IR 05000289/1988001
| ML20148B253 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 03/10/1988 |
| From: | Cowgill C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20148B229 | List: |
| References | |
| 50-289-88-01, 50-289-88-1, IEB-87-002, IEB-87-2, NUDOCS 8803210410 | |
| Download: ML20148B253 (31) | |
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U. S. NUCLEAR-REGULATORY COMMISSION REGION I ,
Docket / Report No.'50-289/88-01- Licensee: OPR-50 l Licensee: GPU Nuclear Corporation P. O. Box 480 ;
Middletown, Pennsylvania 17057 ';
Facility: Three Mile Island Nuclear Station, Unit 1 i i Location: Middletown, Pennsylvania l
Dates: January 9 - February 6,1988 l
I Inspectors: *R. Conte, Senior Resident Inspector .
D. Johnson, Resident Inspector !
Accompanied
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by: S. Peleschak, Reactor Engineer, Region I (RI) j A. Sidpara, Resident Inspecter !
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- Reporting In ector !
Approved by: ( /d 8d C. CowjTTl', ChiepReactor i, ects Section No. lA Dat'e l Inspection Summary: The review was a routine safety inspection (162 inspection hours) assessing performance primarily in the plant operations and maintenance i area In the plant operations area, the inspector reviewed: various plant j transients; verification of correct activities; and, partial loss of the instrument !
air control syste In the maintenance area, the inspector reviewed: the instal- !
lation of expansion jcints into the' reactor building emergency cooling water system; !
battery terminal cleaning; repair of a liquid waste transfer pump; differences in !
the diesel generator injector assemblies; liquid waste disposal transfer line re- !
pair; and other surveillances and maintenance-related activities. A modification t was also reviewed dealing with revised setpoints for the low level interlock on the reactor building sump isolation valve. Also, reviewed were Safety Issue Man- !
agement System (SIMS) Nos. A-26, "Reactor Vessel Overpressure Protection," and l MPA-B-66, "Natural Circulation Cooldown." The review of past inspection findings :
focused on licensee actions related to past violations in tne plant operations, I surveillance, and maintenance area I Inspection Results: Overall, operational performance remained at a high leve Operational mistakes were made and/or equipment malfunctions challenged licensed and non-licensed operators. However, the mistakes were licensee self-identified and operators were responsive to the challenges they received by taking immediate appropriate corrective actions. Weaknesses in the implementation of the indepen-dent verification program continued to be noted. Management attention and involve-ment in these and other facets of operations were noteworth PDR ADOCK 05000289 s G DCD
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The continued trouble-free operation of the plant without any unplanned maintenance j outages is indicative of ar overall effective maintenance program. The negative I findings identified of the inspector were indicative of certain procedural weak-nesses in the areas of control of applicable and non-applicable work instructions and of control of procurement and replacement of parts. In particular, several examples of maintenance-related errors collectively represented a violation of regulatory ,9quirements (paragraph 3.3). The most significant example was the installation of a wrong model (pressure rating) expansion joint in the reactor building emergency cooling water system. Overall, the licensee's initial and im-mediate corrective actions were responsive to the inspector's concerns. However, measures to prevent recurrence may need enhancemen For licensing actions, in particular SIMS Nos. A-26 and MPA-B-66, the licensee effectively translated general licensing actions into plant specific actions with one exception. Also, the licensee took acceptable corrective actions and/or ful-f-illed commitments made in response to the violations reviewe Licensee personnel were cooperative in providing additional information to resolve regulatory issues such as for the fastener sampling required by NRC Bulletin No. 87-0 , -
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DETAILS 1.0 Introduction and Overview '
1.1 NRC Staff Activities The overall purpose of this inspection was to assess licensee activities during the power operations mode as they related to reactor safety, safeguards, and radiation protection. Within each area, the inspectors documented the specific purpose of the area under review, acceptance criteria and scope of inspection, along with appropriate findings / con-clustons. The inspectors made this assessment by reviewing information
. on a sampling basis through actual observation of licensee activities, interviews with licensee personnel, measurement of radiation levels, or ,
independent calculation and selective review of listed applicable docu-ment .2 Licensee Activities During this period, the licensee operated the plant at essentially full power. There were several plant transients as noted below and as ad-dressed in paragraph 2.2.1 of this repor On January 18,1983, at 4:46 a.m. there was a transient fror 100 to 101 percent power, apparently due to a malfunction in the inte-grated control system (ICS). The operators manually stopped the transien ,
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On January 22, 1988, between 5:19 p.m. and 5:33 p.m. there was a transient between 101 and 90 percent power, which was apparently operator induced while switching between manual and automatic ICS control during maintenance work on the IC On January 28, 1988, between 1:10 p.m. and 1:20 p.m. there was a transient from 100 to 97 percent power, apparently due to ICS mal-function in feedwater demand signal processing circuit Also, between 3:57 p.m., January 31, 1988, and 2:10 a.m., February 1, ,
1988, the licensee operated at reduced power (approximately 77 percent)
in order to make repairs to a feedwater pump (FW-P-1A). Coupling grease had leaked out of the feedwater pump coupling causing excessive vibra-tions on the pum .0 Plant Operations 2.1 Criteria / Scope of Review The resident inspectors periodically inspected the facility to determine the licensee's compliance with the general operating requirements of Section 6 of the Technical Specifications (TS)' in the following areas:
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review of selected' plant parameters for abnormal trends; 9
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plant status from a mai'ntenance/ modification viewpoint, including .
plant' housekeeping'_and fire protection measures; '
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control of engoing'and:special evolutions, including' control room personnel awareness of these evolutions;
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control of documents, including logkeeping practices; <
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implementation of radiological controls; and,
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irplementation of_the security plan including access' control, ,
boundary integrity, and badging practice The inspectors focused on the areas listed in Attachment .
-r 2.2 Findings / Conclusions >
2. Power Transients -;
Paragraph 1.2 of this report summarized the various power transients !
that occurred-during this period. Since the Cycle 6 startup, the j licensee continued to observe minor power oscillations apparently -
due to control system "hunting" in certain sections of the IC i The licensee continued to trcubleshoot the problem leading to the .!
installation of an electronic filter circuit into the. suspected l problem section of the ICS to dampen these oscillations. These j actions were not completely successful in giving the licensee de- .
sired result !
l During this period, the power transients were correlatable to main- !
tenance on selected operational activitie_s with the ICS trouble-shooting. Although root causes could not be clearly identified, l they were generally due to an. equipment malfunctioning and/or were 1 operator induced. Appropriate-equipment corrective actions were i taken, such as replacement of selected ICS components. The opera- t tor-induced events were overshadowed by the positive operator re- ,
sponse to these events, es'pecially those actions that mitigated . !
power increases over 100 percent. No plant trip resulted and there '
were no challenges to safety-related equipment. The inspector :
identified no concerns on these event ;
2.2.2 Verification of Correct Activities :
On January 18, 1988, at_9:20 p.m., control room operators observed i that the open and close lights were out for reactor-building emer- l gency cooling valve RR-V4A, "Supply Isolation for the "A" Ventila- !
tion Cooler in the Reactor Building (RB)." After further review, .;
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the operators determined that the breaker for the motor-operated valve at the "1A" engineered safeguards (ES) Motor Control Center (MCC) switchgear in the control building was ope The valve and its MCC breaker are normally closed, but an engineered safeguards actuation system (ESAS) is supposed to open the valve for R8 emergency cooling water flow. With the breaker open, the safety functions would not have been fulfilled. There are two other coolers that could be used in the R Licensee operations personnel attempted to identify the root caus There were workers in the area at the time of event identification, but they provided no additional information on how the breaker was opened. The inspector noted the breaker was located at about foot level on the switchgear panel, but it was not in a main traffic pathway. No cause was determined; however, inadvertent bumping could not be ruled ou The licensee concluded that the ESAS checklist was successful in performing its intended function, since it prompted operators to look at the open/close status for this valve (once per shift). The inspector similarly concluded on the licensee's finding Further, on January 23, 1988, at 11:05 p.m., on-coming control room operators noted that Core Flood Containment Isolation Valves CF-V-19A and B were open and should be closed. The previous shift oper-ated these valves to maintain proper pressure and level in the Core Flood Tanks. However, the shif t did not properly restore this lineup to normal. The on going operator verified no current acti-vities with the subject valves and closed the The valves remained operable in that a containment isolation signal would have closed the valves. Further, the system was still suf-ficiently isolated to prevent Core Flood Tank pressure and level from being out of technical specificatian range requirement The inspector noted that these valves are not listed as "critical valves" (as defined by the licensee in their Independent Verifica-tion Program (IVP) (Administrative Procedure (AP) 1067). Only cri-tical valves require independent verification. The applicable Operating Procedure (0P) called for independent verification on restoration to normal for these valves. This obviously was not completed. The licensee intends to make OP consistent with the AP 1067 program. These program and implementation weaknesses were addressed in NRC Inspection Report No. 50-289/87-1 The inspector considered it noteworthy that the shift personnel alertly identified the incorrect positions for the subject valve At the exit interview, licensee management reiterated their active
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Institute of Nuclear Power Operations (INP0) sponsored program to thoroughly review human performance errors. The inspector _had no-
. additional comments in respons . Partial-Loss of Instrument Air Control System At 3:00 a.m. on January 20, 1988, while workers were working on WOL-V61 pipe repairs (see paragraph' 3.2.5) in the auxiliary building, an instrument air pipe coupling _in'the same area broke causing low air pressure.in:the system. Control room operators responded by assuring back-up. instrument.and service air compressors (four total) .
were operating to make up to the system. Auxiliary operators re-sponded by locally isolating the leak. :As a result of this local isolation, the inlet valves.to parallel letdown filters went closed causing-a loss of. letdown flow; and, as' expected, the letdown relief valve lifted to'the Miscellaneous Waste Storage Tank. This resulted
.in a slight increase'in the effluent monitor noble gas channel for the auxiliary / fuel handling building ventilation system. ,This re-lease was well within technical specifications limit By 3:12 a.m.,
instrument air system _ pressure'was.within normal range _and the reactor plant was relatively unaffected, except as noted abov As a part of the interference removal work for WOL-V61, workers removed the air operator for WOL-V61 while instrument air (IA) re-mained. lined up to the valve operator. This was intentional because isolating IA to the valve.would have also isolated IA to-the letdown isolation valves (see above) eventually causing these valves to go shut. Apparently, as the valve operator for WOL-V61 was moved, a coupling came loose and blew open. No personnel injuries occurre The auxiliary operator (AO) on watch was alert to effectively iso-late the leak with a minimal impact on the plant. The inspector had no further questions on this matte .3 plant Operations Summary Overall, operational performance remained at a high level. Operational mistakes were made and/or equipment malfunctions challenged licensed and non-licensed operators. However, these mistakes were licensee self-identified and operators were responsive to the challenges they received by taking immediate appropriate corrective action. Weaknesses in the-implementation of the IVP program continued to be noted. Management attention and involvement in these and other facets of operations were noteworthy.
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3.0 Maintenance / Surveillance - Operability Review 3,1 Criteria / Scope of Review The inspectors reviewed selected activities to verify proper imple of the applicable portions of the maintenance and surveillance program The inspector used the general criteria listed under the plant operations section of the repor Specific areat of review are listed in Attachment A more detailed review of equipment operability is also addresse .2 Findings / Conclusions 3. Expansion Joints for the Reactor Building Emergency Cooling System Pumps 3.2. Background An improper expansion joint for the "B" pump of the reactor building emergency cooling system (RR-PIB) was installed in September 1987. This problem was identified in NRC Inspec-tion Report No. 50-289/87-24. The licensee prepared a satis-factory evaluation to justify continued operation; and, sub-sequently, they initiated a detailed review on how this occurred. The licensee provided the results of their review on January 13, 1988, as planned. The inspector also looked into the other elevant areas for the conduct of the par-ticular mainten nce activity. The details are as follow .2. Licensee Findings The licensee's findings were documented in an internal mem-orandum dated January 12, 1988, from the Manager of Quality Assurance (QA) Modifications / Operations to Operations and Maintenance Director, TMI-1. A summary of licensee findings is addressed belo (1) The expansion joint was procured from a different vendor as a replacement-in-kin The plant engineering evalu-ation on this replacement-in-kind, as required per pro-cedure PEP-2, Revision 3, dated March 30, 1987, "Plant Modification and Replacement-in-Kind Applicability /
Scope," was not properly implemented prior to the in-stallatio The review was completed on December 11, 1987, following identification of the proble *
(2) The end use for the several joints specified in the pu"chase requisition was not transferred on the purchase order. Therefore, the receipt inspection tags, which are based on the information in the purchase order, did not identify the specific application eithe The lic-
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ensee stated that this was a repetitive problem identi-fled previously by the Quality Assurance Department (QAD) (Quality Deficiency report (QDR) No. DLL-037-87, May 1987). The corrective action was to implement an increased level of review by Materials Management and to provide for a computerized system that automatically transfers information from the purchase requisition to the purchase orde (3) The receipt inspection did not discover the item which was mislabeled on the outer wrapping and, also, it was incorrectly tagged by the vendor. The licensee plans to instruct the Quality Control (QC) receipt inspectors to remove the outer wrapping when identification of the item is questionabl (4) The job ticket (CK-212) indicated the incorrect model number of the expansion joint because the JT was pre-pared using an unapproved copy of the purchase requisi-tion which did not include the end use (see above).
The corrective maintenance (CM) procedure 1407-1, Re-vision 30, does not provide detailed direction for selecting replacement parts. QAD is working with main-tenance to upgrade the applicable part of this procedur (5) The final phase of the installation was witnessed by a QC inspector and the documentation was appropriately verifie The QC inspector did not catch the erro He did not witness the removal of the old joint nor did he compare the new joint with other pump joints. The licensee is still reviewing different ways of enhancing the inspection proces (6) Following the NRC inspection, engineering provided a justification for continued operation. However, no Materials Nonconformance Report (MNCR) was prepared as required. The MNCR was issued by QAD to document the conditio .2.1.3 NRC Findings The inspector concurred with the above-noted licensee find-ings; but, based on tnis review, the inspector had the fol-lowing additional observations. The inspector also reviewed JT No. CK-211 for a similar replacement on the "A" pump (RR-P1A).
(1) Plant Engineering Review (PEP-2) of replacement-in-kind was not performed as required per procedure PEP- This was identified by the licensee as an oversigh Based
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upon past NRC experience, such reviews are performed as necessary and, therefore, this incident is considered'
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an isolated cas .
(2) JT No. CK-212 for the "B" expansion. joint did not in-
'clude the current revision of the installation proce-dure as required by CM 1407-1,, Revision 30.-;There are a few' blanks on.the job ticket which are not filled in -
per the above-mentioned procedur For example, em-ployee number, the cause of failure, regulatory agency, outage codes, etc. Maintenance Procedure (MP) 1407-1~
does not clarify as to the 'use of "N/A" and initials of an authorized person for the data which is not ap-plicabl The job ticket did not specify part number, model number,
. purchase order,-etc. anywhere, except on the inspection tag, (which specifies the wrong model). On the work planning form, page 2, under part 2.1, "Materials," the
purchase. order number is not filled i This incomplete information appears to have contributed to the wrong installation. JT No, CK-211 for the "A" pump' had simi-
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lar error (The final installation of the expansion joint on RR-PIB under JT No. CP-683 did include all the pertinent data.)
(3) The original Bill of Material RN-220 for the expansion ,
joints specifies 160 psig as the design working pressur i The expansion joints installed on RR-P1A and 18-are ~I rated at 150 psig. While'the difference is insignifi-cant and since the maximum discharge pressure is well within the limits, there is no safety significance; however, the discrepancy should have been addressed in the documentation revie (4) The MNCR no. 0176-87, dated'Oecember 14,'1987, could have provided more: specific.information about the com-ponent identification, such as model, part number, et The corporate administrative procedure 1000-ADM-7215.01, Revision 1, "GPUN Material Nonconformance Reports and Receipt Deficiency Notices," as well as the Operational i
, Quality Assurance Plan 1000-PLN-7200.01, Revision 1-00, emphasizes the importance of component identificatio (5) Job Ticket (JT) No. CK-211 for the installation of the joint on the pump RR-P1A included an inspection tag, which shows Model No. 500R, even-though the actual in-stallev joint is Model No. 150R. After the inspector identified this discrepancy to maintenance personnel,
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th_e correct. tag was then included in the job ticke l The normal work / document review process.did not -identify
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and correct this erro :
(6) The work planning sheets of'JT'No. CK-211 specifies'a i torquing procedure (1410-Y-72). The procedure was .
voided on September 10', 1987, after the joint apparently !
-was overtorqued during installation on September 8, 198 ;
The use of this procedure-means excessive pressure on ;
the rubber flange. On January, 15, 1988,-the licensee .
corrected this problem by loosening the' bolts as re . !
quired. (NOTE: The joint for the "B" pump was replaced 3 with the correct model on January 13, 1988, without !
using the torquing procedure. The installation wa i witnessed by the NRC inspector.) ,
(7) For JT No. CP-683 conducted on January 13, 1988, the 5 replacement joint on the RR-PIB showed 150MR model +
number embossed on the outside'of the join (The discrepancy was identified by licensee QC inspectors-witnessing'the replacement.) The-job ticket and other documents indicate 150 Based on the' discussion with the engineering and QC personnel, it was noted that_the letter "M" stands for "Mercer," the supplier, and does not have any significance. The clarification of the model number discrepancy, if documented, would have been an enhancemen The licensee management involvement and respontiveness was quite noteworthy. As soo.n as the problem was iden-tified, the licensee took immediat'e actions to perfor an engineering analysis to justify continued operation and then followed up with the detailed report. The licensee's review was sufficiently thorough in identi-fying several areas that need improvement as a result of the improper installation _of the expansion' joint .2.2 Cleaning of the Station Battery Terminals On January 14, 1988, during the mid-shift (11:00 p.m. to 7:00 a.m.),
the inspector witnessed the subject activity being performed per JT No, CP403, dated December 12, 198 The specific purpose was to assess the quality of work, documentation, adequacy of planning, procedure, as well as procedural compliance. The scope of th activity involved cleaning of minor corrosion products and applying a thin layer of non-ox-id grease on the affected terminal. The following observations were mad . - - . - . . .. . - . _ - - _ _ _ _ _
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(1) The battery terminal activity was being performed by two :
craftsmen; however, the JT was not at' the;worksite. When ,
questioned by the inspector, the JT was brought to.the worksite .
shortly thereafter. Based on the inspector's review, worker li activities were in accordance with prescribed instruction .
(2) The JT specified.a CM 1420-EL-1, Revision' 5, "Troubleshooting '
Unit 1 Battery Chargers."' Revision 5 was not specified on the ;
JT as required by MP 1407-1. The craftsman had marked "N/A" ,
on this procedure since that it was obvious that it was aEwron i procedure for the job. The JT was reviewed and. formally ap--
proved this way prior to starting the job. The planning ~ pro- !
cess, however, did not identify the wrong procedure specified '
on the ticket. The licensee reported that this was due to an *
administrative oversigh !
(3) The JT package included a Preventive Maintenance (PM) Procedure !
-E-72, Revision 6, "Station Batteries Terminal Connection In- ;
s p ec ti on . ". The JT did. not specify the use of this procedur l This condition was discussed by the inspector with the shift <
maintenance superviso The copy of the JT received on January 20, 1988, by the inspector showed the correct procedure and ,
was signed by the maintenance supervisor. The work activity, i when witnessed on January 14, 1988, was in progress without ;
formal review and approvals of the steps that were marked "N/A".
(4) The craftsman did not have the protective clcthing as recom-mended in the E-72 procedure. Also, grease heating was com-pleted in the battery room. The procedure did not include any e grease heating requirements. .This is indicative of weak im- !
piementation of' occupational safety and health measure i h
(5) The terminal clean-up activity was sufficiently defined in JT :
so as to preclude e safety issue. It was quite evident that I the craf tsmen had a very good knowledge of the scope of the :
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3.2.3 Liquid Waste Transfer Pump Repair ;
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, The corrective maintenance activity for JT No. CP-727 on WOL-P78 '
was to correct an oil seal leak and, at the same time, to perform preventive mainter :e. The activity was witnessed by the inspector !
on January 29, 1988, during the mid-shift. The following observa-tions were made, l i
(1) The work was in progress without JT No. CP-727 being at the worksite (Auxiliary Building - 281-foot elevation). The work l was supervised by a maintenance superviso l l
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(2) After the JT was located in the maintenance shop, the inspec-tor's review of the JT indicated that. none- of the .four proce-dures included in.the work package:had "verified" stamp assur-ing verification of the controlled work procedures and this- :
was contrary to Administrative Procedure-(AP) 1001G require-ment (3) The JT did not indicate current revisions of any.of the-four
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procedures as required by MP 1_407- (4) One of the procedures required recording'some data, such as pump coupling align' ment, as well'as certification of the torque wrench. Since the work package was not at-the worksite, it was not clear how this data would be documented. The inspector
] reviewed the final package. The inspector was informed that
the required data will be transferred from the working docu-ments to the controlled copy following. completion of the jo i 3.2.4 Emergency Diesel Injector Assemblies The emergency ~ diesels have two injectors on each cylinder (24 on each diesel). The inspector observed two different types of injec--
tor assemblies on both diesels. The controlled technical manual-shows only one type. The licensee was informed of thi.s discrepancy immediately. Based on licensee's discussion with the vendor, it became clear that both types of injectors are acceptable; however, i the type identified in the technical manual is not the latest mode !
Licensee also has requested the vendor to provide a formal change ;
notice to the manua i The diesels have been tested satisfactorily with both types of in-
". jectors installed. However, the apparent discrepancy was not iden-tified during the recent inspection or during the installatio L
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The injectors were accepted during the recent inspection based on vendor's certificate of compliance, as well as the part number
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identified on the purchase order. Physical verification of the i parts against the approved applicable drawings or comparison against !
the installed units would be an enhancement to identify needed- '
changes to vendor documents. The inspector had no additional com-
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ments on this matte !
l 3.2.5 Waste Disposal pipe Repair I During the week of January 19-22, 1988, the inspector reviewed con-trol room activities and the. Job ticket record associated with the repair to a leaking pipe associated with WDL-V61, Boric Acid Make-Up i System Isolation Valv Two job tickets were involved, one (JT N CP-133) was for the pipe replacement and associated cutting and welding; and, the other (JT No. CP-685) was ~to set up special plant conditions and for restoration to normal. The special plant condi-
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tions were needed because of the difficulty in isolating this sec-tion of pipe. A temporary modification was implemented to use a check valve as an isolation valve. Further a seventy-two hour action statement of the Technical Specifications (TS) (Section 3.2.2.c) needed to be entered because both alternate sources of boric acid make-up (in addition to the Borated Water Storage Tank)
to the reactor coolant system had to be isolate Overall, licensee performance in implementing this work was quite good. The JT CP-685 methodologically set up the isolation bounda-ries, leak checked the affected pipe sections for effective isola-tion, and provided necessary precautions and limitations to opera-tors and workers before actual replacement work was started. Leak-age was detected and the tagout isolation was properly revised to reflect additional valve isolatio No TS action statements were violated. There was substantial quality control involvement and documentation on these activities with no significant finding The safety evaluation for the temporary modification (TM No. 23, dated January 19, 1988) addressed appropriate potential material concerns with the temporary use of a bonnet for the check valve isolation functio Plant conditions were restored to norma The above-noted job ticket packages were still in "the review for completeness phase" two weeks after completion of work. The situ-ation reflected similar administrative control discrepancies as noted elsewhere in this report. The operability / test block was not signed off by the operations department. This verification was accomplished by other means upon tag clearance immediately after job completion. A licensee management representative acknowledged the inspector's observation and indicated that the sign-off should have occurred when operations accepted the components back and re-lied on that action for operabilit The licensee representative agreed to review this matter. The inspector had no additional comment .2.6 Surveillance Observation On February 5, 1988, the inspector witnessed surveillance and test-ing of the emergency feedwater pump per Surveillance Procedure (SP)
1300-3G A/B. The activity was being performed in accordance with the procedure and the communication among the personnel at the pump location, as well as the control room, was ef fective. Also, testing of the check valve MS-V-9 was witnessed per JT Nos. CP-856 and CP-86 Both activities were well don .2.7 Secondary Plant Drain Cooler Foundation I
During the routine plant tour, the inspector found a cracked foun-dations supporting drain cooler "A" in the turbine building, 305- ,
foot elevation. On further examination, it was noticed that all l
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the baseplate. bolts for that foundation were loose. -The equipment is not safety-related, but its failure could result in a secondary plant transient. The matter was reported;to the licensee'represen- .
tatives. The licensee representative initiated a. review of the matter by forwarding it.to plant engineering. No further NRC staff l
- action is required on this matte .
3.3 Summary of Maintenance Observations !
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Paragraph 3.2.1.3(2) identified the installation of the. wrong expansion !
joint on the reactor building emergency cooling system. pump RR-PI !
Paragraphs 3.2.1.3(5) and (6) identified improper component identifica- !
tion, as well as improper torquing of the expansion joint, although in- ,
stallation of the joint was correct on the pump (RR-P1A). Paragraph l 3.2.2(1) and (2) identified the cleaning of the station battery terminals ;
without the job package at the site and without proper procedure change- !
control. Paragraphs 3.2.3(1) and (2) identified the repair on the liquid ;
waste transfer pump (WDL-P78) again without the job package 'at the site !
, and, also, without verified controlled procedures attached to the J t These examples collectively indicate an apparent violation of 10 CFR Part ,
50, Appendix B,. Criteria V and of licensee's NRC-approved Quality Assur- ;
ance Plan, Section 3.0 and ANSI 18.7-1976, Section 5.3 (289/88-01-01). :
At the interim exit meeting of February 1, 1988, the' inspector discussed l proposed licensee corrective actions on their findings (paragraph 3.2.1.2). t Planned action at that time appeared not to be tangible in that most of l the actions were oriented toward counseling personnel. As a result of )
the inspectors' observation and follow-up review, there appeared to be l weak areas within related maintenance / administrative control procedure ;
These areas dealt with the control of applicable /non-applicable work instructions and on the control of procurement and replacement of parts j in the plant to preclude such errors as noted abov Licensee management !
acknowledged the inspectors' comment l The other inspector observations addressed in this'section deal with !
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attention to detail in strictly adhering to related administrative con- ,
trol in the maintenance area. No additional NRC staff action on these i matters is warranted at this tim .4 Maintenance Summary ;
The licensee management involvement and responsiveness was noteworth I The continued trouble-free operation of the plant without any unplanned i maintenance outages is indicative of an overall effective maintenance ;
program. The negative findings described above indicate procedural weaknesses in certain sub-areas of the maintenance progra Howeve r ,-
the licensee, in addition to their own actions, plan to utilize indepen-dent resources to further strengthen their current maintenance progra ,
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4.0 Reactor Building Sump Low level Interlock Modification 4.1 Background / Existing Design -
The inspector reviewed applicable documents that addressed the licensee's action on the reactor building sump low level interlock modification (BA No. 413916). This modification installed the necessary hardware for lowering the reactor building (RB) sump minimum leve Valves WDL-V534/535, isolate or permit flow from the RB sump to the auxiliary building sump. The low level interlock closes this valve at a preset level and it was reset to 15 inches (previously 45 inches).
The low level setpoint alarm (B-3-3 in the control room) was reset to 18 inches (previously 48 inches). The low level setpoint interlock function has been removed from level switches LS-116 D&E and are now provided by the LT-804 instrument channel. The basis for this change was to provide an instrument channel which could accommodate a variable setpoint capabilit The level transmitters are safety grade and provide indication from 0-90 inche The purpose for this modification is to allow more effective use of the miscellaneous waste evaporator by allowing additional water to be pro-cessed, thereby reducing the cycling of the miscellaneous waste evapora-to The basis for the 15-inch setpoint is to maintain the 6-inch drain line from the RB sump to the auxiliary building sump covered with wate This will prevent RB gaseous atmosphere from entering the auxiliary building and causing a radioactive gas hazard during normal or emergency operations while still maintaining a Net Positive Suction Head for decay heat removal pumps on post-accident long-term recirculatio .2 Acceptance Criteria / Scope of fieview The purpose of this review was to:
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ensure changes have been reviewed and approved in accordance with 10 CFR 50.59;
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verify that design changes were reviewed and approved in accordance with Technical Specifications and established QC/QA controls;
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verify that design changes were controlled by approved procedures;
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verify that the licensee conducted a review and evaluation of test results in a reasonable time frame;
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verify operating procedure modifications were made and approved in reasonable time fram . .
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verify operator training programs were revised prior to the modifi-cation being declareo operable; and,
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verify that as-built drawings were changed prior to the modifica-tions being declared operabl The inspector reviewed the following documentation during this inspection:
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Final Safety Analysis Report, Sections 7.3.2.1 (c), 11(b), 6.4.2(e),
and 6.2.2;
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Technical Specifications Section 4.1;
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Operating Procedure (0P) 1104-40, Revision 21, dated June 12, 1987,
"Plant Sump and Orainage System;"
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Surveillance Procedure (SP) 1302-5.25, Revision 10, dated May 6, 1986, "RB Sump Level;"
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OP 1101-4, Revision 58, dated November 16, 1987, "Sumps and Drains;"
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Administrative Procedure (AP) 1043, Revision 12, dated March 12, 1987, "Control of Plant Modifications;"
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Operations Plant Manual, Section P-1, Revision 10;
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Training Handout, dated December 9, 1987, "RB Sump Lo Level Setpoint Change;" and,
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Installation Specification for WDL-V535 Low Level Interlock Modifi-cation, Revision 0, dated July 2, 198 The inspector verified that the control room annunciator and procedures were changed to reflect the setpoint change and that the operators were aware of the change. The inspector also inspected the RB sump monitoring cabinet, which houses the LT-804 instrument channel alarm and interlock modul .3 Findings / Conclusions The inspector found all documents to be adequately reviewed and approved, as well as technically correc Test data was found to be within the l acceptance criteri The written basis for the 10 CFR 50.59 safety i evaluation was found to be technically correct and the questions neces- ,
sary to determine whether the change constitutes an unreviewed safety !
question have been considered by the licensee. The design change was !
incorporated into applicable procedures and the FSAR. As-built drawings l were verified to be updated prior to the modification being declared
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operable. ' Training was completed.in a timely manner. The Fire Hazard Analysis Report (FHAR) _was found to be complete and accurate for .this modificatio ~
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In conclusion, this modification was performed according to licensee i administrative procedures, as well as Technical Specifications and QA/QC controls. Appropriate _ reviews-and approvals were evident. Procedure i reflected'the setpoint change in a timely manner. The installation-and
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-turnover'of this modification was found to be in accordance with proce- :
dure .0 Safety Issue Management System Item Verification
' Introduction ;
The inspector verified proper implementation, on a sampling basis, of l licensee actions related to the below-listed NRC Safety Issue Management System (SIMS) items. The generic irspection approach for a SIMS. item ;
was:
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research various licensee and NRC correspondence,. including safety evaluation reports (SER's) to identify key assumptions, commitments, t or other licensee actions to be taken to resolve the safety issue; :
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identify any additional items which need to be verified as deline-ated in the related NRC Temporary Instruction or other inspection procedures;
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verify proper implementation of the items planned above; and, !
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assess licensee performance related to that implementation and re-j l lated to dissemination of the issue and-its resolution to licensee ;
personnel who need to know, such as by procedural upgrading and
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trainin l
5.2 Reactor Vessel Overpressure Protection (SIMS No. A-26) j 5. Background ;
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The "Technical Report on Vessel Pressure Transients" in NUREG 0138, !
November 1976, summarize's the technical considerations and,-also !
addresses the safety concerns related to the overpressure protection of pressurized water reactor,(PWR) vessels at low temperature. The i NRC letter dated August 11, 1976, requested the licensee (then Met-
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Ed) to design and install necessary modifications to mitigate the j consequences of pressure transients at low temperatures. The'1etter also required the licensee to implement proper administrative con- l trols as an interim measure until the hardware changes-were com- .
pleted. The licensee submitted a proposed action plan on October ;
9 15, 1976. Following NRC reviews and in response to additional j 4 ;
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safety concerns, the licensee upgraded their action plan and also prepared other relevant documentation, including Technical Specifi-cation Change Request No. 74, dated March 13, 1978. The technical specification was revised through Licensee Amendment No. 56, dated March 28, 198 The safety evaluation of licensee's actions on this safety issue was performed by NRR on July 28, 1980, and licensee actions were found to be adequate, complete, and in compliance with the design criteria specified in 10 CFR Part 50, Appendix The design cri-teria for the system performance included: (1) necessary operator action; (2) single failure susceptibility of components; (3) test-ability of the system; and, (4) seismic and other safety grade criteria (IEEE Standard 279 requirements).
The NRC safety evaluation also assessed licensee's evaluation of design basis events and mitigating controls'.
5. Findings / Conclusions The inspector verified that the licensee's design documents reflect all the hardware and procedural changes as identified in the above referenced safety evaluatio Key design aspects were: (1) an alarm if the system pressure exceeds 485 psig (manually enabled) when the temperature is less than 275 F; (2) an alarm if high pressure in-jection (HPI) valves are not racked-out at temperature less than 275 F; (3) an alarm associated with pressure levels and the system pressure; (4) an interlock on the core flood tank discharge valve so that they will not open until the system pressure is reduced to 600 psig; (5) an alarm indicating position of the pressurizer relief block valve; and, (6) testing provision for these functions. The cooldown and start-up procedures include the necessary steps to reflect the above-noted desig The inspector also verified that the licensee's training program indicated that appropriate training was completed. Records on this matter were quite extensiv All the requirements related to the overpressure mitigating system were satisfactorily complete !.3 Natural Circulation Cooldown (SIMS No. MPA-B-66)
5. Background While St. Lucie Unit I was cooling down under natural circulation conditions on June 11, 1980, flashing of coolant produced a void in the reactor vessel upper head, forcing water into the pressurize The reactor was successfully brought to cold shutdown. Based on the NRC review of the event, MPA item B-66 was initiated. This MPA
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requires that all pressurized water reactor's (PWR's) implement procedures and training programs to ensure the capability to deal with such events. Licensees were requested by Generic Letter (GL)
No,'81-21-to provide an assessment of-their facility procedures and training program, including:
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a demonstration (e.g., analysis and/or test) that controlled = t natural circulation cooldown from~ operating conditions to cold shutdown conditions, conducted'in accordance with plant.proce-dures, should not result in reactor. vessel' voiding;
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verification that supplies of safety grade auxiliary feedwater are sufficient to support plant cooldown methods; and,
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a description of plant training programs and the provisions of emergency procedures (e.g., limited cooldown rate,' response to rapid change in pressurizer ~ level) that deals with preven-tion or mitigation of reactor vessel voidin ;
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Licensee responses to this issue are embodied in the following let-ters: December 7,1981; July 20,1983; April 4,1984; June 26,1984; and, July 24, 1985. The NRC safety evaluation report (SER).was-issued June 5, 1984. Previous inspections (NRC Report Nos. 50-289/
84-31 and 85-17) confirmed licensee procedural commitments in re-sponse to this issue. However, both the NRC SER and inspection reports noted that additional analysis was needed in order to de- :
termine a maximum cooldown rate (beyond 10 degrees F/hr) for which a void would not form at the' reactor vessel upper head (RVUH). The additional analysis was completed and documented in the latest licensee letter (July 24,1985). '
The results of this review are addressed belo I In conjunction with this inspection, the following procedures were reviewed on a sampling basis:
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Operating Procedure (0P) 1102-11, Revision 69, effective
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August 18, 1987, "Plant Cooldown;" and,
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The Abnormal Transient Procedure (ATP) Series 1210-1 to 1 For further guidance to the inspector, NRC staff issued Temporary Instruction No. 2515/86, dated April 7, 1987. This inspection
- documents the review required by that T .3.2 Natural Circulation Demonstration The licensee opted to address this concern by computer analysis, rather than actual plant tes Licensee preliminary analysis was reviewed and found to be acceptable as noted in the NRC scaff SER.
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The revised analysis (letter, dated July 24,1935) identified that a void would not form at the RVUH area provided that the cooldown was limited to no greater than 50 degrees F/br and reactor coolant system (RCS) pressure and temperature were held above a minimum pressure versus temperature curve (MPTC). This curve is a combina-tion of the natural circulation cooldown curve and fuel pin com-pression curve for natural circulation which already existed as a plant limiting condition. There is a section of the pressure versus temperature domain that the natural circulation curve is more re-strictive than the fuel pin compression curv The inspector noted that OP 1102-16, "Natural Circulation Cooldown,"
was cancelled and its control measures were properly incorporated into a revised OP 1102-11, "Plant Cooldown." In particular, the latest MPTC was properly reflected in OP 1102-1 However, it was notea that the Abnormal Transient Cperating Guide (AT0G) procedure 1210-10, Revisior 15, effective January 13, 1988,
"Abnormal Transients Rules, Guides, and Graphs," did not address the MPTC for natural circulation, nor did it address the special measures to prevent a void formation in the RVUH. Upon further :
review, the licensee's letter of June 4, 1984, introduced termin-ology that the naturul circulation restrictions established by analysis apply to non-emergency situations. Non-emergency situ-ations are not clearly define Further, this appears to conflict with emergency procedure scope in that a loss of reactor coolant pumps (RCP's) is covered by emergency procedure; namely, 1202-14,
"Loss of RC Flow /RC Pump Trip," along with the ATOG series. Also, the staf f's SER does not make the above-roted distinction. It would appear that the reality of how the licensee would manat.e all design basis events involving natural circulation (NC) is not accurately ,
reflected in the staff's SE Further, the following ATOG procedure would direct the operator to a natural circulation cooldown and/or the use of Operating Procedure (0P) 1102-11, Revision 70, effective December 11, 1987, "Plant Cooldown:" 1210-5, Revision 12, effective December 11, 1987, "0TSG Tube Leak / Rupture;" 1210-6, Revision 9, effective December 30, 1987,
"Small Break LOCA Cooldown;" and, 1210-9, Revision 11, effective December 30,1987, "HPI Cooling-Recovery from Solid Operations."
Paragraph 3.18 of ATP 1210-5 directs the use of OP 1102-11 but, also, indicates "do not limit cooldown to 50 F/ hour." Paragraph 2.13 of ATP 1210-6 assumes reactor coolant pumps (RCP's) are off, does not reference OP 1101-11, and requires that a cooldown of 100 F/ hour be established. Paragraph 2.10 of ATP 1210-9 directs the use of OP 1102-11 without exceptions or restrictions. It would appear that the actions of ATP's 1210-5 and 1210-6 would lead to a void forma-tion at the RVUH during a design basis event in light of the licen-see's analysis on natural circulation cooldow That void formation ,
is not a safety issue as long as it does not interfere with decay ;
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heat removal. More importantly, no notes / cautions exist in ATP !
1210-5 and 1210-6, nor do general rules exist in 1210-10 that.would i-caution the operator about this situation and/or provide the.guid- l ance.of 0P 1102-11 on detecting and' collapsing the void before it !
interfered with natural circulation .for ATOG ~ actions. : j The inspector consulted with the NRR project manager and lead tech - }
nical reviewer on this issue for NRR.- The draft SER recognizes that 1 certain design basis events may cause voids to form in the RVU ;
This information apparently was not transcribed into.the final .NRC- 4 staff SER. Beyond this administrative oversight, the NRR represen- ,
tatives indicated that sufficient guidance must exist for all design- !
basis events in order for the operator.to recognize and take action ie to collapse the void in the RVUH if it were'to form on a-natural circulation cooldow !
The inspector concluded that the AT0G procedures do not adequately ;
provide this guidance either directly or by reference in'ATP 1210-5, ';
"0TSG Tube Leak / Rupture," and 1210-6, "Small Break LOCA-Cooldown."
l The inspector also agreed to further meet with licensee representa- l
tives on this issue in the next inspection period. The results of l that meeting will also be documented. Accordingly, SIMS No. MPA- !
B-66 will remain open but all actions of the related Temporary In- i struction (TI) 2515/86 were completed and the TI will therefore be 1 closed.
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l 5. Safety-Grade Sources of Water for Natural Circulation Cooldown i
The NRC SER accepted the licensee response to this concern. Essen- !
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, tially, the emergency feedwater system (EFW) (now safety grade) ,
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would be used for natural circulation cooldown. Several sources !
of water are available. Condensate storage tanks (CST's)~, condenser ;
J hotwell, on-site one-million gallon storage tank, and an unlimited ,
supply of river water from the reactor building emergency cooling !
water system (last resort). The inspector independently' confirmed )
plant configuration sources as noted in the licensee's response to ;
this ite i
5. Training Program and Implementation l i
The license 2's training program description in their letter response, !
dated December 7, 1981, is minimal. Essentially, it states that [
operators would be made aware of the St. Lucie event,' procedural ;
actions would be reviewed at the 1982 simulator sessions, and !
training on the then applicable procedure would be accomplishe .;
This was acceptable by the NRC staff's SE i
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The inspector discussed this item with training department personnel involved in both classroom and simulator training. The inspector also interviewed four licensed operators associated with two shifts to ascertain their knowledge of the event and licensee procedures for natural circulation cooldow In addition to the above, the inspector reviewed selected sections of the following dccument Exhibit 2 to 7811-PG0-2613, Revision 4, "Licensed Operator Requalification Training"
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7811-PGD-2611, Revision 5, "Replacement Operator Training Program"
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Lesson Plan No. 11.2.01.281, Revision 0, "Natural Circulation"
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Lesson Exercise 11.7,01.015, Revision 0, "Loss of Off-Site Power"
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Lesson Exercise 11.7.03.020, Revision C (Replacement Operator Training Program), "Loss of Off-Site Power" Based on this review, the inspector concluded that the licensee provides appropriate training to their operators on this event but, more importantly, the training was substantial on how to conduct a natural circulation cooldown with respect to the concerns and restrictions imposed subsequent to the St. Lucie event. Operator interviews confirmed general familiarity with the St. Lucie event, but it also confirmed familiarity with the licensee s procedural requirements. The in plant training on natural circulation was extensively reviewed by NRC staff during the TMI-1 Restart process of October-December 198 The inspector had additional observations in this are Certain lesson plans / exercises still reflected the cancelled proce-dure OP 1102-1 The training representatives were aware of that and plans exist to revise these documents to reflect the current procedur The in-classroom lesson plan for operator requalification program was not set on a specified frequenc Plans exist to do so on a frequency yet to be determined by the licensee. A new cycle of training on this plan is set for 1988 Cycle It was last com-pleted in 1984 and during the 1985 in plant training. Simulator training on natural circulation is usually provided during the loss of of f-site power drill l The inspector had no residual concerns in the training area for this SIMS ite * O
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5.4 SIMS Issues Summary For the two SIMS issues reviewed i .e, the licensee appropriately translated generic licensing actions into plant specific action Specifically, they properly incorporated requirements and commitments into procedural requirements and/or training plan / exercise elements, except as noted above for the residual issue dealing with natural
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circulation as described abov .0 Licensee Actions on Previous Inspection Findings !
6.1 Introduction For these items listed below that were previously identified violations, the inspector reviewed the licensee's response and corrective to:
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verify the licensee responded in a timely manner;
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verify measures taken to correct the item and avoid recurrence were completed and within the specified time frame; and,
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verify licensee commitments were complete .2 (Closed) Violation (289/86-06-01): Failure to Properly Implement Facility Procedures The licensee responded to this violation on August 11, 1986, and provided a supplemental response un August 26, 1986. Details of NRC evaluation -
of the licensee's responses are documented in NRC Inspection Report N /86-17. The licensee further responded to this violation in a letter dated March 5, 1987. A meeting between the licensee and NRC was held on June 8, 1987, to discuss the licensee's responses to this violatio Tne details of this meeting are provided in NRC Inspection Report N ;
289/87-0 .
I Licensee corrective actions have been verified in these previous inspec-tion reports. Due to an oversight in Inspection Report No. 289/87-08, this item was not closed. This item is, therefore, being closed in this inspection repor !
l 6.3 (Closed) Violation (289/86-12-02): Inadequate Safety Evaluation for (
Change to Procedures Described in FSAR and Modifications and Single l Failure Analysis f or Back-Up Instrument Air ;
i The comprehensive single failure analysis on the air supply system to the ( ergency feedwater valves is discussed in NRC Inspection Report N /. 06 and currently remains as an open item (289/67-06-08).
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The emergency ~ feedwater pump Surveillance Procedure (SP) 1301.11.42 has
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been revised and reviewed in NRC Inspection Report No. 289/87-09. -Ad-
~ditional review of safety evaluations and design verification was com- *
pleted in NRC Inspection-Report Nos.. 289/87-08 and 87-1 The inspector reviewed the revised procedure for installation of tempor--
ary shielding (9100-IMP-3282.01, Revision 2, dated November 3, 1986) and performed an in plant review to inspect temporary shielding for conform-ance to this procedure. -This revision has incorporated all the correc-tive measures discussed in the licensee's response to the violatio The inspector verified that temporary shielding was in conformance with Radiological Controls Procedure 9100-IMP-3282.01. Upon discussions with cognizant licensee personnel, it was determined that temporary shielding is reviewed / approved on a -semi-annual basis. - Approved methods exist .for . .
determining the acceptability for continued use of existing temporary shielding. It is noted that the use of- temporary shielding .is not ex-cessiv The determination to maintain a temporary shielding instal'lation is made-by the radiological engineer who is responsible for the installatio Technical Functions Work Requestse a e written for -installation existing beyond one refueling outage, currently. eighteen months, and a cost / bene-fit analysis is done and the installation is dispositioned in accordance with the results of-this analysis. The inspector had no further question This item is close .4 (Closed) Violation (289/86-12-08): Failure to Take Prompt Correctiv_e_
Action on known Conditions Adverse to Quality The conditions adverse to quality were repetitive examples of out-of-specification log readi tgs and of drawing deficiencies.. The. inspector reviewed auxiliary operator (AO) log sheets for the period December 28, 1987, to January 9, 1988. Out-of-specification data for important-to-
safety sy; . ems were properly identified on the -log' sheets. There was not an excessive amount of log entries which were out-of-specificatio Appropriate reviews of the log sheets were evident. This portion of the violation is close The control room drawing deficiencies and GPUN drawing procedure revision have been reviewed and closed in NRC Inspection Report No. 289/87-2 Corrective measures taken are adequate and have been implemented in a timely manner. This item is close , , _ , , _ , , _ - . _ __ _ . _ , . _ . _ . _ _ __ __ __
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6.5 (Closed) Violation (289/86-17-02): Failure to Follow- Procedures Associated with Engineered-Safeguards' Actuation System Testing-In a meeting of June 8, 1987,1this item was discussed. The detatis of this meeting and topic were documented in NRC-Inspection Report N /87-0 The inspector verified that crew members were briefed and understood the minimum requirements for independent verification. Corrective actions were completed in a timely manner and adequately addressed the' violatio .6 (Closed) Violation (289/86-17-11): Source Range Instrumentation Inoperable Without a Proper Safety Evaluation The licensee provided a response to this violation in a letter dated March 5, 1987. The corrective actions provided in that letter were determined acceptable in NRC Inspection. Report No. 289/87-0 The inspector verified that the corrective actions discussed in the lic- ;
ensee's response have been implemente The individuals involved have '
bean instructed in the proper methods of altering plant systems. All Instrument and Control (I&C) technicians have been instructed in the.
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proper use of AP 1013, Revision 25, effective date October 28, 1987,
"8ypass of Safety Functions and Jumper Control." A sioilar problem with the improper implementation of AP1013 was identified in NRC Inspection Report No. 50-289/87-19. The lice'nsee is currently preparing a training !
seminar on the general use of AP 1013 which will_ provide _several days '
of classroom instruction. This seminar should be provided within the next month. Follow-up actions will be reviewed under violation 289/
87-19-0 Difficulty in verification of training records was experienced. The y licensee could not locate the training records of the I&C technicians j for AP 1013 (spacific to this violation). The inspector verified train- '
ing records for past violations were readily available'. The I&C super-i visor then agreed to re-instrtet the I&C technicians on the proper use of AP 1013 and provide verification of such training. This has been
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verifie The corrective actions taken are adequate and have been verified to be complete. This item is close .7 .( Closed) Violation (289/87-09-04): Pressurizer Platform Installation
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Design Modification Failure to Properly Review and Verify Design l
The inspector verified GPUN Specification SP 1101-32-022, Revision 1,
"Nuclear Grade Fiberglass Insulation Systems," incorporated the neces-sary changes to avoid further occurrences of this type. The corrective actions were incorporated in a timely manner and all commitments were completed, i
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The changes incorporated included sections to clarify thermal growth and obstruction clearance requirements. An evaluation is now required prier to installation of any replacement insulation which reduces clearances between adjoining or adjacent component Quality control verification is also required. Drawings showing items to be insulated that could produce irterferences due to thermal movement or growth must also be
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provide t These measures are adequat This item is close .8 (Closed) Unresolved Item (289/87-09-05): Licensee Incorporate Fcur Channel Level Cnecks into Site Procedures *
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This issue concerned the fact that not all four channels of Once-Through Steam Generator (OTSG) start-up and operating range level transmitters for the new Heat Sink Protection System (HSPS) are provided with indica-tion in the control roo The licensee was considering discontinuing .
the four channel compariso However, there were several level trans- !
mitters problems since initial installation. Further, NRC staff gave the licensee a positinn that four channel checks were needed to meet applicable TS. Accordingly, the level comparison was incorporated into e formal Surveillance Procedt e (SP) 1301.4.1, Revision 39, "Weekly Checks."
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The inspector had reviewed the data compiled via this document in NRC Inspection Report No. 50-289/87-24 and it was satisfactory. The licensee plans to continue recording and comparing the "blind" OTSG level trans-mitters until permanent corrective action is taken. At present, the .
inspector's concern about the "blind" 0TSG level transmitter channel checks has been resolved and this item is close .9 (Closed) Unresolved Item _(289/87-13-011: Licensee to Resolve Improper i Installation of Anchor Bolts in the Chill Water System fr' Control '
Building Ventilation System Licensee actions related to this item were reviewed in a region-based ;
inspection (Report No. 50-289/87-18). However, due to an administrative oversight, this item was not closed in the report. Accordingly, the item is closed administratively by this report.
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6.10 (0 pen)_NRC Temporary Instruction 2500/26 (289/25-00-26): NRC Bulletin 87-02, Fastener Testing l l
The licensee completed selection and testing of a representative sampie !
of fasteners (studs, bolts, and nuts) to comply with the requirements i of NRC Bulletin 87-0 I The inspector participated in the initial selection process, but the licensee later determined that some bolts / nuts selected for the "r.ot important to safety" classification were not traceable. Another selec-
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tion of bolts / nuts was made where material identity was known. The lic-ensee subsequently informed the' inspector of this action and also noted this in the initial response to the bulletin in a letter dated January !
15, 1988.- 1 ;
The inspector reviewed the types of bolts / nuts selected for the alternate
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. sample and concluded that the sample selection was made in an appropriate i manner. This item remains open-pending review of the licensee warehouse--
- program and.NRR-comment on the test result '
6.11 NRC Information Notice (IN) No. 85-38 (289/85-IN-38): Lo'ose Parts Obstruct Control Rod Drive Mechanism This IN was issued to alert Eabcock & Wilcox (B&W) licensees of a poten-tial problem with control rod binding due to failure of a' leaf spring i in one control rod drive mechanism (CRDM) at Davis-Besse. The potential safety issue was also addressed in:an NRC letter cated December 4, 1985 ;
(OD 85-19) to a 10 CFR 2.206 petitioner who called for a shutdown of all
- . B&W-designed reactors. The petitioner's request was denied based on the NRC staff's SER attached to the above-noted letter. The licensee at TMI
? had used a different orocedure/ process for control rod unlatching and initially determined that the inspection recommended by the IN was.not necessary. However, as noted in the subject SER, the licensee committed to verify tne position of their leaf spring at the next refueling outag '
, Subseqt.ently, the licensee, with B&W assistance, performed; the recom- l mended inspection of the leaf spring assembly. This was performed in April 1986 during the "5M" eddy current outage. The inspe: tor reviewed i the results of the inspection and discussed the evolution with licensee engineering persennel. The work was completed in accordance with JT N CH-25 There were no problems similar to that for the Davis-Bess ~
facility. The inspector concluded that the licensee had taken adequate ccrrective action in response to IN No. 85-38 and the NRC staff's request
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as noted in the stibject SE ,
6.12 Past Inspectien "indings Summary
- The licensee took appropriate corrective action snd/or fulfilled commit-i ments made in response to the above-nated violations. Generic licensing actions were effectively translated into plant specific action, along
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, with meeting commitment date Licensee representatives cooperated fully with NRC staff on the bolt sample selection process-for NRC Bulletin N . The licensee provided necessary information to resolve the sub- ;
ject unresolved items.
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- 7.0 Exit Interview
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l- The inspectors discussed the inspection scope and findings with licensee man-acement at an interim exit meeting on February 1, 1988, on the maintenance
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area and at a final exit meeting conducted February 9,1988.. Senior licensee i personnel attending the final exit meeting included the following:
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T. Broughton, Operations and Maintenance Director, TMI-1 J.'Colitz, Plant Engineering Director, TMI-1 J. Fornicola, Manager, Quality Assurance Modification / Operations H. Hukill, Director, TMI-I D. Shovlin, Plant Material Director, TMI-1 C. Smyth, TMI-1 Licensing Manager P. Snyder, Manager, Material Assessment A representative of the Commonwealth of Pennsylvania, Ajit K. Bhattachayyra, also attended the meetin The inspection results as discussed at the meeting are summarized in the cover page of the insper. tion 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 deviation Unre-solved items discussed during the exit meeting are addressed in Section I
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ATTACHMENT 1 NRC INSPECTION REPORT NO. 50-289/88-01 ACTIVITIES REVIEWE0 Plant Operations
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Control room operations during regular and back shift hours, including fre-quent observation of activities in progress 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
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Areas outside the control room
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Selected licensee planning meetings
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Plant transients of January 18, 22, and 28,1988
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Partial loss of instrument air on January 20, 1988 During this inspection period, the inspectors conducted direct inspections during the following back shift hour Day /Date Time Wednesday, 1/13/88 3:00 a.m. - 7:00 Week of 1/18-22/88 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> between 6:00 a.m. - 7:00 Sunday, 1/24/88 12:30 p.m. - 2:00 :00 p.m. - 10:30 Friday, 1/29/88 3:00 a.m. - 7:00 Maintenance / Surveillance
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Expansion Joints for the Reactor Building Emergency Cooling System Pumps - 1 JT Nos. CK-211, CK-212, and CP-683 l
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Cleaning af the Station 3attery Terminals - JT No. CP-603
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Liquid Waste Transfer Pump (WDL-P78) Repair - JT No. CP-727 l
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Waste Disposal Pipe Repair - JT Nos. CP-133 and CP-685 l
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Surveillance Observations - JT Nos. CP-856 and CP-869 l
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- Attachment 1 2
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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 belo ;
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TABLE RCS LEAK RATE DATE All Values GPM DATE/ TIME (NUREG 1107) CORRECTED DURATION Lg N Lg -
G "U "U 1/14/88 0.4279 0.43 -0.07 0.03 0.0365 0220 2 Hours 1/17/88 0.4071 0.41 0.03 0.13 0.1346 0030 2 Hours
- 1/19/88 -0.1845 -0.18 -0.93 -0.83 -0.8232 0917 2 Hours 1/21/88 U.3631 0.36 0.12 -0.02 -0.0143 1803 2 Hours 1/22/88 0.4616 0.48 0.27 0.37 0.3581 1542 2 Hours 1/26/83 0.1817 0.18 0.01 0.11 0.1187 1530 2 Hours 2/4/88 0.4707 0.47 0.01 0.11 0.1151 l 0048 l 2 Hours G = Identified gross leakage U = Unideni'fied leakage L - Licensee calculated N = NRC calc lated
- Declared invalid by licensee due to water additi n to make-up tan Columns 2 and 3, 5 and 6 correlate + 0.2 gpm in accordance with NUREG 110 N u
is corrected by adding 0.1044 opm to the NUREG 1107u N due to total purge flow through the No. 3 seal from RCP's.