IR 05000236/1988016
| ML20207F771 | |
| Person / Time | |
|---|---|
| Site: | Millstone, 05000236  |
| Issue date: | 08/11/1988 |
| From: | Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20207F764 | List: |
| References | |
| 50-336-88-16, NUDOCS 8808230205 | |
| Download: ML20207F771 (33) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
50-336/88-16-Docket'No.
50-336 License No.
DPR-65 Licenseei'
Northeast Nuclear Energy Company
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P.O. Box 270 Hartford, CT 06101-0270 Facility.Name: Millstone Nuclear Power Station, Waterford, Connecticut Inspection At: Millstone Unit 2 Dates:
June 14 through July 29, 1988 Inspectors:
William J.- Raymond, Senior Resident Inspector Peter J. Habighorst, Resident Inspector Joseph Golla, Division of Reactor Safety David Jaffe, licensing Project Manager, Office of Nuclear Reactor
. Regulation (NRR)
Reporting Inspector:
. Peter J. Habighorst, Resident Inspector Approved by:
N O b k )v 8/t/ /89 E. C. McCabe, Chief, Reactor Projects Section 18 Date Inspection Summaiy: June 14 - July 29,1988 (Report 50-336/88-16)
' Areas _ Inspected: Routine NRC resident and specialist inspection of plant operations, surveillance, maintenance, emergency preparedness, physical security, NRC bulletins and information notices,-previous identified items, storage battery adequacy, Licensee Event Reports (LERs), and committee activities.
Results: No unsafe conditions were identified. Good performance was identified on responses to NRC Information Notices, and good initiative was evident in the'-
undertaking of a safety system functional inspection by the Itcensee.
l 0808230205 OBOB1L PDR ADOCK 05000336 G
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TABLE OF CONTENTS l(
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~1.0 Persons Conts.cted....................................................
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' 2. 0 Summary of Facility Activities.......................................
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J 3.0 Licensee' Actions on Previously Identified Items......................
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' 3.1 (Closed)'UNR 87-25-03: Follow-up Actions Needed to Prevent P.ecurrence of H103 Breaker' and SBL Control Switch Failures....
3,2 (Closed): Follow-up on Items-Identified During Facility Tour....
3.3 TI 2515/73: Inspection of NRC Bulletin 85-03, Motor-0perated
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di Valve Common Mode Failure Oue to Improper Switch Settings.....
3.4 (Closed) IFI 84-07-01: Modify Reactor Crc at System High Point
' Vent S/ stem to /. void :nadvertent Ac'..+ un se to Hot Short
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Circuit...........................
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3.5 (Closed) Nov. 87-29-01: Failure to Control Use of Overtime......
3.6 (Closed) IFI 85-03-04: ' Refueling Cavity Water Sea 1..............
4.0 Plant Tours andl Operational Status Reviews..........................
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4.1 Safety System Operability........................................
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4.2 Review of Plant Incident Reports................................
4.3. Housekeeping Tour...............................................
' 5. 0 Containment Integrated Leak Rate-Test (CILRT)........................
6.0 Physical Security....................................................
7.0- Su ve111ance.........................................................
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'8.0 Maintenance.............................
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9.0 Emergency Preparedness Ort 11.........................................
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10.0 NRC Bulletin 88-05, Nonconforming Materials Supplied by Piping Supplies Inc.' (PSI) and West Jersey Manufacturing Company (WJM)....
l 11.0 Review of Plant Changes and Testing (10 CFR 50.59 Review, Annual
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Operating Report)...........-......................................
12.0 NRC Regional Temporary Instruction (TI) 87-07, "Storage Battery Adequacy Audit".....
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Table of Contents
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PAGE 13.0 NRC Information Notice (IN) 88-38, "Failure of Undervoltage Trip tttachments of General Electric (GE) Circuit Breakers".............
14.0 Sa fety Systera Functional Inspection (SSFI)...........................
15.0 Review of Licensee Event Reports (LERs)..............................
16.0 Review of Periodic Reports.........................................
17.0 Committee Activities.................................................
18.0 Management Meetings..................................................
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DETAILS 1.0 Persons Contacted NNECO S. Scace, Millstone Station Superintendent J. Keenan, Unit 2 Superintendent J. Riley, Unit 2 Maintenance Supervisor F. Dacimo, Unit 2 Engineering Supervisor D. Kross, Unit 2 Instrument and Controls J. Smith, Unit 2 Operations Supervisor The inspector also contacted other members of the Operations, Radiation Pro-tection, Chemistry, Instrument and Control, Maintenance, Reactor Engineering, Station Services Engineering, and Security Departments.
2.0 Summary of Facility Activities The unit began the inspection period in a hot shutdown condition.
This shut-down, which began on June 7, was for replacing the upper gripper coils on the Control Element Drive Mechanisms (CEDMs).
The unit returned to power on June 15, and remained at full power throughout the rest of the inspection period.
No CEDM problems recurred. Also, three separate leaks in the service water system were detected, assessed and repaired successfully.
3.0 Licensee Action on Previcusly Identified Itema (92701)
3.1 (Cl_osed) Unresolved Itta 87-25-03: Follew-up Actions Needed to Prevent Recurrence of H103 Breaker and SBL Control Switch Failures This iten concerned the 6.9 Kilovolt (KV) Bus 25A failure to transfer to the Reserve Station Service Transformer (RSST) due to H103 tie breaker inability to close.
The #111ure was due to a bcnt shaft on the racking mechanism shock absorber (attributed to opeaator error during breiker operation). Additionally, as described in inspection report 50-336/87-25, an interlock switch (52/T.S) failed, allowing the breaker spring charging motor to operate even though the breaker was not racked up fully.
The inspectcr reviewed procedure MP-2701J, Revisien 7, Planned Maintenance (EQ), prescribing routine 4.6 KV and 6.9 KV breaker preventive mainten-ance. The Plant Operations Review Committee (PORC), in meeting 2-88-105 on May 4,1983, authorized a change to MP-2701J to replace the interlock switches (52/IS) for all 4.6 KV and 6.9 KV breakers at a refueling outage frequency. No inadequacies were noted.
The training department implemented a "lessons learned" from this breaker malfunction in lesson plan MP-0P-FUND-2131-1 Appendix A.
Training on this event was provided to licensed and non-licensed operators in cycle 1 of the 1983 requalification training schedule.
The inspector reviewed
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the lesson plan and roster of licensee attendees and concluded the train-ing adequately described and promulgated actions to prevent recurrence.
This item is closed.
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3.2 (Closed) Follow-up on Items Identified During Facility Tour The Regional Administrator toured all three Millstone Units on August i
19, 1987.
During this tour, generic and unit specific concerns were t
identified.
The items and their status are summarized below:
a.
The overall condition of the plant areas and equipment was good, with the exception that in the
'A' ESF (Engineered Safeguards Fea-
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ture) room there was a relatively large number of material discre-pancies. Two concerns from the findings in the 'A'
ESF room _in-clude: (i) a potential operability issue with the 'A' containment
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spray pump due to a missing nut on the pump suction flange and the
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work control closecut process that. allowed the condition to occur; and (ii) the number of leaks constituted a potential significant post-accident leakage source term from the containment.
Internal licensee memo PSE-SA-87-271 dated 9/17/8i ua:ce&nts Stress Analysis Engineering review of the seismic qualification of the
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subject Raised Face Slip-on (RFS0) flange with one missing bolt.
Based on engineering calculation 79-176-911GP dated 8/19/87, the
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RFSO flange joint is adequate to withstand the design loads in the
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as-found condition of only eleven bolts in place instead of the required twelve. The design loads for this evaluation were ex-tracted from Millstone 2, Bechtel Problem Number 46.
Plant Engi-r neering has verified the leak tightness of flange in the as-found condition (Reference Calculation 2-ENG-147).
Correction of the material discrepancies was undertaken on a line-item basis.
Each item was inspected by the !!censee and addressed by work order or on-the-spot correction. A If>t of leaks was com-
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piled and also corrected on a line-item basis. This aspect of the
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I inspection remains open pending verification of correction of the
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identified leaks, of replacement of the missing flange bolt, and of adequacy of ongoing addressal of such items.
b.
The lubrication program for rotating equipment should be reviewed j
to ensure that excessive grease does not enter the motor windings.
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i A plant tour was conducted inspecting motors that have been lubri-cated in the past 4 months. Twenty-five (25) motors were inspected.
The licensee could not determine from visual inspection if any motors were leaking past their seals and into the motor windings.
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A history report was submitted to Production Management Maintenance System (PMMS) tv determine the number of motor failures attributed to grease in the windings; none were noted.
In addition, discus-
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sions with the department electrical maintenance supervisor revealed i
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no motor failures related to over greasing.
NRC resident and re-gion-based inspections have since identified no motor greasing in-adequacies.
This aspect will receive ongoing routine NRC review.
c.
The program for using calibration stickers and trouble reports should be implemented consistently at all three units.
Millstone Station uses a standard set of Trouble Report tags and stickers which are stocked at the warehouse.
These fluorescent tags and stickers include the applicable Trouble Report or Authorized Work Order (AWO) number.
Procedure ACP-QA-2.02C was revised to clarify the use of Trouble Report stickers; uniform use has since been observed.
Representatives from all 3 units' I&C departments met to evaluate the use of calibration stickers.
For installed equipment, a review of applicable standards indicated that stickers should display only the actual calibration date and the identity of the person perform-ing the calibration; calibration due dates need not be documented on the stickers.
Inspector review noted that calibration frequency, grace periods, and scheduling dates for installed equipmer.t is controlled "fa the PMMS, PORC-approved procedures, and the Unit Technical Specifica-tions.
Calibration is also subject to system or unit availability.
The use of the due date block for installed equipment is, therefore, not necessary.
Measuring and Test Equipment (M&TE) does require filling out the Date Due block.
This practice is consistent in the station M&TE Lab.
The sticker used for M&TE equipment will remain unchanged, with the Date Oue filled in as required.
The sticker for installed plant equipment has been redesigned without the Date Due block and will be utilized stationwide on all future equipment calibrations.
Both stickers will comply with the following documents: (1) NV Quality Assurance Program Topical Report (NVQAP); (2) ANSI N45.2.4-1972/IEEE Std 336-1971; (3) Regulatory Guide 1,30; (4) ACP-QA-9.04.
The new stickers are on order.
Since the licensee reassessment, NRC control room inspections have identified only one case of recording a calibration due date on an installed equipment calibration sticker.
Calibration stickers will be routinely examined during NRC inspections.
d.
The plant ventilation system appears to be a vehicle for tae spread of dirt throughout plant equipment areas.
This creates the poten-tial for an adverse impact on operation of electrical equipment and on long-term reliability of componercs.
It is reccmmended that actions be taken to minimize the spread of dirt.
Based on several
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observations during-the tour. actions should 'also be taken to clean
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the ventilation intake screens on electric motors (this finding. is related_to item b above).
Engineering has written a PDCR evaluation'to install intake filters in the Turbine Building, PDCR MP2-87-051. Motor intake screens are cleaned during. annual Preventative Maintenance (PM). This aspect.
P remains open_pending confirmation of continued cleanliness..
e.
The control 'of radiological areas was generally good and there were relatively few areas with significant contamination levels.
However, a work effort that was contrary to good contamination control in-volved a fire hose that was run from the Unit 2 refueling floor to the aerated waste tanks on the -45 ft. elevation. Although the hose had a sleeve to protect.it;from contamination on the refueling level, there apparently was no sleeving in place around coupling in C e hose segments.to limit the spread of contamination.
The responsible Engineer was cautioned about the spread of contami-nated fluid due to the lack of sleeving around the fire hose coup-lings.
The hose was immediately removed as it was no longer re-quired for the job. ~ This aspect of contamination control will be
. routinely evaluated during'NRC' inspections, f.
A white chalky substance was observed on top of the electrolyte in the C&D main station batteries, 201 A&B. What is its source and does it have an impact on battery operation and life?
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A letter dated 9/23/87 from C&D power Systems to the licensee states the following.
The white chalky substance is minute fragments of
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glass fiber that came loose during the handling of the mat and separator assembly and an excess amount of binder used to adhere the glass mat to the separator. All materials used in the manufac-ture of the cells are compatible with other components in.the cell itself.
No detrimental effect regarding cell life or performance is anticipated. There is no adverse performance condition which may develop.
The inspector reviewed the licensee's actions and noted during routine inspections that housekeeping was good (see Detail 4.3), with one excep-tion being the miscellancous equipment around the fans in the ESF room.
This problem was identified to the licensee and improved performance has
since been observed. Also, the 11 specific material discrepancies iden-tified during the Regional Administrator's tour have been corrected.
No inadequacies were noted.
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3.3 Temporary Instruction (TI) 2515/73:._ Inspection of NRC Bulletin 85-03, Motor-0perated Valve Common Mode Failure Due to Improper Switch Settings The licensee has identified the selected safety-related valves, the valve maximum differential oressures, and the program to assure valve operabil-ity by letter dated June 11, 1986.
Review of this response indicated the need for additional inr'ormation which was requested by an NRC Re9 son I letter dated July 29, 1987.
Review of the licensee's September 4, 1987 response found that the selection of the applicable safety-related valves was addressed, the valve maximum differential pressures are in accordance with the bulle-tin, and the program to assure valve operability requested by action item e. of the TI is acceptable.
No inadequacies were noted.
By letter dated June 27, the inspector was notified by the NRC's Office of Nuclear Reactor Regulation (NRR) of close-out of TI 2515/73 item "e" to NRC Bul-letin 85-03.
The results of the inspections to verify proper implementation of this program and the review of the final response required by action item f.
of the TI will be addressed in a future inspection.
3.4 (Closed) IFI 84-07-01: Modify Reactor Coolant System High Point Vent System t; Avoid Inadvertent Actuation Oue to Hot Short Circuit This item concerned the electrical design of reactor coolant system (RCS)
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high point vents oer TMI-Task Action Plan (TAP) Item II.B.1.
In 1984, the licensee's electrical design for the RCS high point vents did not implement grounded shield leads in control wiring or additional switch conttets to break both connections in each vent valve isolation solenoid.
Therefore, TMI-TAP Item II.B.1 clarification A(8) was not met in that the RCS vent system was not protected against inadvertent actuation due to hot shorts.
This item was initially reviewed in routine inspection report 50-336/84-07.
The susceptibility of the RCS vent valves to inadvertent opening due to a hot short was subsequently reviewed by the licensee.
The review con-cluded: i) two valves in series must open to create a vent path from the RCS; and ii) two separate hot shorts are needed to cause a spurious opening of one vent valve.
Inspector review of the licensee's 10 CFR 50 Appendix R Compliance Review, Revision 0, and Piping and Instrument Drawings (P&ID) 25203-26014; 25203-39405 Sh. 65; and 25203-32007 Sh. 41 concluded that no single failure would open a path for radioactive fluid from the RCS to the atmosphere via the pressurizer or reactor head vents due to hot short circuits.
In NRC Generic Letter 85-01, the "Fire Pro-tection Policy Steering Committee" defined criteria for circuit protec-tion as "safe shutdown capability should not be affected by any one spurious actuation or a signal resulting from a fire in any plant area."
To this end, the inspector reviewed the licensee response to the NRC
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Safety Evaluation Report (SER) of September 1983, and to Generic Letter 86-10. "Implementation of Fire Protection Requirements-in April, 1986.
No inadequacies were noted.
3.5 { Closed)NOV 07-29-01: Failure to Control Use of Overtime The licensee was requested to review the 1937 outage to determine the extent to which NE0 1.09, "Overtime Controls for Personnel Working at the Operating Stations," was not implemented.
Licensee investigation concluded seven individuals exceeded the NEO 1.09 guidance.
In routine o
resident inspection report 50-336/87-29, inspector review found the lic-ensee's corrective actions to the Notice of Violation acceptable, with further NRC review of the control of overtime planned to verify ongoing compliance with the licensee procedure.
The NRC received an allegation concerning a J;1y 17, 1987 incident, in which licensee personnel worked 20.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> without prior supervisor authorization.
The workers performed electrical inspections on Environ-mentally Qualified (EQ) safety-related valves.
The inspector reviewed this allegation in routine resident inspection report 50-336/88-06.
In this review, the inspector interviewed the alleger, the alleger's foreman, another involved worker, and the Millstone 2 superintendent.
The four people interviewed stated their belief that the safety-related valve in-spections were completed correctly.
The NRC requested the licensee to
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investigete the event and report the findings to the NRC. The inspector reviewed the licensee's response and concluded the safety-significance in this case was minimal; however, supervisory oversight of employee work activities in this particular case was marginally acceptable. Also, as documented in inspection report 50-336/88-06, the inspector interviewed first line supervisors to determine the amount of interface occurring between workers, the criteria utilized for granting overtime, and cap-ability of workers to work excess hours especially on safety-related work.
No inadequacies were noted at that time.
The inspector concluded the licensee actions fulfilled the requirements of NE01.09 and ACP 1.19, "Overtime Controls for Personnel Working at the Operating Station," which implements TS 6.2.2.g.
This item is closed.
3.6 (Closed) Inspector Follow Item (IFI) 85-03-04, Refueling Cavity Water Seal NRC Bulletin 84-03, issued on August 24, 1984, described the failure of the refueling cavity water seal at Haddam Neck Nuclear Power Station.
The Bulletin requested applicants and licensees to investigate the pos-sibility of a similar occurrence at their facility.
Northeast Utilities responded by letter dated November 29, 1984.
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a.
Seismic Support for Refueling Cavity Drain Line The' licensee identified failure of the refueling cavity drain line
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as a potential source of loss of refueling cavity water. To de-crease the likelihood of failure, the licensee committed to upgrad-ing the supports for this line by adding restraints and achieving leismic Category 1 status up to the first isolation valve. That upgrade was performed under PDCR 02-25-85.
During this inspection, the drain line was inaccessible for direct inspection due to its location inside containment.
Based upon review of PDCR 02-25-85 and calculation 84-124-348GP, the upgrade of the refueling cavity drain line was accomplished using conservative design criteria.
The licensee's internal correspondence shows that the upgrade was completed on March 7, 1985.
b.
Update the procedures to include Response to a loss of Refueling Cavity Water Inventory Abnormal Operating Procedure (A0P) 2578, "Loss of Refuel Pool and Spent Fuel Pool Level," was reviewed. The procedure addresses the
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following areas: evacuation / assignment of personnel; communications; health physics; monitoring of spent fuel pool make-up and cooling
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systems; and disposition of spent fuel in transit when leakage oc-
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The procedure provides appropriate operator response to failure of the refueling cavity or nozzle dam rubber seals. The content of AOP 2578 is consistent with the event analysis presented in the licensee's November 19, 1984 response to IEB 84-03, c.
A Demonstration will Verify that the Operator Actions Required by the procedure to Place Reactor Fuel in a Safe Location can be Performed Within the Time Interval Assumed in the Analysis The licensee's November 19, 1984 response to IEB 84-03 states that,
"Even if it were conservatively assumed that it would take the operator ten (10) minutes to diagnose a refueling cavity pool seal failure, plant personnel would still have approximately eighteen (18) minutes to secure any irradiated fuel in transit.
This time interval will be verified as sufficient to allow plant personnel to perform all necessary actions identified in the operating proce-dure regarding a cavity seal failure." Procedure AOP 2578 contains the following provisions for securing irraciated fuel:
If the fuel assembly is near the core and time allows, then
.y position it in a core location.
If near the upender and the transfer carriage is in containment and time allows, then:
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position the fuel assembly in the upender.
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put it in the horizontal position.
iii transfer it to the spent fuel pool.
If unable to perform the above steps, then lower the hoist over a clear area in the south saddle until load on the dillon (load cell) begins to decrease.
Do not attempt to ungrapple the fuel assembly.
Note: Approximately 50 minutes are needed to com-pletely close the transfer tube gate.
In-Service Test T85-11. "Movement and Timing of a Fuel Assembly to a Safe Storage Location" was conducted on March 11, 1985.
The test was prepared to confirm the time required to secure irradiated fuel in transit.
The data from T85-11 indicated that:
The time required to move a fuel assembly from above the core
to a safe location at the South Saddle was 12 minutes and 34 seconds.
The time required to move a fuel assembly from above the core
to the spent fuel pool upender was 18 minutes and 15 seconds.
The time required to stroke the Transfer Tube Isolation Valve
(2-RW-280) is 27 minutes, 11 seconds to open and 11 minutes, 44 seconds to close.
Based upon the above, the licensee has confirmed that the time re-quired to secure an irradiated fuel assembly in transit during a refueling cavity draindown is within the available time identified in their November 29, 1984 submittal.
d.
The Licensee's Submittal does not Address the Effects of a Failure of the Steam Generator Nozzle Dam System Procedure A0P 2578. Rev. O, February 25, 1985 specifically addresses failure of the rubber seals on both nozzle dams of one hot leg nozzle.
The maximum expected leakage was calculated to be 3210 gpm compared to the maximum leak rate for a refueling cavity seal, cal-culated to be 6493 gpm. The actions required per AOP 2578 would not differ depending on the source of the leakage since the refuel-ing cavity would drain to the reactor vessel flange in either the refueling cavity seal or nozzle dam seal failure mode. Therefore, time would be available to secure irradiated fuel in transit and to plan recovery measures in the event of a nozzle dam seal failure.
Based on the preceding, the licensee has satisfactorily addressed the remaining concerns associated with IEB 84-03.
This item is close.
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4.0 Plant Tours and 0.oerational Status Reviews (71707)
The insper*or observed plant operations during regular and backshift tours of the foiscwing areas:
.ontrol Room Auxiliary Building Vital Switchgear Room Enclosure Building Turbine Building Intake Structure Diesel Generator Room Fence Line (Protected Area)
Containment Building Control Room instruments were observed for correlation between channels, pro-per functioning, and conformance with Technical Specifications. Alarm condi-tions in effect and alarms received in the control room were reviewed and discussed with operators.
Posting and control of radiation, contamination, and control of high radiation areas were inspected. During plant tours, logs and records were reviewed to ensure compliance with station procedures to determine if entries were correctly made, and to verify correct communication and equipment status.
Records reviewed included various operating logs, turnover sheets, and tagout logs.
.4.1 Safety System Operability (71710)
Emergency systems were reviewed to verify they were operable in the s.andby mode.
The systems reviewed were the High Pressure Safety Injec-tion, Low Pressure Safety Injection, and Auxiliary Feedwater System.
The status of the Hot Shutdown Panel and emergency diesel generators were also inspected.
Tha review considered proper positioning of major flow path valves, proper operation of indications and controls, and visual inspections for proper lubrication, cooling, and other conditions.
Re-ferences used included the Final Safety Analysis Report (FSAR); plant instrument and piping diagrams (P& ids) 25203-26015, "Safety Injection and Containment Spray Systems; Operating Procedures OP-2604E, Facility I HPSI valve lineup, OP-2322, Rev. 9, Auxiliary Feedwater, and OP-2346A, Rev. 9, Emergency Diesel Generators.
No inadequacies were noted.
4.2 Review of Plant Incident Reports (PIRs) (71707)
The plant incident reports listed below were reviewed during the inspec-tion period to (1) assess the significance of the events; (ii) review the licensee's evaltations; (iii) verify whether the licensee's response and corrective actions were proper; and, (iv) verify that the licensee reported the events in accordance with applicable requirements.
The PIRs reviewed were: 88-47, 88-48, 88-49, 88-50, 88-51, 88-52, 83-53, 88-54, 88-55, and 88-56.
The following items warranted inspector followup:
PIR 88-49, "Dropped Rod #23 in Group 2." See inspection report
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50-336/88-1.
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PIR 88-50, "Reactor Shutdown due to an inoperable CE0M #23".
See
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inspection report 50-336/8S-13.
PIR 88-52, "Wide Range Nuclear Instrument Channels B and 0".
See
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inspection report 50-336/88-13.
PIR 88-56, "Service Water Flanges with low Hardness Values".
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Detail 10.0 of this report.
No inadequacies were noted.
4.3 Housekeeping Tour (71707/54834B}
The resident inspector accompanied the unit superintendent on a tour of the turbine building on July 20. A purpose of the accompaniment was to assess licensee management awareness to routine housekeeping and ongoing activities. The superintendent discussed with the inspector the fre-quency of detailed facility tours. The superintendent visits all loca-tions within the facility monthly. Consideration is being given to making such inspection visits weekly.
The tour identified 18 minor deficiencies in housekeeping, equipment operation, and temporary system installation.
The superintendent re-ferred the deficiencies to the appropriate departments for resolution.
During the tour, the inspector witnessed the following activities: In-Service Inspection (ISI) of the auxiliary feedwater discharge line, painting in the upper elevation of the turbine building, and condensate pump air removal maintenance.
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Overall, the inspector assessed turbine building cleanliness as good.
Licensee management expressed a strong desire to improve cleanliness as it relates to safe operation. No inadequacies were noted.
5.0 Containment Integrated Leak Rate Test (CILRT) Results Review (70323)
The NRC reviewed the licensee's February 1988 CILRT results documented in accordance with 10 CFR 50, Appendix J, paragraph V.B.3.
These results were summarized in a technical document entitled "Reactor Containment Building Integrated Leak Rate Test" and were attached to the licensee's letter dated May 2, 1988 to the NRC.
The report contains a test summary, presentation of results, and other information such as a description of plant and computer software and data analysis techniques.
The total time calculational method of Bechtel Nuclear Topical Report BN-TOP-1 for reduced duration tests was utilized. This method is acceptable per 10 CFR 50, Appendix J, which stipulates that Type A tests be conducted in ac-cordance with American National Standard N45.4-1972, "Leakage Kate Testing of Containment Structures for Nuclear Reactors."
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The purpose of the test was to demonstrate that leakage through the contain-ment building and systems penetrating containment do not exceed that allowed by technical specifications. The test was conducted with containment isola-tion valves and containment pressure boundaries in an as-left condition.
The containment met the leakage criterion in both the as-found and as-laft condi-tion.
The test was witnessed by an NRC region based inspector and was fol-lowed by a successful leakage verification test.
Inspection findings are documented in NRC Region I Inspection Report No. 50-336/83-04, s.
Type A Test Parameters 1.
Test Method Absolute 2.
Calculation Method Total Time (reduced duration per BN-TOP-1)
3.
Test Pressure 54 psig (accident pressure)
4.
Test Duration:
5ttbilization Period 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />
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Data Gathering for Leakage Calculation 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
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Verification Leak Rate Test 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />
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b.
Test Results wt. %/ day 1.
Maximum Allowable Leak Rate 0.375 2.
Measured Leak Rate (Lam)
0.100 3.
95% UCL (upper confidence level) plus cor-0.138 rections for isolated penetrations, valves not in post accident position, etc.
This is the containment as-left leak rate.
4.
95% UCL plus corrections plus penalty leak- 0.201(as-found)
age.
(Penalty leakage is add-on Type C leakage due to improvements.) This ad-justs the result to the as-found value.
5.
Conclusion Acceptable in both the as-left and as-found conditions
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6.0 Physical Security (81064)
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Selected aspects of site security reviewed during inspection tours included access controls, personnel and vehicle searches, personnel monitoring, place-ment of physical barriers, compensatory measures, guard force staffing, and follow-up on response to alarms and degraded conditions.
No inadequacies were noted.
7.0 Surveillance (61726)
The inspector observed the folicwing surveillance tests for conduct in ac-cordance with approved procedures, for results compliance with Technical Specification (TS) and administrative requirements, and for deficiencies.
No unacceptable conditions were identified.
SP-2401G, Rev. 8, RPS Bistable Trip Test, Channel A - June 30, 1988.
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M2-88-06350 and M2-88-05351, Internal Camera Inspection of "B" EDG Air
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Receivers, July 18, 1988.
SP-2405A, Rev. 3, Seismic Events System Channel Checks; SP-24059, Rev.
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2, Seismic Channel Functional Test; SP-205C, Rev. 3, Seismic Channel Calibration.
The inspector reviewed the surveillance program associated with the intake structure seismic monitor (NE 9450).
This seismic instrument is utilized to predict plant responses and, in particular, the unit's intake structure to seismic motion. The monitor is a triaxial time history accelerograph.
It features central recording on magnetic tape cassettes with a remote transducer element and a separate triggering unit. Both the transducer and the trigger-ing unit are connected to a recording and playback system in the control room.
Upon a seismic event, the system is activated within 0.1 second.
The sensi-tivity of the accelerograph is 0.00199 to 1.09g.
The inspector reviewed the three most recent surveillances.
All data met the acceptance criteria.
No inadequacies were noted.
8.0 Maintenance (62703)
On July 19, the inspector observed routine preventive maintenance on the "B" Emergency Diesel Generator (EDG).
Prior to observation, the inspector veri-fled the licensee's operability run of the "A" EDG and entrance into Technical Specification (TS) action statement 3.8.1.1.b, Electrical Power, A/C Sources.
The action statement requires the licensee to verify operability of the re-maining EDG within one hour and at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter, and to return the inoperable EDG to operable status within seventy-two (72) hours.
No in-adequacies were note m
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The inspector verified required administrative approvals for authorization to perform "B" EDG preventive maintenance.
The authorized work orders were M2-8S-06350 and M2-88-06351.
The inspector independently reviewed portions of the tag-out.
No inadequacies were noted.
The inspector observed portions of the bi-weekly, monthly, and quarterly PMs for the "B" EDG, per Maintenance Procedure 2701J-19. Maintenance personnel appeared well versed and knowledgeable, The inspector had no further ques-tions.
9.0 Emergency Preparedness Drill (82701)
On June 23, the licensee conducted an unannounced Emergency Preparedness Call-Back Drill which exercised the response capabilities of the station and cor-porate emergency response organizations.
The drill objective was to verify the emergency response organization could be.atified by radiopager message and respond to the site within the time sonstraints in NUREG 0654, Table B-1,
"Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants " Table B-1 iden-tifies members of the licensee's organization who are to respond within 30 and 60 minutes of receiving the radiopager message.
The drill required all Millstone Station and Corporate personnel on call to telephone the call-back recorder system.
Responders were to provide an esti-mate of how long it would take to travel to the site.
Personnel were not re-quired to actually respond.
State and local officials did not participate.
However, the licensee informed the State of Connecticut Office of Civil Pre-paredness that the drill was to be conducted. No inadequacies were noted in the drill methodology.
The drill was conducted at 8:00 pm on June 23.
Six of the 58 station and corporate employees involved did not respond to the radiopager message.
Three af these six individuals were contacted by telephone, and back-up personnel for the remaining three individuals were also contacted by telephone.
The licensea determined that the acceptance criteria of augmenting shift coverage per the NUREG Table B-1 requirements within 30 minutes was met.
The licensee also reported that one of the three call-back recorders was tied-up for ap-proximately 10-15 minutes for no apparent reason.
Licensee corrective actions included: an investigation into why six indivi-duals did not respond to the radiopager message; plans to increase the number of call-back recorders at the station; and plans to re-emphasize to the emer-gency plan organization the importance of responding to radiopager messages.
The licensee plans to conduct another unannounced call-back response drill within the next few month _ _ _ _ _ _ _.
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i Inspector review of the guidance of NUREG 0654, Table B-1 concluded that the licensee marginally satisfied the guidance.
The licensee's proposed correc-tive actions are considered appropriate.
The inspector will follow-up the licensee corrective actions and future call-back response drills in a subse-
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quent routine inspection.
10.0 NRC Bulletin 88-05, "Nonconforming Materials Supplied by Piping Supplies, Inc.
(PSI) and West Jersey Manuf acturing Company (WJM) (92717)
Copies of certified material test reports (CMTRs) for material supplied by
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PSI and WJM apparently contained false information on material supplied to the nuclear industry. On May 6, 1988, the NRC issued Bulletin 88-05, Noncon-forming Materials Supplied by PSI and WJM.
The NRC requested licensees to review purchasing records to determine of ary American Society of Mechanical Engineer (ASME) or American Society of Test Materials (ASTM) materials have been supplied by PSI since January 1,1985, or from WJM since January 1,1976.
Licensees were also requested to review the location of installed components in safety-related systems, the suitability for intended service, and conform-ance with code and procurement specifications.
Supplement 1 of NRC Bulletin 88-05 was issued in response to test data for two flanges supplied by WJM to the Shearon-Harris nuclear power plant.
The results demonstrated material properties significantly below the ASME and ASTM specifications.
Supplement 1 to Bulletin 88-05 requested licensees to test installed flanges and fittings, within 30 days of receipt of the supplement, for conformance to ASME and ASTM specifications and, if deviations were pre-sent, to justify continued operation of the facility.
.
The inspector reviewed the licensee's activities in response to Bulletin 88-05, and associated Supplement 1.
The licensee's Quality Services, Purchasing, and Engineering departments reviewed purchase orders and work orders to de-termine the number of flanges and fittings supplied by WJM or PSI.
That re-view identified a number of spare flanges and those installed in safety-related systems.
Licensee review identified the following vendors who sup-plied WJM and PSI fittings to the Millstone Station: Radnor, Tyler-Dawson, Guyon, Cunningham Supply, and Pullman Power.
The licensee's corporate Gene-ration Engineering department coordinated activities at the facility and de-veloped conformance testing on installed flanges on safety-related systems.
The results of testing of installed flanges were compared with the CMTRs sup-plied by the vendor and with the applicable ASME and ASTM code standards.
The licensee tested the potential nonconforming materials using an Echo-tip instrument. The instrument tests for Brinnell hardness values.
Five readings were taken on the flange surface.
The highest and lowest values were deleted; the three remaining readings were averaged to determine a hardness value.
The licensee reported to the inspector that the hardness technique was ac-ceptable to the Electric Power Research Institute (EPRI),
The inspector dis-cussed the EPRI position with the NRC (NRR) staff, and concluded that it was acceptable, t
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The 35 PSI or WJM flargos and location identified at Millstone were:
Number of ASME/ ASTM Flanges and Size Manufacturer location Specification
(24")
WJM
"B" Service Water SA 105 Discharge Header
(24")
WJM Service Water return SA 105 from RBCCW Heat Exchanger
(24")
WJM Service Water header to SA 105 RSCCW Heat Exchangcr
(24")
WJM Service Water System SA 105
(3")
WJM Service Water System SA 105
(10")
WJM Service Water supply to SA 105 Emergency Diesel Generators (EDG)
(8")
PSI Service Water supply SA 105 to EDGs For the listed flanges, licensee review elso concluded that ten 24" and eight 3" flanges were replaced before issuance of NRC Bulletin 98-05.
Independent inspector review of documentation confirmed that conclusion.
The licensee conducteo in plant testing for the remaining 17 installed flanges.
Of those, three service water flanges to the EDG produced Brinell Hardness values below the ASTM specification.
The ASTM specification range for SA-105 material is 137-187.
The hardness values for these three flanges ranged be-tween 131 and 135.
Licensee investigation revea19d these flanges were in-stalled in October 1936 and manufactured by PSI. The licensee also reported to the inspector that two other PSI flanges were inaccessible for testing.
NRC Bulletin 88-05 asks the licensee to notify the NRC of any deviation from ASME/ ASTM in safety-related flanges and of any inaccessible flanges or fit-tings identified in safety-related equipment or systems. On July 14 at 3:25 p.m., the licensee notified the NRC Operations Center of their findings.
The inspector reviewed the licensee's justification for acceptability of the flanges installed in the service water system.
The licensee concluded the following. Allowable stress in the system stress analysis is based on a 60,000 psi tensile strength for the piping, not flanges. The flange tensile strength required for the worst case must be equal to or greater than 32,000 psi.
This is based on actual stress. When compared to ASME Section III Appendix 3, this indicates a margin between required minimum tensile and measured tensile
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strength of 31,000 psi. The reduced strength of the flanges tested does not affect the stress analysis for any conditions previously analyzed such as seismic, deadweight and thermal concerns.
The inaccessible PSI flanges were located in the same general plant area.
The inspector verified the flanges were inaccessible for testing.
These flanges were from the same heat number as the tested flanges which deviated from the ASME/ ASTM standards.
The licensee concluded the inaccessible flanges were also acceptable: the lowest tensile strength found for the accessible flanges of the same heat of metal was 63,900 psi (for a Brinell hardness of 131), and the tensile strength required under worst case conditions is 32,000 psi.
No inadequacies were noted in the licensee's justification for continued operation (JCO).
The inspector observed hardness testing of four (24") service water dischargt-flanges in the intake structure.
The licensee's inspection team consisted of a maintenance worker, a fire watch, a station engineer, and a corporate engineer. The maintenance worker prepared the flange surface, the station engineer recorded the hardness data, the corporate engineer tested the flange, and the fire watch observed grinding activities. No inadequacies were noted.
The licensee has identified twenty-one flanges in their warehouse manufactured by either WJM or PSI.
These flanges will be sent to Dirats Laboratory for chemistry and tensile property verification, and to the licensee's laboratory for metalographic examination.
As of July 20, the licensee had completed hardness testing of all identified flanges manufactured either by WJM or PSI. On August 3, 1938, after the in-spection period for this NRC report, the NRC issued Supplement 2 to NRC Bul-letin 88-05.
For operating facilities, that supplement temporarily suspended the field measurements, testing, record review, and JC0 preparation requested by the bulletin and Supplement 1, pending NRC review of data and analysis and NRC determination of appropriate additional action.
This bulletin remains open for Millstone 2 pending the results of that NRC determination.
11,0 Review of Plant Changes, and Test (10 CFR 50.59 Review, Annual Operating Report) (37702/37703)
By letter dated February 23, 1933, the licensee reported the January 1, 1987 to December 31, 1987 plant changes, test and experiments to Millstone 2.
A sample of safety-significant activities, detailed below, were reviewed to de-termine the adequacy of the safety evaluations prepared per 10 CFR 50.59(b)(1).
The safety evaluations for the activities identified below were reviewed.
PDCR 02-94-86, Check Valve 2-SW-13A and 2-SV-13B Replacemant.
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PDCR 02-003-87, Installation of 2-51-657 Manual / Air Actuator.
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PDCR 02-106-86, Provide vital power for AFW room sump pumps.
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PDCR 02-100-86, Replacement of Diesel Generator differential relays.
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PDCR 02-005-87, Steam Generator tube stabilizers.
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PDCR 02-71-85, Vital AC switchgear chilled water system.
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E0P-2525, Standard Post-trip Actions, Rev. 2.
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Licensee procedure ACP-QA-3.08 (NE0 3.12), Revision 3, is used for preparing Safety Evaluations. The procedure covers the following.
Responsible initiating and reviewing organizations / individuals.
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Guidance on safety evaluation content including determining the existence
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of unreviewed safety questions.
Procedures and responsibility for conducting supplemental safety evalu-
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ations (integrated safety evaluation).
Numerical guidelines associated with key safety analysis parameters for
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determination of unreviewed safety questions.
A detailed discussion of each criterion associated with the definition
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of "unreviewed safety question" in 10 CFR 50.59(a)(2).
Procedure hEO 3.12 contains a Safety Evaluation Worksheet for documentation of the safety evaluation.
Step 6.1.3 of the procedure states in part:
"As a general practice, the Safety Evaluation Worksheet provided in At-tachment 8.8 shall be used to document the safety evaluation.
If the Safety Evaluation Worksheet is not used, the preparer shall ensure that the content of the safety evaluation satisfies the requirements of this procedure.
Approval by the Manager /NUPOC Department Head shall be re-quired for safety evaluations not documented by the Safety Evaluation Worksheet."
In each case examined, the unreviewed safety question criteria of 10 CFR 50.59(a)(2) were properly addressed.
The most comprehensive safety evaluation was associated with E-2525 where the impact of reactor coolant pump tripping was assessed for specific accidents addressed in the FSAR.
For PDCRs where components were replaced on a one-for-one basis, (e.g., PDCRs 2-100-86 and 2-71-85) the safety evaluations were evaluated as minimal but adequate. No unreviewed safety questions were identified by the licensee or by subsequent inspector review.
The Safety Evaluation Worksheet was available in only two cases PDCR 02-003-87 and E0P-2525.
Not using the worksheet may have violated Step 6.1.3 of NEO 3.12.
This item is unresolved pending determination of whether worksheets were and are being appropriately used to develop safety evaluations (UNR 88-16-01).
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Because acceptable safety evaluations were found in all cases examined, the licensee was found to be in compliance with 10 CFR 50.59.
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12.0 NRC Region I Temporary Instruction (TI) 87-07, "Storage Battery Adequacy
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Xudit" (62705)
The primary objective of Regional TI 87-07 was to determine if the licensee's program is adequate to assure that storage batteries will, in accordance with the current licensing basis, remain operable.
The inspector evaluated the licensee's program under the guidance provided in Regional il 87-07.
The results of the inspection are documented below:
12.1 General Battery Information The vital 125-volt DC supply system consists of two station batteries
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(201A and 2018), two isolated switchgear busses, and associated distri-bution panelboards.
Each of the two DC switchgear busses is supplied by one station battery and two half-size battery chargers, with an iden-tical pair of standby chargers for use with either of the two 60-cell station batteries.
Each switchgear bus supplies two independent and physically separated 125 volt de panelboards for vital loads, two 125 volt de panelboards for nonvital loads, and two separated 125 volt DC/120 volt AC static inverters for vital instrumentation.
The following is general information concerning Millstone Unit 2 vital station batteries 201A and 2018.
Manufacturer:
C&D Type ICV-33 Rating:
8 hr. (amp-br. ) - 2320.
I minute (amp) - 2560.
Voltage: Nominal:
125 VDC.
Minimum:
105 VDC.
Maximum:
140 VDC.
Qualified seismic life: 40 years (see Detail 12.3).
Design electrical life:
12-15 years.
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A e: 8.5 years
Placed in service: November 1979.
Plan for replacemar.t:
The battery capacity tcat must demonstrate the battery has 80% of the manufacturer's rating.
If a battery cannot meet the 80% rating, che bstte' y is at end of life.
Maintenance procedure
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SP 2736F pla*
the batt ery capacity and b used to predict when the bat-i
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tery will reach 80% capacity (decrease its terminal potential to 105 volts in less than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />).
The present capacity of battery 201A is 91.3%, 2018 is at 89.95.
12.2 previous Licensee Actions The inspector reviewed licensee actions on the following NRC Information Notices (ins):
IN 83-11. "Possible Seismic Vulnerability of Old Lead Acid Storage
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Batteries."
IN 84-83, "Various Battery Problems."
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IN 85-74, "Station Battery Problems."
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IN 86-37, "Degradation of Station Batteries."
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In regard to IN 83-11, the licensee had all Unit 2 operator license holders read and acknowledge the information provided concerning the potential seismic vulnerability of lead-acid storage batteries. No in-adequacies were noted.
The seismic lifetime and qualifications of the station batteries are discussed in report detail 12.3.
IN 84-83 discussed various emergency DC power supply battery problems at nuclear power plants. One item was adequacy of storage battery capacity.
The licensee determined in early 1935 that the station bat-teries' capacity was acceptable.
The inspector reviewed original and recent engineering calculations on emergency and normal capacity loads on the vital 125 VDC system using PAS 3-156-802-GE DC load current profile.
This showed that worst case battery loading is less than 80% of battery capacity.
Another specific NRC concern in IEN 84-83 was the use of non-compatible lubricants to clean battery posts of anti-corrosion grease during rework of the intercell connections.
The licensee uses specific cell mainten-
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ance kits purchased from the battery manufacturer.
The kits contain terminal lubricants which are qualified by the vendor (C&D Batteries)
for use on batteries, The inspector reviewed the vendor technical manual and compared the maintenance kits used in the plant.
No inadequacies were noted.
IE Information Notice 86-37 addresses significant problems occurring in station batteries. Recently, two nuclear power plants declared their station batteries inoperable.
One was due to substantial degradation of the cell plates and the hook area where the plates attach to the cell posts, and the other was due to specific gravities below the Technical l
Specification (TS) limi.
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The licensee examined the station batteries and found no evidence of flaking / degradation of the cell plates or the hook area.
Inspections of these areas have been added to Surveillance procedure 2736B, "Complete Battery Cell Measurement." The inspections are performed quarterly.
Spec.ific gravities below Technical Specifications due to inadequate float voltage were not evaluated by the licensee as a problem with the station batteries.
The Maintenance Department routinely reviews specific gravity surveillances and compares them with previous historical data.
The re-views have indicated no specific gravities below the minimum specifica-tion.
No inadequacies were noted.
12.3 Seismic Lifetime The inspector discussed the seismic lifetime for the station battery racks, cells, chargers, and associated disconnect switches with the licensee.
The licensee's evaluation concluded the seismic lifetime is forty years.
The inspector reviewed Plant Design Change Request (PDCR)
2-153-79 for replacement of both station batteries.
The PDCR provided vendor seismic qualifications conform 6nce criteria to the licensee.
The vendor manual states that, if the replaced station batteries are in-stalled per the vendor's approved procedures, the original seismic quali-fications are maintained.
The inspector reviewed licensee work orders 2-535-79, "201B Battery Cell Replacement," and 2-509-79, "201A Station Battery Changeout." The Quality Assurance (QA) inspection plan in each of the work orders documents the vendor torque specifications for battery racks and floor anchor bolts.
No inadequacies were roted in the records of replacement of station batteries in 1979.
The Final Safety Analysis Report (FSAR), section 8.5.2.2 describes the simulated seismic environment of the station battery chargers, station batteries, disconnect switches, and individual battery cells. The in-spector reviewed the description in the FSAR and discussed the design specifications, results, and seismic justification with the licensee.
The inspector had no further questions in regards to this matter.
12.4 Electrical Sizing and Qualification The inspector reviewed the initial station battery sizing per Bechtel calculation E-205-3 in August, 1971. The calculation concluded the eight-hour loading, using C&D Type LCU battery ceils, was 1888 Ampere-hours (AH).
This provides an about 18*. margin for the 8-hour battery rating of 2300 AH supplied by the vendor.
The inspector verified calcu-lations to independently derive the same conclusion.
In 1979, the licensee replaced the station batteries (201A and 2018).
The licensee's battery surveillance inspections and subsequent i.1 spec-tions by the manuf acturer (C&D Batteries) had disclosed chemical erosion of the positive battery posts in a number of cells.
The inspector ques-tiened the licensee's maintenance department on root cause failure of
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the chemical erosion of the positive battery posts. According to the licensee, the lead coating on positive posts was remnved due to friction between the posts and individual cell covers from minor thermal expansion and contractions of the cell.
Subsequently, the protective lead coating was removed, and the electrolyte (sulfuric acid) resulted in chemical corrosion of the copper terminal posts.
This condition, evident as red buildup in the electrolyte, was visually noted.
The licensee then re-placed the batteries with ones of a proven post seal design.
The in-spector reviewed licensee surveillance data on installation and capacity checks for the installed station batteries. All surveillance require-ments were met, and was in accordance with the vendors recommended in-stallation testing.
No inadequacies were noted.
The inspector reviewed the licensee cc.r.mitments set forth in the FSAR (Section 8.5.3.2) for emergency loading conditions for the station bat-teries. Accordingly, each station battery is sized to supply total con-nected loads for one hour without non-vital battery charger support.
Loading on the battery is assumed with one of the two vital batteries out of service.
The inspector reviewed the DC load profile under emer-gency conditions and verified the amp-hour loading was within the battery capacity rating.
In May 1987 the licensee generated engineering calculation PA83-156 802-GE, Rev. O, "Determine Class IE Station Battery loading." The purpose of the calculation was to determine the S-hour battery loading profile to satisfy loss of all A/C conditions.
The licensee utilized worst case initial conditions (60 degrees F electrolyte temperature, 1.75 VDC cell voltage, and an aging factor of 1.25) and concluded the vital batteries during a station blackout can be expected to supply emergency loads in excess of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
The licensee evaluated selective load shedding of (non vital) DC components to extend the life of the batteries in excess of twelve (12) hours.
The licensee incorporated the load reduction in SP 2345C, "125 VDC System." The inspector reviewed the procedure to verify the incorporation of load reduction steps in SP 2345C as recom-mended by engineering calculation PA83-156-802-6E, Rev. O.
The inspector reviewed the results of engineering calculation PAS 3-156-802 and the emergency loading results located in FSAR Table 8.5-1.
The inspector questioned the licensee on the differences Letween the 6.9 KV and 4.16 KV breaker trip coil DC control power actuating currents.
Engineering calculation PA 83-156-802 listed the 6.9 KV and 4.16 KV breaker switchgear trip coil loads as 1.9 amps, but the breaker mar;fac-turer's rating was 6.0 amps.
The 480 volt switchgear tripping coils were listed at 2.7 amps; according to manufacturer's ratings, this should be 1.9 amps.
The licensee determined the discrepancies were due to person-nel error, in that the individual who performed the calculation recorded incorrect values. With the incorporation of changes due to the higher trip coil currents for all the breakers, the first minute battery load profile increases from 510 to 595.7 am;s.
Since, during actual condi-
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mately one second before the breaker is tripped, this revised load as-sumption is very conservative.
During the last refueling outage, the battery service test performed used a load current prarile of 510 amps for one minute, 201 amps for 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> and 58 minutes, Jad 241 amps for one minute (=1614 ampere-hours). On July 28, the liransee prepared a justification for continued operation, based on the re:ent engineering calculation used for TS surveillance 4.8.2.3.2.d.
The station batteries are manufactured by C&D company and have one-minute rating of 2560 amps and an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> rating of 2320 ampere-hours.
Using the data from the last capacity discharge test, battery 201A has a capacity of 91.3% and battery 2018 has a capacity of 89.9%.
The last service test at 510 amps used 21.8% of the battery capacity for the first minute for hattery 201A and 22.2% of the battery capacity for battery 201B, If the new value of current (595.7 amps) were used for the one minute duration, 25.5% of the battery capacity for battery 201A would have been used and 25.9% of the battery capacity for battery 201B would have been used.
The total battery capacity used during the last service test was 76.19%
for battery 201A and 77.38% for battery 201B. With the new va?ue of current for one minute, the total battery capacity used would have been 76.26% for battery 201A and 77.44% for battery 2018. Therefore, the battery service test with the new one minute service current is less than the 804 capacity value b3 which battery replacement is planned.
As noted earlier in this report section, the 1888 ampere-hour calculated loading provides an 18% margin to the 2300 ampere-hour rating of the battery.
Correlation of this 18% margin to the 80% capacity replacement criterion and to the service test loading was not accomplished during this inspection and will be further reviewed (UNR 88-16-01).
The Millstone Unit 2 Battery Li'e is 12 to 15 years. Maintenance Pro-cedure SP2736F plots the battery capacity and thus can predict when the battery will decrease its terminal potential to 105 volts in less than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (80% of the manufacturer's rating).
The battery trending data is enclosed in the Battery Service Test (SP2736E) and Battery Performance Discharge Test (SP2736F).
FSAR section 8.5.4.2 itemizes the data which is collected.
Further NRC review of the expected end-of-life battery voltage drop-off characteristics to assess the utility of this trending data is planned.
12.5 Battery Ventilation and Protection from Ignition Hazards The inspector reviewed the station service battery room ventilation system.
The battery rooms are ventilated by cable vault air f rom the switchgear rooms, exhausting through roof exhausters to the atmosphere.
The two battery rooms are intercor.nected to assure venting following
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failure of either of the two 50% capacity exhaust fans (F112A and F1128).
The fans are powered from vital AC Motor-Centrol Centers (MCCs) 22-2E and MCC 22-2F.
Each fan has a rated capacity of 4125 cubic feet / minute (CFM). A flow switch is provided on the cable vault supply far to the respective battery rooms to monitor air flow temperature and to alarm on low flow conditions.
The flow switch maintains supply air to the battery rooms between 70 degrees F and 104 degrees F.
According to the FSAR, the components in the ventilatten system are designed to 3perate in an environment of 110 degrees F, atmospheric pressure, and 100% re-lative humidity.
The inspector noted no inadequacies.
The inspector reviewed control room operatar actions in the event of a loss of one or both exhaust fans. According to IEEE 484-1981, a loss of battery ventilation may be hazardous during battery charging or dis-charging operations due to hydrogen gas build-up.
In operations proce-dure (0P) 23158, "Non-radioactive Ventilation System," Rev. 3 Change 1, for a loss of one exhaust fan, the operator is to verify the non-affected fan is operating.
If both fans are lost, the operator is to monitor battery cells for gassing and install portable booster fans.
The licen-see verifies both exhaust f ans are operating on a shift basis (approxi-mately 8-hour intervals). No inadequacies were noted.
The licensee does not have installed hydrogen monitoring capabilities for the station battery rooms. A charging or discharging battery pro-duces both hydrogen and oxygen.
The licensee's basis for no hydrogen detection is a highly reliable and constant exhaust air flow throughout the battery rooms and battery exhaust flow surveillance.
In the event of a loss of normal pover (LNP), the ext vst fans sequence within 20 seconds onto the vital buses supplied by he emergency diesel generators.
Additionally, the inspector independently calculated, using IEEE 4S4-1981, that for the worst case conditions, a fully charged battery with an as-sumed exhaust fan out of service resulted in fa* less than 2% hydrogen accumulation of the total battery area.
The inspector reviewed licensee
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technical specification (TS) surveillances for the station batteries with respect to precautionary steps concerning ventilation system operation.
No inadequacies were noted.
On June 29, the inspector toured both vital station battery rooms.
The inspection included review of individual cell electrolyte levels, top-to-jar seals, caps on filler openings, individual cell flash arrestors, location of heat sources, location of loose equipment and combustibles, and general housekeeping of the area. Minor deficiencies were noted on station battery 2018.
Cells 40, 14, 25 appeared to have indications of electrolyte on top of each cell and surrounding floor around the area.
The Exide computer battery had a terminal post cap seal crack. The above discrepancies were discussed with the licensee, and the licensee provided appropriate resolution (cell cleanup and cap replacement).
Overall, the vital battery rooms were in good condition. No loose equipment or com-bustible material was present. On the tour, the inspector discussed with a plant equipment operator the required tour s of the battery room. The
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operator explained the tour frequency within the battery rooms as twice an 8-hour shift. The licensee's tour consisted of general housekeeping, adequate exhaust flow, cell electrolyte levels, and local battery voltage readings.
No inadequacies were noted.
The inspector reviewed the Fire Hazards Analysis for the battery rooms.
The battery rooms are designated as fire zones A-22 and A-23.
The review consisted of early warning ionization fire detection, location of fire extinguishers, fire hoses adjacent to the area, and consequences of an assumed fire.
No inadequacies were noted.
The inspector noted one potential water-carrying pipe inside the battery rooms, a drain line from the cable vault to remove water from fire sup-pression system activation.
The inspector, in discussions with the lic-ensee, reviewed past NRC concerns on the drain line from the cable vault room.
NRC inspection report 50-336/84-15 detailed a noncompliance with 10 CFR 50 Appendix R, Section II.C.4. and IIIG.2., concerning inadequate redundant trains separated by a 3-hour fire barrier.
The unsealed pene-trations of the drain pipe were identified as the noncompliance.
NRC inspection report 50-336/86-02 closed out the above item by verification of licensee's corrective actions and actual inspections of the battery room.
The inspector reviewed seismic qualification for the drain line connection.
The licensee evaluated and installed the seismic fasteners for the drain line per idCR 2-54-85.
No inadequacies were noted.
12.6 Electrolyte Temperature and Specific Gravity Industry code IEEE 484-1981 identifies a basis for rated battery per-formance as a cell temperature of 77 degrees F.
Different cells should have the same electrolyte temperatura, insofar as is practicable.
The inspector's tour of the station battery rooms did not identify localized neat sources (such as radiators, steam pipes or space heaters) which could alter temperatures of individual cells.
The inspector reviewed tFe last tt n weekly surveillance data sheets, $P-2736A, "Battery Pilot Cell Surve llance," Rev. I for both 201A and 2018 station service battery pilot cell;. All pilot cells were within 5 degrees F of each other as recommended in IEEE 484-1981.
The acceptaore criteria for individual pilot cells corrected to 77 degrees F is an electrolyte specific gravity greater than or equal to 1.200. All surveillances reviewed met the acceptance criteria.
No inadequacies were noted.
12.7 Charging The inspector reviewed maintenance procedure MP 2720F6, "Station Battery Cell Charging," Rev. 3.
The licensee utilized vendor (C&D Batteries)
recommended objectives for conducting individual cell charging.
The acceptance criterion for a charged cell is an individual cell voltage of 2.35 volts +/-0.05 volts. Also, cell specific gravities shall not vary by more than 5 points (0.005).
No inadequacies were identifie _ _ _ _ _
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The inspector reviewed the accomplishment of equalizing charges on sta-tion service batteries.
This is a charge given at a voltage higher than nominal float voltage for a definite number of hou.$ depending on the value of the charge voltage.
The licensee accomplishes an equalizing charge on the station batteries following the performance of SP 2736E,
"Station Service Test," or SP 2736F, "Battery Performance Discharge Test."
The restoration from the above surveillances is accomplished with operat-ing procedure (0P) 2345C, "125v D.C. System," SP 27368, "Complete Battery
Cell Measurement," and SP 2736G, "Battery Charger Capacity Test." The i
inspector was unclear as to what other criteria that may require an equalizing ctarge on the station batteries other than in 18-month or 60-month surveillances.
In reference to $P-2736A Rev. 1, "Battery Pilot Cell Surveillances," and SP-2736B, "Complete Battery Cell Measurements,"
procedures for TS (weekly and 90-day) surveillance requirements precau-tion step 6.3 adds...
"If the battery must be in ' equalize' notify operations." According to the vendor manual, minimum acceptable cell voltage and specific gravity readings should be utilized when determining the need for an equalizing charge.
This item will be tracked as open pending licensee review and evaluation of other required conditions for placing an equalizing charge on the station batteries in accordance with vendor recommended practices (UNR 83-16-02).
OP2345C specifies the instructions for batteries on oper, current.
How-ever, cautions do not exist to preclude r9 storing vital batteries after extended periods on open circuit. TSs require one battery and charger to be operable in Modes 5 & 6.
T5 surveillarce requirements do not have to be performed on inoperable equipment.
Licensee evaluattun of the need
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for further guidance on restoration of an open-circuited battery to ute
(e.g., performance of an equalizing or freshening charge) appears appro-priate.
Lead-acid batteries lose charge when removed from a voltage source that is higher than the open circuit potential of the battery.
As this charge
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is lost, the electrochemical process produces lead sulfate in both the
positive and negative plates of each battery.
If left uncharged for a significant period of time, lead sulfate will begin to form in large
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c rystal s.
These large crystals may be difficult to reduce through acrmal charging and may thereby inhibit the electrochemical process needea to
sustain a healthy lead-acid battery.
Higher than normal charging poten-tials or more sophisticated rer,edial approaches may be necessary to re-cover the affected battery. With severe sulfation, replacement may be necessary.
Therefort, batteries should be given prompt initial charging
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and, if required to remain out of service af ter initial charging, the i
charging process may need to be repeated at least every three months for up to a year from initial shipment.
Procedures may need to address the need for a freshening or equalizing charge when the battery is open cir-cuited for an extended period of time.
This is an unrosolved item (UNR 83-16-03) pending licensee determination of the need for such charges, i
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The inspector reviewed the licensee's control of battery water quality.
Based on discussions with the licensee's maintenance department, water quality is confirmed by chemistry prior to use in the cells.
The in-spector noted that TS surveillance proceaures SP-2736A and SP-2736B re-quire the water container to be marked "distilled" and the amount o'
water added to a cell to be documented.
No inadequacies were noted.
A review of specific getvity readings following a battery charge as it related to electrolyte stratification was conducted. The inspector re-viewed licensee addressal of this issue in NRC Information Notice 85-74.
The licensee takes a complete set of individual cell voltage readings following a recharge ano equalization.
Specific gravity readings are not taken at the immediate completion of the charge because of potential electrolyte stratification, with remixing of electrolytv taking a number of days. The inspector verified this conclusio!. with the vendor's tech-nical manual. The individual call raadings for specific gravity are taken at three-month TS surveillance intervals (SP-27368) with a minimum of two days of float operation prior to the surveillance.
The inspector reviewed the establishment and maintenance of float voltage in accordance with the vendor's instructions.
The licensen maintains the recommended nominal float voltage for each of the station batteries as recommended by the vendor.
Industry code IEEE-389 specifies that a single-cell charger, if used, not violate class IE independence from non-class IE equipment.
The lic-ensee determined that, if an individual cell charger is powered from an AC source compatible with the battery, there would not be a separation problem.
Procedure MP 2720F6 was revised to require compatible power supplies between individual cell chargers and station batteries. The inspector verified implementation of this procedural step in MP 2720F6.
No inadequacies were noted.
12.8 performance Tests and Replacement Criteria Periodic capacity te:ts assure adequate battery capacity.
They include acceptance tests to assure adequate initial capacity, servie; tests to show the al,ility to carry the design load profile, and periodic capacity tests to assure maintenance of capability.
323 are relevant industry codes, The inspector reviewed the licensee's acceptance testing for the November 1979 replacement of station batteries under product acceptance test work orders 2-535-79 and 2-509-79.
The licensee evaluation was well docu-rented, and acceptance criteria were satisfiad.
No inadequacies were noted in testing and evaluation of results.
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and Battery Performance Discharge Test (SP-2736E) and Battery Performance Discharge Test (SP-2736F) for both station batteries. No inadequacies were noted. - All acceptance criteria were satisfied.
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13.0 NRC~Information Notice (IN) 88-38, "Failure of Undervoltage Trip Attachments
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of General Electric (GE) Circuit Breakers" (92717)
The-NRC issued IN 88-38 on June 15, 1988 to alert licensees of potential
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problems resulting from the failure of undervoltaga trips of GE circuit
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breakers AK-2-15 and AK-2-25.
The failure mechanism was internal binding of the undervoltage device linkage.
The internal binding was due to inadequate
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clearance between the armature and trip paddle.
The inspector reviewed IN 88-38 for applicability to Millstone 2.
GE AK-2-25 breakers with undervoltage devices are installed as reactor trip breakers (RT8s). AK-2-25 breakers'are also used in other safety systems; however, those other breakers are not equipped with undervoltage trips and are not affected by information promulgated by IN 88-38.
The inspector reviewed past NRC ins and bulletins related to GE AK-2-25 breakers.
The documents reviewed were: IN 83-18, Failure of Undervoltage Trip Function of Reactor Trip System Breakers, IN 85-58, Failure of IE Type AK-2-25
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Reactor Trip Breakers, and NRC Bulletin 83-04, Failure of Undervoltage Trip Function of Reactor Trip Breakers.
Inspector review of actions taken en the above ins and of the licensee response to and action on Bulletin 83-04 con-cluded that the licensee was responsive to past NRC concerns in relation to the referenced bulletins and information notices.
The inspector reviewed licensee surveillance procedures related to internal binding of undervoltage device linkages.
Station procedure PT-21429, Reactor Trip Switchgear Shunt Trip and Undervoltage Device Test, establishes a uniform method of inspection and test of GE Undervoltage Trip devices used in reactor trip switchgear, and determines the response time of the breaker to a 1 css of control power. A technical reference utilized to develop PT-21429 is GE Service Advice Letter (SAL) 175-9.3 (April.15,1983).
The SAL vas provided
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by the NRC to all power reactors to inform them of potential problems with GE AK-2-25 oreakers.
The licensee incorporated the SAL preventive maintenance (PM) recommendations into procedure PT-21429.
One specific pH was an adjustment between the.ndervoltage trip device and the trip paddle on the t-ip shaf t.
The action develops a positive trip and 1/32 of an inch of additional travel of the armature.
The 1 ten is addressed in step 7.4.1 of PT-21429.
The overtravel ensures the trip paddle moves more than the amount necessary to trip the breaker. A turning screw on the front
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face of the breaker allows for adjustment between the armature and trip paddle.
The same PM is also performed by the licensee in maintenance procedure 2701J-1, both at a refueling outage frequency.
The licensee reported to the inspector l
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that records do not show that-any Millstone' Unit 2 RTB has failed to trip when required.
This statement was supported by the licensee's response to.NRC Bulletin 83-04.
The inspector reviewed TS surveillance procedures SP-2401A, "Manual Reactor Trip Test," and SP-2401D, "Reactor Protection System Matrix Logic and Trip Path Relay Test," against TS 3.3.1.1, "Reactor Protection Instrumentation."
The inspector verified the test was within the acceptance criteria for the
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past three surveillances.
No inadequacies were noted.
The inspector conciuded that the licensee's surveillance program adequately addresses the concerns in NRC IN 88-38.
The licensee's active role with the vendor (General Electric) has been accompanied by good performance to date on the safety-related RTBs.
The insrector has no further questions on this matter.
14.0 Safety System Functional Inspection (SSFI)
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By letter dated April 4, 1988 the licensee notified the NRC of tht'r decision to assess the operational readiness of a selected safety system to function under operational and analyzed accident conditicas at Millstone 2.
The in-ternal assessment is similar to the NRC's SSFI.
The NRC has encouraged indi-vidual licensee participaticn in such an evaluation program.
The licensee's program structure generally followed the format of NRC's IE Manual Chapter 2515, Appendix C, as a guideline for the self-3ssessment program.
The licensee's safety system for selected assessment was the Reactor Building Component Cooling Water (RBCCW) system.
RBCCW transfers heat from structures, systems and components important to primary plant and safety system operation to the service water system. The licensee selected the RBCCW system based on its importance to vital systems and on the recommendation from the licen-see's-corporate Probability Risk Assessment (PRA) branch.
The assessment team consisted of 11 individuals not associated with the lic-ensee's staff at Millstone 2.
One staff member frcm Millstone 2 acted as liaison between the unit and the assessment team.
Team members were affili-ated with Millstone 1, Millstone 3, Haddam Neck, and the licensee's corporate office.
The objective of the licensee's SSFI was to assess the operational readiness including: whether the RBCCW is capable of performing safety functions; whether testing is adequate to determine if the RBCCW system performs all of its safety functions; whether RBCCW maintenance is adequate to ensure oper-ability; whether training is adequate to ensure proper operation; human fac-tors relating to RBCCW; and whether management controls including procedures are adequate to ensure that RBCCW will fulfill its intended functions.
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As of July 22, the assessment team had compiled a list of 32 observations.
The observations concerned issues such as replaced component seismic review, temporary installations, drawing control and errors, seismic restraints for temporary equipment, labeling, and training task job analysis.
The observa-tions were presented to the Millstone 2 staff daily for disposition and for concurrence on the ac'equacy of the disposition from the assessment team.
The SSFI is scheduled for completion by August 5, 1988.
Final report issuance is to be early in September.
The inspector will review the final report in future inspections.
The inspector concluded that the SSFI incentive, basis, working format, and utilization of licensee manpower was good.
15.0 Licensee Event Report..ns) Review (92700)
A Licensee Event Report submitted during the period was reviewed to assess LER accuracy, the adequacy of corrective actions and compliance with 10 CFR 50.73 reporting requirenents, and to determine if there were any generic
.m-plications or if further information was required.
The LER reviewed was:
88-009, Unrecoverable Dropped CEA Due to Upper Gripper Failure.
The inspector reviewed the contents of the LER in routine inspection report 50-336/88-13.
No inadequacies were noted.
16.0 Review of Periodic Reports (90713]
Upon receipt, a periodic report submitted pursuant to Technical Specifications were reviewed.
This review verified that the reported information was vclid and included the NRC required data, and that test results and supporting in-formation were consistent with design predictions and performance specifica-tions.
The inspector also ascertained whether any repor+.d information should be classified as an abnormal occurrence.
The following report was reviewed:
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Monthly Operating Report for Millstone Unit 2 for June 1988 17.0 Committee Activities (40700)
The inspector attended Plant Operations Review Committee (PORC) meetings on July 6 and July 13.
The inspector noted by observation and from the written record that committee administrative requirements were met for the meetings, and that the committee discharged its functions in accordance with regulatory requirements.
The inspector observed a thorough discussion of ma+.ters before the PORC and a good regard for safety in the issues under consideration.
No inadequacies were identified.
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'18.0 Management Meetings (30703)
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Periodic meetings.were held with station management to discuss inspection
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' findings during the inspection period. A summary.of findings was al so 'di s-cussed at the conclusion of the inspection.
No' proprietary information was-
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covered within the scope of the-inspection. -No written material as given-
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to'-the licensee during the inspection period.
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