation / cooling system and their support auxiliary sub systems].
The licensee's site upper manage-ment also indicated their willingness to cooperate with the NRC and report other actuations if NRC staff needs the information and after their own review of the NRC staff's needs/ reasons.
The inspector stated that the only other actuations of interest that come to mind may be the control room ventilation actuation into the recirculation mode for control room habitability and prevention of a toxic environment in the control room. Again, the licensee reiterated that such an actuation'would not be considered an ESF F
actuation, but they would have to review a NRC staff request for such reports.
The inspector concluded that the problem herein lies with the lack of definition of an ESF system in the FSAR.
NRC staff guidance for
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rule 10 CFR 50.72 (NUREG 1022 and Supplement 1) defers to the
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regional office or FSAR for a definition of an ESF system.
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This matter was discussed with representatives of the NRC's Office
of the Analysis and Evaluation of Operational Data (AEOD) and Office-1 of Nuclear Reactor Regulation, Operational Event Assessment Branch.
I During these discussions, it was learned that many licensee repre-I sentatives do consider the control room ventilation an ESF system and, as such, they report to the NRC staff inadvertent actuations j
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However,- these representatives confirmed
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that the regulatory guidance on what constitutes an ESF system is not.
.i clear and it is highly dependent on how the licensees write their FSAR's, j
AE00 is considering additional guidance in that regard.
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Accordingly, this matter is considered closed pending additional action.by NRC staff on further defining ESF systems.
9.4 { Closed) IE Information Notice (289/86-IN-53)-
Improper Installation of Heat Shrink Tubing Over Electrical Splices and Terminations This notice describes a potential generic safety problem involving improper installation of heat shrink tubing over electrical splices and terminations. The inspector reviewed the licensee's' program for the installation of heat shrink tubing.
This review consisted of the following actions:
verifying that appropriate training on the use of heat shrink tubing has been implemented; that Quality Control
- inspections are performed on an appropriate number of splices; and, that the procedure used to control heat shrink tubing work is in accordance with station administrative controls and manufacturer's i
recommendations.
The inspector's review indicated that the licensee has adequately addressed the concerns discussed in Information Notice 86-53.
Based on the above, this item is closed.
9.5 (Closed) Unresolved Item (289/87-10-02):
Hydrogen Re;ombiner
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Missing Surveillance Test The completed surveillance procedure and data for post-maintenance
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testing of the "1A" Hydrogen Recombiner on May 5, 1987, was found by the licensee. At the time of the previous inspection the record for SP 1303-11.46 A/B', " Hydrogen Recombiner Functional Test," was missing because it was misfiled in the surveillance recoros files in the control room. The enmpleted record indicated a satisfactory test.
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I While looking at this flie, the inspector also verified proper quarterly
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completion of this test as required by TS.
The inspector also reviewed Revision 7, Ated July 27, 1987, to Preventive Maintenance (PM) Procedure M-148, " Hydrogen Recombiner Lubrication." The revision, as committed to by the licensee in a previous inspection, alleviates inspectors' concern for unnecessarily operating associated containment isolation valves and it calls for coordinating the PM at the time of surveillance procedure completion.
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9.6 '(Open) Unresolved Item (289/87-11-04):
Review Licensee Correc
- tjve Action for Technical Adequacy of Fire Protection Reviews In a letter, dated July 27, 1987, the licensee reported two noncompliance related to the fire hazards analysis report.
The first concerned a conduit containing the control cable for the solenoid. valve for IC-V4, L
which was f" nd unprotected in FH-FZ-6.
The conduit contains the control
. cable for IC44.which is unsheathed Rockbestos cable, and conventional cables. A design drafting error occurred in the design process and, as a
resu't, the wrong conduit in the zone was wrapped.
Additionally, in FH-FZ-1, two control circuits for fan AH-E-15B were found to be unprotected. AH-E-ISB.is relied upon for ventilation in AB-FZ-7. -The noncompliance in FH-FZ-1 is similar to those zones where an exemption was previously granted.
This exemption allowed for the use of the roving firewatch in lieu of protection as re-quired by Section III.G.2 of Appendix R.
As a compensatory measure for the noncompliance, the existing roving firewatch was extended to include FH-FZ-1.
Also.as a result of "as-built" reviews conducted by the licensee's
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contractor, a licensee representative verbally reported to the resident inspector on September 3, 1987, additional "as-built" information concerning fire protection capability.
For areas already covered by the roving fire watch, new equipment has been
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identified in those areas as needing protection, which apparently
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was not previously known.
For example, the fire watch passes through a hallway in the intermediate building (IE-FZ-5), which was
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not specifically identified as a fire zone with cables (that need protectf on)' providing control and/or power for a safety-related fan l-in the. diesel generator building. Another fire zone (CB-FA-31) in the control building has power cables for fans in the intermediate building for'the emergency feedwater pumps.
No new areas need to be covered by the fire watch patrol, but the recent
, identifications as reported on September 3, 1987, warrant procedure revl? ions.
The licensee plans to revise Emergency Procedure (EP) 1202-31, Revision 30, effective July 6, 1987, " Fire." This EF lists affected components for fires in various areas or zones in the plant.
The licensee also plans to provide this information to the Office of Nuclecr Reactor Regulation, which is to review the final fire hazard analysis report and issue a safety evaluation report on the matter.
This item rer.tains open pending licensee action as noted above and further NRC staff reviLw.
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10.
Exit Interview The inspectors discussed the inspection scope and findings with f
licensee management at a final exit interview conducted September 4, 1987, Senior licensee personnel attending the final exit meeting included the following:
Licensee H. Hukill, Director, TMI-1 C. Hartman, Manager, Plant Engineering, TMI-1 C. Incorvati, Audits Supervisor, TMI-1 S. Otto, TMI-1 Licensing
.L. Ritter, Administration,. Plant Operations L. Robinson, Representative - Media Relations M. Ross, Director, Plant Operations, TMI-1 D. Shovlin, Manager, Plant Maintenance, TMI-1 C. Smyth, Manager, Licensing, TMI-1 N.RC L. Bettenhausen, Chief, Reactor Projects Branch 1, Division of Reactor Projects The inspection results as discussed at the meeting are summardzed in the cover page of the inspection report.
Licensee representatives did not indicate that any of the subjects discussed contained proprietary or safeguards information.
Unresolved Items are matters about which more information is required in order to ascertain whether they are acceptable, violations, or deviations.
Unresolved items discussed during the exit meeting are addressed in Sections 2, 3, and 9.
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ATTACHMENT 1 NRC INSPECTION REPORT NO. 50-289/87-13 l
ACTIVITIES REVIEWED
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I Plant Operations
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Control room operations'during regular and back shif t. hours, including frequent observation of activities in process and periodic reviews of selected sections of the shif t foreman's log and control room operator's log and selected sections of other control room daily logs Areas outside the control rocm
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j During this inspection period, the inspectors conducted direct inspections during the following back.chift hours.
Date Time 7/10/87
_ 9:00 p.m. - 11:00 p.m.
7/11/87 12:00 p.m. - 1:00 p.m. (Saturday)
7/11/87 5:15 p.m. - 7:15 a.m.
7/11/87 8:30 p.m. - 10:30 p.m.
7/12/87 6:00 a.m. - 8:00 a.m.
7/12-13/87 11:00 p.m. - 2:30 a.m.
7/22/87 6:00 a.m. - 7:00 a.m.
_8/22/87 9:30 p.m. - 10:30 p.m.
9/3/87 6:00 a.m. - 7:00 a.m.
Maintenance JT CL D61 - CO-V5 requested February 8,1987, completed June 21, 1987,
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air inlet leak on' solenoid operated air valve JT CN 119 - Requested June 13, 1987, completed June 17, 1987 - CA-V1 will
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not close; cleaned torque switch contacts JT CN 221 - Requested July 1,1987, completed July 1,1987 - Repair
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contact circuit for MS-V-14A, valve JT CM 914 - Requested May 29, 1987, completed June 26, 1987, request
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cables for NI-1 per attached change modification request Surveillance SP 1303-4.14, Revision 14, April 16, 1986, " Reactor Building 30 psig
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Analog Channels" L_-_-_-_____-______=_-_-_.
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. Reactor Coolant' System-(RCS) Leak Rate The inspector selectively reviewed RCS leak rate' data for'the past 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 1107.
Licensee
.(L) and NRC (N) data are' tabulated below.
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RCS LEAK RATE DATA
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DATE/ TIME
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CORRECTED DURATION L
N N
N L
g g
g g
U 07/23/87
.5683
.57-0.08 0.02
.0172 11:45 p.m.
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2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 07/31/87
.1481 0.15-0.09 0.01
.0118 4: 10 p.m.
2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />
'08/02/87
.0798 0.08-0.15-0.05 0410 3:45 p.m.
2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />
08/05/87
.1606 0.16-0.04 0.06
.0600 4:13 p.m.
2. hours-08/09/87
.7930 0.79-0.06 0.04
.0418 4:02 p.m.
i 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />
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08/12/87
.1539 0.16-0.02 0.08
.0847 4:17 p.m.
.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> G = Identified gross leakage U = Unidentified leakage L = Licensee calculated N = NRC calculated Columns 2 and 3; 5 and 6 correlate 1 0.2 gpm in accordance with NUREG 1107.
(N is corrected by adding 0.1044 gpm to the NUREG 1101 N due to u
u total purge flow through the No. 3 seal from RCP's.
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ATTACHMENT 2 l
REGION I ADMINISTRATOR
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INSPECTION / TOUR AT TMI-1 JULY 16, 1987 Specific Discrepancies 1.
348-foot elevation Fuel Handling Building (FHB) - Debris under RM-L-9 and NS/IC closed cycle surge tanks 2.
329-foot elevation FHB (by elevator) - PC head cover adrift on cable conduit in contaminated areas 3.
305-foot elevation FHB, Spent Fuel (SF) Cooler room, southeast corner - Protective cover being off of position indicator for ventilation damper represents an electrical shock hazard 4.
SF Cooler. Room - Debris in contaminated area adjacent to pumps and under SF-V3 5.
305-foot elevation Auxiliary Building (AB) by ES switchgear -
cautionary labeling for switchgear on floor, not on appropriate panel s 6.
Closed Cycle Cooling Room - Very dirty at motor intake ventilation louvers 7.
"1A" Makeup Pump Room - An RTD conduit connector cap looked loose (in contaminated area), previous cap " impression mark" was exposed 8.
281-foot elevation AB, a Decay Heat (DH) vault - Pump casing vent / drain (DH-V-148) coupling ' leak wrapped with cloth (boric acid soaked)
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A DH vault - DH suction piping drain had active leak with end flange off and adrift; potentially a contaminated area not so marked 10.
Heat exchanger vault - NR-V-19 MOV mounting nut and lock washer appeared to be not properly set 11, 322-foot elevation Control Building, Battery Room - Tight plywood (
between cells may not allow for thermal expansion of the cells on battery discharge during testing or needed use 12.
B" Diesel Generator Building - Speed changer cable insulation interfer-
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ente with fuel rack operating linkage
"A" diesel out to be checked for similar situation 13.
Multiple contaminated areas - Need clearing, obvious not touched as frequently as non-contaminated areas J