IR 05000336/1988022
| ML20206D008 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 11/04/1988 |
| From: | Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20206C992 | List: |
| References | |
| 50-336-88-22, NUDOCS 8811160502 | |
| Download: ML20206D008 (16) | |
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U.S. NUCLEAR REGULATORY COP. MISSION
REGION I
Report No.
50-336/88-22
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Docket No.
50-336 License No.
DPR-65 Licensee:
Northeast Nuclear Energy Company P.
Box 276 Hart 7ord,Cf 06101-0270 Facility Name: Millstone Nuclear Power Station, Unit 2 Inspection At: Waterford, Connecticut Dates:
September 13, 1933 - October 12, 1938 Reporting i
Inspector:
Peter J. Habighorst, Resident Inspector Inspectors:
William J. Raymond, Senior Resident Inspector Peter J. Habighorst, Resident Inspector Approved by:
Ot C. b i)*
I4 lBe E. C. McCabe, Chief, Reactor Projects Section IB,
Date Inspection Summary:
9/13 - 10/12/88 (Report 50-336/88-22)
Areas Inspected: Routine NRC resident inspection (157 regular hours, 6 backshift
~'ou~rs} o'f Flani~ operations, surveillance, maintenance, previous identified item:,
h Temporary Instruction (TI) 2515/98, review of Plant Incident Reports (PIRs), Mate-rial Equipment Parts List (MEPL) Discrepancies, periodic reports, and committee activities.
Results: No unsafe conditions were identified.
One violation was identified in-volving procedural adherence during a calibration of the containment gaseous radi-ation monitors (Section 5.2).
Additional follow-up is warranted on resolution of MEPL discrepancies (Section 7.0), and Emergency Operating Procedure (EOP) revisions in response to NRC Inspection 50-336/SP-10 (Section 3.1).
kkE A&K 0500o336 2 001 Joe o
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TABLE OF CONTENTS
PAGE i-1.0 Persons Contacted....................................................
2.0 Summary of Facility Activities.......................................
3.0 Previously Identified Items (92701)..................................
3.1 (Closed)NED 88-10-02: Emergency Operating Procedures (EOPs) Did Not Include a Step to Eliminate Voiding in the RCS Steam
Generator Tubes as Per CEN-152 Guidelines.....................
3.2- (Closed) UNR 88-13-01: Wide Range Nuclear Instrument Environmental Qualification and Justification for Continued Operation.....................................................
3.3 (Closed) IFI 88-10-01: Evaluation of E0Ps to Clarify the Proper Temperature Band for Utilizing the 30 Degrees F Subcooling Margin / Net Positive Suction llead (NPSH) Requirements for Operation of RCPs.............................................
4.0 Fa c i l i ty To u r s ( 7170 7 )...............................................
5.0 Plant Operational Status Reviews (71707).............................
5.1 Plant Incident Reports ( PIRs)...................................
5.2 Inoperable Containment Gaseous Radiation Monitors (PIR 88-69)...
5.3 Containment Isolation Valve (2-RC-045) Inoperable...............
5.4 Safety System Operability (71710)...............................
5.5 10 CFR 50, Appendix R, Electrical Backfeed Walkiown (64100).....
v.0 Surveillance Testing (61726).........................................
7.0 Material Equipment Parts List (MEPL) Olscrepancies in the Production Management Maintenance System (PMMS) (92701).........................
8.0 Tsmperary Instruction (TI) 2515/98 On High Temperature Inside Con:ainment/Drywell (71707)........................................
9.0 Refueling Outage Meeting (30703).....................................
10, 0 Pe r i od i c R e p o r t s ( 9 0 713 ).............................................
11.0 Committee Activities (40700).........................................
12.0 Maintenance (62703)..................................................
13.0 Temporary Instruction TI 87-07, Battery Audit (62705)................
14.0 Manag(eent Meeting (30703)...........................................
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DETAILS
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1.0 Persons Contacted
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Inspection findings were discussed periodically with the below persor.nel.
S. Scace, Millstone Station Superintendent
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J. Keenan, Unit 2 Superinter, dent J. Riley, Unit 2 Maintenance Supervisor 0. Kross, Unit 2 Instrument and Controls J. Smith, Unit 2 Operations Supervisor The inspector also contacted other Operations, Radiation Protection, Chemistry,
Instrument and Control, Maintenance, Reactor Engineering, Station Services
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Engineering, and Security personnel.
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2.0 Summary of Facility Activiti x
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The uqit began the inspection period at full power and remained at this level
for the duration of the period.
However, during the period the unit's full
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power output decreased approximately 12 Kde. The licensee identified fouling in the feed flow venturi and the resultant impact on the secondary plant i
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calorimetric calculation as the rrimary cause. Another contributor was a tube l
leak in the 'A' Motsture Separator Reheater (MSR) second stage heat exchanger.
The licensee continues to track MSR performance and feedwater venturi fouling.
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t On October 5, the itcensee's Safety System Functional Inspection (SSFI) team
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presented their self-assessment audit results to licensee upper management.
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Numerous audit observations were made; no immediate action items were identi-
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fled.
The SSFI inspection findings will be presented to the NRC at the t'
Region I Office on November 29, 1938.
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I Near the end of the inspection period, the licensee conducted pre-refueling i
outage work on the intake screen bays. Other items of interest during the (
period were: inoperable containment gaseous radiation monitors (Report Detail
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5.2), TI 2515/98 Containment Temperatures (Report Detail 8.0); and MEPL/PMMS
Incorporation Discrepancies (Repoit Detail 7.0),
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j 3.0 Previously Identified Items (92701)
Closed) NED S8-10-02: Emergency Operating _ Procedures (EOPs) Did Not
(iETude 'a Step to Eliminate Voiding in the RCS Steam Generator Tubes 3.1 l
as PAREif-162 GMines
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t This item concerned a January 30, 1985 licensee letter regarding the Millstone Unit 2 Procedure Generation Package (Rev.1).
By this letter,
the licensee committed to conform to the guidance provided by NUREG 0399
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j and the Combustion Engineering Energency Procedure Guidelines (CEN-152),
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and to implement an E0P verification program to evaluate the correctness d
of the procedures and to ensure the generic and plant specific technical information had been incorporated.
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CEN-152 Rev. 2 provides guidance on steam generator U-tube voiding and subsequent elimination in section 5, "Loss of Coolant Accident Recovery Guideline," and Secticn 6. "Steam Generator Tube Rupture Recovery Guide-line." NRC specialist Inspection Report 50-336/88-10 reported the lic-ensee did not incorporate this guidance into plant-specific E0P 2532,
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"Loss of Primary Coolant," and E0P 2534, "Steam Generator Tube Rupture,"
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or justify the omission of this guidance.
The licensee responded to Report 50-336/88-10 by letter dated September
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22, 1988. The licensee's reported corrective action was incorporation of steam generator U-tube void elimination into E0P 2532 in July 1988.
I The Plant Operations Review Committee (PORC) approved the change in i
meeting 2-88-131 on August 3.
The inspector verified incorporation into E0P 2532 and reviewed the licensee's t.aining lesson plans to verify the
incorporation of S/G V-tube void information was included in operator
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training.
No inadequacies were noted. This item is closed.
y The licensee committed to incorporate CEN guidance on steam generator U-tube void formation into E0P 2534, "Steam Generator Tube Rupture," by February 1, 1989. The licensee has guidance in E0P 2534 with reference
to E0P 2540, "Functional Recovery," to ensure the Heat Removal Safety i
Function is r.t jeopardized in the interim.
This item is unresolved l
pending determination of the adequacy of licensee actions (UNR 88-22-01).
i 3.2 (C1_oted) UNR 88-13-01: Wide Range Nuclear Instrument Environmental
QualiTication and Justification for Continued Operation (JCO)
i This item concerned a 10 CFR 21 letter f rom Gamma-Metrics (G-M) to the l
NRC on Feoruary 19, 1988. The letter discusseo potentially inadequate
cable soldering techniques by G-M, and questionable environmental quali-q fication of the wide range nuclear instruments (WRNIs). The inspector reviewed and documented licensee actions on the 10 CFR 21 letter in In-spection Report 50-336/88-13, issued July 7,1988. The inspector ques-
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tiened the lack of a Justification for Continued Operation (JCO). The licensee determined that two of the four WRNIs were pot.entially affected
by the condition reported, and all requirements of the Millstone Unit i
2 Technical Specifications were satisfied with the existing installation.
The licensee concluded that no JC0 was needed based on a licensee cor-i porate engineering EEQ evaluation of the two affected WRN! monitors on June 8.
That evaluation concluded there was no operability ouestion
based on the failure mechanism (submerged moisture intrusion at 67 psig),
and that the consequence of failure of the detector cabling (arcing-i induced electrical noise) was an increase in observed neutrun flux indi-
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cation.
The licensee initiated Plant Incident Report (PIR) 83-52 to
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track the potential problem.
The licensee has committed to replace the WRNI in the upcoming refueling outage in February, 1989.
The inspector concurred with the licensee's evaluation.
3.3 (Closed) IFI 88-10-01: Evaluation of E0Ps to Clarify the Proper Tempera-
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ture Band for Ott11 zing the 30 Degrees of Subcooling. Margin / Net Positive luction Head (NPSH) Requirements for Operation of RCPs
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This item concerned a discrepancy between CEH-152, Revision 2, whic5 specifies a NPSH curve entitled "Typical Post-Accident Press;te-Tempera-ture limits" and the licensee's E0P incorporatio.i of the 30 degrees F sub-cooling margin pressure-temperature limits for RCP start NRC specialist team inspection (Report 50-336/88-10) compared the required
minimum NPSH curve to the 30 degrees F subcooled margin curve and con-
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cluded the 30 degrees F subcooled margin is conservative for RCS tempera-l tures greater than 480 degrees F.
For RCS temperatures less trian 480
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degrees F, the pressure for the 30 degrees F subcooling margin does not satisfy the RCP NPSH requirements. Use of the 30 degrees F subcooled
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margin may result in damage to the RCPs below 480 degrees F and thereby degrade the post-accident RCP heat removal capability. The licensee agreed to evaluate the use of a caution ster or otherwise clarify the proper temperature band for the 30 degree subcooled margin.
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I By letter dated September 22, the licensee responded to Inspection Report 50-336/83-10.
The use of 30 degree F subcooling as an acceptance cri-j terion was clarified by a caution step in the E0Ps. The licensee's re-view committee approved this change to the E0Ps in PORC meeting 2-88-113 I,
on August 3, 1988.
The inspecto reviewed the affected E0Ps and verified the caution stepr were incorporated in E0P-2534, Revision 5, "Steam i
Generator Tube Rupture," E0P-2536, "Excess Steam Demand," Revision 4, and Step 3.19 of E0P-2532, "Loss of Coolant Accident," Revision 4.
E0P-i i
2532 requires starting prerequisites for RCPs per Of 2301C, "Reactor l
Coolant Pump Operation," and verifies sufficient NPSH for RCP operation
prior to starting, i
The inspector reviewed training lesson plan RQ2-534-1 for E0P-2534.
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lesson plan has the instructor point out the difference between the RCP l
NP5H curve and the 30 degrees F subcooled margin curve, and between the j
more limiting curve and RCS temperature.
Based on discussions with the licensee, lesson plan RQ2-534-1 was ad-
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ministered for licensed operator requalification training in 1933.
The
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inspector had no further questions.
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4.0 Facility Tours (71707)
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The inspector observed plant ooerations during regular and backshif t tours
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of the following areas:
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Cnntrol Room Auxiliary Building Vital Switchgear Room Enclosure Building
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Turbine Building Intake Structure Diesel Generator Room Control rcom instruments were observed for correlation between channels, proper functioning, and conformance with Technical Specifications. Alarm
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conditions in effect and alaras received in the control room were discussed
with operators.
The inspertor reviewed periodically the night order logs, tagout log, Plant Incident Report (PIP.) log, and bypass Amper log. Discus-
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sions with operation department staff on each of the respective logs took
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place.
No inadequacies were noted.
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Posting and control of radiation, contaminations, and control of high radi-i ation areas were inspected.
DurinC plant tours, icgs and records were re-
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viewed to ensure compliance with station procedures *.) determine if entries
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were correctly made, and to verify correct communication and equipment status.
j Selected aspects of site security were verified to be proper during inspection tours, including site access controls, personnel searchas, personnel monitoring, i
placement of physical barriers, gt.ard force staffing, and response to alarms l
and degraded conditions.
No inadequacies were noted.
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5.0 Plant Operational _ Status Reviews (71707)
l 5.1 Plant Incident Reports I
I The plant incident reports (PIRs) listed below were reviewed during the
inspection period to (1) determine the significante of the events; i
(ii) review the licensee's evaluation of the events; (iii) verify the l
licent e's response and corrective actions were proper; and, (iv) verify
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that the h;ensee reported the events in accordance with applicable re-
quirements, if required. The PIRs reviewed were: number's 88-66, 80-67,
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88-60, 88-69, 88-70 and 83-71.
The following items warranted inspector i
fol bwup.
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PIR 88-69 (Detail 5.2) and PIR 88-71 (Detail 5.3).
f 5.2 Inoperable Containment Gaseous Radiation Monitors (PIR 88-69)
f On Septenber 21 at approxinately 2:00 p.m., with +he unit at 100's power,
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the licensee found two containment gaseous radiation monitors (Ru 81238 i
and RM 8262B) in the "test" position. The licensee returned the two
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radiation monitors modules to "operate." The "test / operate" toggle i
switch is utilized in TS surveillance SP 2404AK, Revision 0, "Containment Process Radiation Monitor Gaseous and Particulate Instrument RM-8123A/B and RM-8262 A/B Functional Test." The gaseous radiation monitors are l
used in applications covered in the following TS Limiting Conditions for
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Operation (LCOs): 3.4.6.1, "Leakage detection systems;" 3.3.3.1, "Radi-t ation Monitoring;" and 3.3.2.1, "Engineered Safety Feature Actuation Sys-
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tem Irstrumentation." The monitors were required to De operable in the ll
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present plant condition (Mode 1) for the TS LCOs concerning containment leakagt detection and radiation monitoring systems. The TS LCOs were
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not entered by the licensee during the established time interval between j
August 23 until September 21; the licensee did not then recognize that
the monitors were inoperable.
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The inspector used the Preventive Maintenance Management System (PMMS)
tracking system to verify no corrective or preventive maintenance was
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conducted on RM 81230 or RM 8262E, between August 23 (completion of SP 2404AK) and September 21. The inspector discussed the noncompliance with (
TS requirements with station management on September 23. The licensee
acknowledged this issue as a noncompliance with TSs 3.4.6.1 and 3.3.3.1.
i The inspector reviewed OP-2619A-1, Revision 6, "Control Room Shift I
Checks." The operators record the chart readings for containment gaseous
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radiation monitors every eight hours.
The inspector noted a dif ference
in chart recorded values when the monitors were in "test" and in "oper-
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ate."
In "opt rate," the mudule records rapid continuous minor fluctu-ations about a defined radiation level, whi!e in "test" the level is well
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defined with no fluctuations.
The inspecter questioned control room I
operators whether radiation module test / operate switches were checked i
routinely or on a shift turnover basis.
The operators responded that
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the switch in question was not routinely -hecked.
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The inspector reviewed the safety significance of inoperable containment I
gaseous radiation monitors.
The gaseous radiation moni',or system is part i
of the containment leakage detection system. Alternat.e systems (con-l tainment sump level system, and containment particula'.e system) are used
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to provide detection of e.ontainment leakage to the op.erators, and were
operable during the time period of interest.
The licensee condurts
periodic containment inspections of potential leakpaths. Monthly at-
power containment entries are conducted to acquire Safety I $ ction tank j
(SIT) boric acid sa ples, Analyses of 12 remote containment gaseous
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samples were made between August 23 and September 21 per the licensee's
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chemistry tracking system.
(The containment particulate and gaseous
samples are analy:ed prior to routine containment venting.)
Inspector
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review concluded there was limited safety significance to the inoperable
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gaseous radiation monitors during this period.
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t The licensee reported to the inspector the root cause for the event was
personnel error during performance of TS surveillance 2404AK on August
23.
The inspector reviewed SF-2404AK and the assactated data sheets on l
August 23. This review concluded the procedural steps were not performed
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as prescribed.
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On July 15, 1938 the licensee responded to routine Inspection Report
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$0-336/88-07. This inspection noted a Violation for failure to follow l
procedures during calibration of the control room radiation monitors, t
The licensee's response described actions to prevent recurrence: the
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licensee committed to enhance awtreness of all Instrument and Control
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personnel by emphasizing proper and accurate data collection and the i
necessity for procedural compliance. The inspector concluded that the radiation monitor inoperability chowed that the licensee had again failed y
to adhere to procedure SP2404AK on August 23.
This is a repeat violation
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(VIO 88-22-02).
5.3 Containment Isolation Valve _(2-RC-045) Inoperable (PIR 88-71)
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On October 8, at 1:30 p.m. containment isolation valve 2-RC-45 did not close electrically during in-service inspection testing under procedure
SP 21133-1.
The licensee entered Limiting Condition for Operation
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3.6.3.1.b.
The TS requirement is to isolate the affected penetration by deactivation of the automatic velve (2-RC-45).
The cause of the failure of 2-RC-45 was its instrument air pressure p
regulator. The regulator adjusts 'nstrument air pressure to 20 psig from i
a header pressure of 100 psig.
Licensee investigatico determined the l
regulator fai'ed high, resulting in the electrical solinoid being unable to stroke against the air header pressure.
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Licensee initial ution was to isolate instrument at. to 2-RC-45 and fail j
the' valve shut per the TS action statements.
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On October 9, the itcensee repaired the pressure regulator using Author-
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ized Work Order (AWO) M2-88-01699, "Replaced Air Regulator." SP-21333 i
was reperformed successfully on 2-RC-45.
The licensee exited the LC0 l
at 2:00 p.m.
The inspector had no further questions.
5.4 Safety Su tem Operability (71710)
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l Emergency Systems were reviewed to verify they were operable in the
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standby mode.
The systems reviewed were the Service Water System, l
Auxiliary Feedwater System, and Containment Spray System.
The status j
of the Fire Shutdown p&iel and emergency diesel generators was also in-i spected.
The review considerW proper positioning of major flow p ch
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valves, proper operation of indications and controir, and visual inspec-tions for proper lubrication, cooling and other conditions.
References
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j used included the Final Safety Analysis Report (FSAR); the Technical j
Sp?cifications; plant instrument and piping diagrams (P& ids) 25203-26015,
"Safety Injection and Containment Spray Systen ;" 25203-26008, "Circu-i i
lating and Service Water, Screen Wash and Chlorination Systems;" 25203-26005, "Condensate and Feedwater;" and Operating Pro:edures (ops) 2322,
I Revision 9 "Auxiliary Feedvater," 2346A, Revision 9. "Emergency Diesel l
Generators." ar.d '!612C-1, Revision 14. "Service Water Facility 1 Valve l
Lineup." Minor valve identification tag discrepancies in the service water system were discussed with and corrected by the licensee. Good
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cleanliness was cbserved in the emergency diesel generator rooms.
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i 5.5 10 CFR 50, Appendix R Electrical Backfeed Walkdown (64100j
On October 7, the inspector, the NRC Licensing Project Manager, and the j
licensee conducted a walkdown of AOP 2579A, "Contingency Fire Procedures for Hot Standby Appendix R Fire Area F-1," Revision 2, Steps 4.14 and 4.15.
The purpose of the walkdown was to measure the time required to restore electrical power during the recovery from a major fire in Fire
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Area F-1 (Millstone Unit 2 Control Room). The walkdown specifically addressed the operations required to energize the three vital 4.16KV
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Millstone 2 buses (24C, 240, and 24E) and one vital Millstone 2 480 volt
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bus (22F) from 4.16 KV Millstone 1 Bus 14H.
The steps to accomplish this I
evolutian are in procedure AOP 2579A, Steps 4.14-4.15.
As part of the demonstration, the inspector observed manual charging and racking up of 4.16KV and 480V breakers at the training facility.
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The licensee assumed the followiny initial conditions prio to the walk-down: Millstone Unit I has shu.vown and experienced a Loss of Normal
Power (LNP); the 14H bus is energized from the Emergency Diesel Genera-l tors; a major fire occurred resulting in evacuation of the Unit 2 control room and loss of the Millstone 2 diesel generators.
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The approximate time interval to accomplish steps 4,14 and 4.15 to /JP i
2379A was 30 minutes; however, the inspector noted plant evolutions and j
conditions during an actual event may vary this time interval.
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hJ inadequacies were noted.
6.0 Surveillance Testing (61726)
i The inspector observed portions of surveillance tests to assess performance f
in accordence with approved procedures and Limiting Conditions of Operation, j
removal and restoration of equipment. 0A/QC involvement, and deficiency review and resolution.
The following surveillance activities were reviewed:
l SP2404AB, "Control Room Ventilation Radiation Monitor Calibration,"
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l steps 6.1.3 - 6.1.7, on October 7, 1988.
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0F-2620, "Control Rod Exercises," and SP-26010 "Main Steam Isolation
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l Valve (MSIV) Testing," on Septenber 16, 1988.
OP2304A, "Isolation of Letdown / Charging Systens," on (eptember 28. 1983.
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j Station Battery Quarterly Surveillance on October 5, 1988.
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No inadequacies were noteo.
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7.0 Material Equipment Parts List (MEPL)_ Discrepancies in the Production Manage-
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l ment Maintenance System (PMMS) (92701)
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10 CFR 50, Appendix B, Criterion II requires the identification of structures,
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systens, and components covered by the license;'s Quality Assurance Program.
The licensee's Quality Assurance Program (NN AP) Topical Report, QAP 2,0 and t
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Apper. dix A outlines the items and systems generically. A specific parts list,
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the MEPL, has been developed for each unit at Millstone Station.
The purpose
of Administrative Control Procedure ACP-QA-4.03.B is to assure the preparation
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and mair.tenance of the MEPL.
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The licensee's PMMS is an automated compute system which stores informatun i
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on specific components installed in Norther t Utilities (NU) plants. ACP-QA-
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4.03C, "Use of PMMS Data Base to Indicate quality Assuranco Program Appite-
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j ability," establishes the method of control and use of the PMMS data base to
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j indicate Quality Assurance (QA) applicability.
The ha-d copy MEPL and the j
associateri component /sys em evaluation remain the 905 rning document to de-
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termine QA applicability. On October 1, 1984 the licensee first utilized the
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PMMS system to generate Authorized Work Orders (AW0s) for Millstone 2 com-r ponents.
The PMMS includes QA indicators that must be applied when mainten-
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i ance is performed.
The PMMS QA field indicator will contain either a "Y",
"N", or "U" defining whether the component is under QA program controls, not j
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under QA program controls, or has yet to be classified.
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Northeast Utilities Service Company (NUSCO) has the responsibility to audit i
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j the PMMS OA indicators against the MEPL at intervals not to exceed the unit's
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i refueling cycle. NUSCO conducted the audit for the first time in 1983.
(NRC
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l review of licensee actions for PMMS audit times at other intervals is de-
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l scribed in routine Inspection Report 50-245/88-12.)
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During mid-August 1988 the inspector was informed by the licensee of the re-i sults of NUSCO's audit of PMMS indicators versus the MEPL listing.
The lic-I l
enste informed the inspector that 83 potential discrepancies exister' for com-f l
porents under the jurisdiction of the Unit 2 Maintenance department., with i
j approximately 630 discrepancies in the Instrument. and Control area.
The dis-
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crepancies irvolved components with "N" or "U" in the PMMS data base that
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should be "Y" per the MEPL evaluation.
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On September 20, the inspector discussed the NtlSCO audit findings with de-l i
partment heads in I&C and in Maintenance. According to the Unit 2 maintenance l
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depart ent heaa, all 83 discrepancies had been resolved. The plant personnel l
verified that, in cases identified by the NUSCO audit, no discrepancies
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existed betn en the MEPL evaluation and incorporation into the PMMS Jata base.
i No maintenance work under a PMMS generated AWO was done under non-0A guidance
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i when the activity was evaluated as QA from the MEPL evaluation, The inspector
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I selected four MEPL evaluations: MP2-CD-196, "Clean Liquid Pad Waste System;"
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I MP2-CD-210. "Reactor Coolant System;" MP2-CD-223, "120 Volt Vital instrument
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A.C.;" MP2-CD-469, "Step Cap and Stop Nut for Se vice Water Motor Actuators,"
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that were identified as discrepancies by the 1988 NUSCO audit The inspector
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reviewed the MEPL evaluattun for appropriate QA classifications.
No inade-
quacies were noted.
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The inspector discussed the audit findings with the Millstone 2 I&C department head. According to the licensee, insufficient manpower to re-evaluate the
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audit findings in ths I&C department exists.
The licent*e implemented a
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short-term program such that, if the PMMS indicator is
"N" or "U" on a AWO,
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j the department will citduct the activity as "Y".
At the end of the inspection period, the licensee identified three components marked as "N" and conducted repairs as if under a QA category during a two week period.
The licensee is
actively involved in acquiring a contract vendor engineer to evaluate the j
discrepancies in MEPL/PMMS evaluations.
The inspector will continue to follow licensee actions in the !&C Department as it pertains to PMMS/MEPL evaluations.
The licensee plans to re-audit the data base using current (June 1, 1988) PMMS i
p-intouts.
The licensee estimates the audits will take approximately four weeks per unit.
The audit sequence will be Millstone 1, Mil' stone 2, Con-necticut Yankee, and Millstone 3.
No safety work 1,iadequacies were identified by the inspector's review.
The inspector will centinue to follow licensee r
actions and NUSCO audit results.
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8.0 Temporary Instruction (TI) 2515/93 on High Temperature Inside Cortainment/
hell (71700 D
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I The objective of TI 2515/98 is to review containment temperatures at Millstone
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2 to dete rmine whether or not high containment temperatures are a problem.
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In Soller Water Reactors (EWRs), the containment chillers and ventilation duct
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work are not safety-related.
In Pressurized Water Reactors (PWRs), the fan
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coolers and filters are safety-related; however, the duct work is not safety-i related.
Pn'Rs are required to keep the ambient temperature i* side the con-l l
tainment within a valua specified in the TSs.
But, actual +.em ature pro-l l
files may not have been used to calculate expected equipment operating life-
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l time. Also, temperatures might be obtained ftom non-conservatively located
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'ensors (e.g., in the flow stream of the containment air coolers).
Tempera-
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l tures therefore may not represent the actual average ambient temperature of i
l the containment, especially in the region of the upper plenum.
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The inspector reviewed average containment temperatures recorded by the lic-h i
ensee between Aprii - September 1988.
TS 3.6.1.5, "Primary Containment Ayer -
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age Air Temperature," requires that the value shall not exceed 120 degrees
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F.
Surveillance procedure SP2619A, Step 11 "Primary Containment Average Air l
Temperature Verific:. tion," surveys the associated TS on a daily basis.
The t
following are the average air temperatures between April - September 1988
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utilizing 5 arbitrary data points from $P2619A:
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April 107.54 degrees F i
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103.7 degrees F June
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July 103.6 degrees F
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August 107.9 degrees F
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September -
109.1 degrees F
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These results satisfy in TS 3.6.1,5.
A basis for TS 3.6.1.5 is to ensure containment peak air temperatures do not exceed the design temperature of 288 degrees F during a Loss-of-Coolant Ac-cident (LOCA). The Final Safety Analysis Report (FSAR) 5.2.5.4.3 reports the containment ventilation system will eliminate temperature gradients across the secondary shield walls, the refueling pool walls, and the operating floor.
No inadequacies were noted ir. the TS basis or description in the FSAR.
The inspector reviewed the TS containment air temperature values and assessed whether the t.(erage temperature readings adequately reflect containment con-ditions, and if a significant difference due to temperature sensor location or stratification of containment atmosphere produced "hot spots." TS 3.6.1.5 requires that temperature be measured based on the arithmetic average of two sensors (38 foot elevation Southeast Containment Wall, and the 38 foot South-west Containment Wall). According to SP 2605B, "Containment Leak Rate Test -
Type A," the TS temperature sensors, T-8108 and T-8109, are located at the 44 ft, containment elevation, 145 degrew. and 263 degrees azimuthal, and 60 and 58 feet from the containment centerline, respectively.
SP26058 describes the representative volume fraction of both temperature sensors combined as
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11.6% of the total containment volume. The inspector compared the arithmetic average reading of T-8108 and T-8109 for September 20 and September 22 to all
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eighteen air temperature instruments located in containment with respect to arithmetic volume fraction.
Temperature sensors T-8108 and T-8109 averaged
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to 109 degrees F, and the volumetric-weighted average was calculated to be l
approximately 115 degrees F.
The difference was a result of "hot spots" in temperature sensors T-8097, T-8112, T-8110, and T-8111.
The sensors are located near steam generators and the containment auxiliary circulating fan discharges. The temperatures recorded ranged between 130 degr F to /.41.2 degrees F for the four temperature sensors. The inspector questioned the licensee on the technical justification for selection of T-8108 and T-8109 for TS 3.6.1.5, "Containment Air Temperature" requirements.
This la an ui-
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resolved item (UNR 88-22-01).
The inspector reviewed the licensee's equipment environmental qualification program for selected equipment, as it compares to localized "hot spots' in containment.
In one case, the licensee has shortened the qualified lifetime for Valcor Air-Operated solenoid valves (RC-422, RC-414, RC-415, RC-416, RC-
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417, RC-423, RC-424, and RC-425).
The solenoid valves are used for venting the pressurizer and the reactor vessel.
The inspector reviewed the EEQ Main-tenance Summary Book for the associated solenoid valves.
1.ie licensee assumed a constant temperature of 157 degrees F in the pressurizer block house (loca-tion of Valcor soleroid valves). The 'icensee calculated a reduced lifetime of the solenoids from 40 years (based on 120 degrees F) to twelve years based on 157 degrees F.
The 120 degrees F basis for lifetime was used in Valcor Engineering Qualification Test Report QR52600-515.
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The inspector questioned the licensee staff on the extent of effects on quali-L fication lifetime as it relates to safety related instrumentation.
The lic-
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ensee currently has established a program to identify localized containment
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elevated temperatures and the affects on lifetime and operation of EEQ equip-
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ment. The inspector will follow licensee actions on this program in future r
inspections.
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The inspector reviewed bulk containment temperatures.
The licensee has re-
ported three separate times where bulk containment temperature exceeded the f
l TS 3.6.1.5 requirement of 120 degrees F.
Licensee event reports (LERs) sub-mitted to the NRC were 78-31, 80-03, ano 80-30.
The maximum containment tem-I
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perature recorded from the associated LERs was 120.8 degrees F.
In all cases,
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temperature exceeded the limit for less than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
The high temperature j
in all three cases was a result of inadequate adjust:
...t in the service water
temperature control valve to the Reactor Building Component Cooling Water i
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j (RBCCW) heat exchanger. The control valve was set too high to maintain the
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j Reactor Coolant Pump (R9) seal cooling water at 90 degrees F.
The RCP seal L
i water is cooled by RBCCW.
The licensee corrected this problem by establishing j
a localized Temperature Control Valve for the RCP seal cooling water to main-tain the recommended 90 vegrees F.
The inspector had no further questions.
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9.0 Refueling 0utage Meeting J30703)
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On September 13, the inspa ctor attended a Millstone 2 Refuel Outage Meeting.
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l The meeting agenda focuset on outage schedule reviews department updates and
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j overall comments on the upcoming outage from the unit superintendent.
The i
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outage start date is scheluled for February 4,1989.
The length of the outage i
(from and to full power operation) is scheduled for 46 days 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Major l
work activities discussed at the outage meeting included: steam generator inspection and repair; reactor coolant pump (RCp) seal instrumentation; ser-l i
vice water repairs; RCP motor replacement; snubber overhaul; EEQ and valve
work; and preparations for future steam generater replacement. The unit
superintendent stre, ed communications, performing the activity right the
"first" time, and foi the unit to maintain a questioning attitude on outage
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activities. No inade.
cies w ee noted, i
i 10.0 Periodic Reports (90
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Upon receipt, a periodic report subm:tted pursuant to Technical Specifications l
was reviewed.
This review verified th.' the reported information was valid e
and included the NRC required data, and tnot test results and supporting in-
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formation were consistent with performance specifications. The inspector also
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ascertained whether any reported information should be classified as an ab-
normal occarrence.
The following report was reviewed:
-- Monthly Operating Report for Millstone 2 for September 1988.
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i No inadequacies were identified, j
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11.0 Committee Activities (40700)
The inspector attended meetings 2-88-146, 2-88-147, 2-88-151, and 2-88-152 of the Plant Op e*tions Review Committee (PORC) on September 21, September 22, October 5, and Oc ober 6, respectively.
The inspector noted that commit-tee administrative req f rements were rnet, and concluded that the committees discharged their functuns in accordance with regulatory requirements. The inspector observed thorvugh discussion of matters before the PORC and a good regard for safety in the issues under consideration by the committee.
The issues presented to PORC over the four observed meetings included the follow-ing.
Plant Design Change Request (PDCR) MP2-88-090, "Control Room Ventilation
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Radiation Monitor."
In-service Test T-88-42, "Boric Acid Reduction from Boric Acid Storage
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Tank tr Charging Purrps."
MP2->;U1J. revision to "Monthly Maintenance Walk-through."
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MP2-T417J, revision to "Wide Range Nuclear Instrumentation Calibr
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PDC'c MP2-88-032, "Temporary Change to Lower Fa Upper Reservoir Oil Set-
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nt.nt from 78% - 76.6%."
No inadequactes were noted.
12.0 Maintenance (62703)
The inspector observed and reviewed s.tected portions of preventive and cor-rective maintenance to verify compliance with regulations, use of administra-tive and maintenar.e procedures, compliar.c= *ith codes and standards, proper QA/QC involvement, use of bypass jumptrs and safety tags, personnel protection, and equipnent alignment and retest. The following activity was included:
AWO M2-88-09716 "Repair of Static Inverter No. 6."
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No inadequacies were identified.
13.0 Temporary Instruction (TI) 87-07, Battery Adequacy Audit (62705)
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As documented in Inspection Report 50-336/SS-16, the inspector reviewed Tl 87-07 for the vital station service batteries for Millstone 2.
In order to complete the inspection activities under Tl 87-07, a revics was performed of selected safety-significant batteries for the Millstone $Lation.
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The wet cell batteries review included the Millstone 1 gas turbine battery, diesel fire pu.sp battery, security system battery, and the Emergency Operating
Facility (EOF) battery.
The inspection included the licensee's maintenatice l
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l and surveillance programs in place to assure that each selected battery will
perform its design function.
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r The first battery reviewed was the Millstone 1 Gas Tutbine battery. According i
to Updated Safety Analysis Report (USAR) Section 8.3.1.1.5.2, the gas turbine i
battery supplies direct current (D.C.) power to gas turbine auxiliaries when l
a A.C. power is lo:t.
Table 8.3-4 of the USAR describes a 56 cell, 125v 0.C.,
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340 amp-hour (AH) eight-hour rated battery.
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The taspector reviewed OP-339, Revision 13. "Gas Turbine Generator." The procedure describes checks under form OP 339-2.
The inspector reviewed re-
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l sults of the monthly gas-turbine TS operability surveillance.
No inadequacies (
were noted.
On October 3, the inspector toured the gas turbine battery room at Millstone
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The inspector noted ro n temperature via a Chromalot readout at 76 degrees
F.
The inspector noted minor amounts of sediment on the bottom of cells 21 25, 26, 48 and 49. Good ventilation air flow within the area was noted. The
j sediment formation on the bottom of the cells was discussed with the licensee.
L This will be followed during futcre inspections.
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The diesel Fire Pump battery surveillance indicated that battery surveillances
..e required by Millstone 1 TSs as listed below:
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SP788.3, Revision 3 - (Weekly) TS 4.12. A.H.1.
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SP788.4, Revision 3 - (Quarterly) TS 4.12. A.1. 4.2.
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SP788.5, Revision 1 - (18-Month) TS 4.12.A.1.H.3.
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i The inspector reviewed the three most recent surveillance data sheets for
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i SP788.3 and SP783.4, and the last SP788.5 18-month surveillance. No inade-
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quacies were noted.
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On October 3, the inspector toured the diesel fire pump room.
The area near the fire pump battery indicated sufficient Chemical Burn First Aid equipmont
l and good overall ventilation directed to the top of the battery. The room
temperature was 72 degrees F.
No inadequacies were noted, j
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TM inspector reviewed the licensee's surveillance program for the security
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and Emergency Operating Facility (EOF) diesel batteries.
The review consisted i
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of corrective and preventive maintenance on the batteries from January 1,1984 i
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until September 15, 1983 using the PRMS system.
The maintenance program for
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the security diesel battery is located in procedure 2701J-44, and consists
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j of monthly and annual FMs. Millstone Un n 2 has responsibility for mainten-
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ance of the security battery. The monthly PM consists of pilot cell voltage
i readings, total battery voltage reading, and individual cell specific gravity i
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readings. Annual preventive naintenance (PM) checks individual cell voltage.
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i The licensee's PM program requires replacement of the security diesel bat-teries every 6 years (Procedurc 2701J-54, Revision 2). No inadequacies were noted.
The EOF diesel battery PM program is monthly and annual checks of electrolyte level, specific gravity, individual battery voltage, and total battery voltage.
The EOF diesel battery, per the licensee's PM program, is replaced every 2
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years.
No inadequacies were noted in the PM program between January 1, 1984 and
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September 15, 1988. The inspector concluded that the present licensee sur-veillance and maintenance program for the selveted wet cell batteries is adequate.
j 14.0ManagementMeeting(30703]
Periodic meetings were held with station management to discuss inspection
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findings during the inspection period. A summary of findings was also dis-l cussed at the conclusion of the inspection. No proprietary information was
covered within the scope of the inspection.
No written material was given l
to the licensee during the inspection period.
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