IR 05000336/1990001
| ML20033G926 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 04/02/1990 |
| From: | Haverkamp D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20033G921 | List: |
| References | |
| 50-336-90-01, 50-336-90-1, NUDOCS 9004130076 | |
| Download: ML20033G926 (14) | |
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U.S. NUCLEAR REGULATORY COMMISSION
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REGION I
Report No.:
50-336/90-01
Docket No.:
50-336 License No.
DPR-65 Licensee:
Northeast Nuclear Energy Company P.O. Box 270
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Hartford, CT 06141-0270 Facility Name: Millstone Nuclear power Station, Unit _2 Inspection at: Waterford, Connecticut Dates:
January 20, 1990 - March 5, 1990 t
Reporting Inspector:
peter J. Habichorst, Resident Inspector Inspectors:
William J. Raymond, Senior Resident Inspector Peter J. Habighorst, Resident Inspector Brian Hughes, Operator Engineer, PWRS, OB; ORS
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Guy S. Vissing, Project Manager, Project Directorate I-4, NRR Approved by:
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w [- j 3 TItITc Donald R. Haverkamp, Chigf
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Reactor Projects Section 4A l
Division of Reactor Projects
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Inspection Summary:
Inspection on January 20,1990 - March 5,1990 (Inspection Report No. 50-336/89-24 Areas Inspected:
Routine NRC resident and specialist inspection of plant operations, surveillance, maintenance, previously
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identified items, engineering / technical support, committee activities, evaluation of licensee self-assessment, Licensee Event Reports (LERs), and Security Event Reports.
Results:
1.
General Conclusions on Adequacy, Strength or Weakness in Licensee Program Properly planned and executed routine containment entries were noted. (Section 4.1)
One example of insufficient attention to detail on a procedure for valve line-ups was noted.
This appears to have been an isolated case based on past inspection review. (Section 3.3)
9004130076 900404
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PDR ADOCK 05000336 O
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Adequate licensee follow-up actions on previous NRC Bulletins werenoted.(Section6.1and6.2)
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Violations Within the scope of this inspection, a violation was noted, which concerned failure to document a Licensee Event Report (LER) for a condition prohibited by the plant's Technical Specification. Specifically, the control room emergency air clean-up system was not aligned in the recirculation mode as required by technical specification 3.7.6,1.a.
,d no LER was submitted to NRC (Section 3.4)
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Unresolved Items Two previously unresolved items were closed.
(Sections 5.3.1 and 5.3.2) No unresolved items were opened during the inspec61on period.
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TABLE OF CONTENTS
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1.0 Persons Contacted......................................
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2.0 Summary of Facility Activities.........................
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3.0 Plant Operations (IP 71707/71710/93702)................
3.1 Control Room Observations.........................
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3.2 Plant Tours.......................................
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3.3 Stand-by Readiness of Engineered Safety Features Systems and System Wa1kdown......................
3.4 Review of Plant Incident Reports..................
4.0 Radiological Controls (IP 71707/92701).................
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4.1 Posting and Control of Radiological Areas.........
5.0 Maintenance / Surveillance (IP 62703/61726/92702)........
5.1 Observation of Maintenance Activities.............
5.2 Observation of Surveillance Activities............
5.3 Previously Identified Items.......................
5.3.1 (Closed)UnresolvedItem 89-16-02, Slow Start Time for "A" Emergency Diesel Generator...........
5.3.2 (Closed) Unresolved Item 89-17-05 Licensee
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Acquisition of Design Specification and Control of Specification for Emergency Diesel Generator
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Voltage Contro1...................................
6.0 Engineering / Technical Support (IP 92702)...............
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6.1 NRC Bulletin 83-05, ASME Nuclear Code Pumps and Spare Parts Manufactured by Hayward Tyler Pump Company........................
6.2 NRC Bulletin 83-07, Apparently Fraudulent
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Products Sold by Ray Miller, Inc..................
6.3 Temporary Instruction RI-86-03, General Electric HGA Relays in Safety-Related Applications.............
6.4 Temporary Instruction No. RI-86-02, General Electric Type AK-F-2-25 Breakers...........................
7.0 S e c u ri ty ( I P 71707 )....................................
7.1 Security Tours....................................
7.2 Illegal Substance Discovered Outside Protected Area....................................
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Table of Contents
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Page 8.0 Safety Assessment / Quality Verification (IP 71707).............................................
8.2 Committee Activities..............................
9.0 Management Meetings (30703)...........................
The NRC inspection manual inspection procedure (IP) or temporary instruction (TI) that was used as inspection guidance is listed for each applicable report section.
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i DETAILS
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1.0 Persons Contacted Interviews and discussions were conducted with licensee staff and management during the report period to obtain information pertinent to the
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areas inspected.
Inspection findings were discussed periodically with the supervisory and management personnel identified below.
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- S. Scace, Nuclear Station Director, Millstone Station l
- J. Keenan, Nuclear Unit Director, Millstone Unit 2 J. Riley, Maintenance Manager, Millstone Unit 2 J. Becker, Instrument and Controls Manager, Millstone Unit 2
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J. Smith, Operations Manager, Millstone Unit 2
- Attendee at post-inspection exit metting on March 23, 1990.
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2.0 Summary of Facility Activities Millstone Nuclear Power Station (Millstone 2 or the plant) operated at rated thermal power throughout the inspection period.
Daily management meetings provided good discussion and awareness of reactor coolant system pressure boundary operating concerns associated with the steam generator tubes. The awareness was manifested in observation performance of the steam jet air ejector radiation monitor, monitoring trends of leak rate rate calculational results (primary to secondary), and chemistry department sampling frequency and recommendations.
The calculated leakage rates averaged six to seven gallons per day (GPD) with maximum values at fourteen OPD during the inspection period.
The required limit as prescribed in the Technical Specifications is 144 GPD.
NRC Activities A Region I specialist inspection of acoustic valve monitor system alle-gations and IE Bulletin 80-06 was conducted between January 22-26, 1990.
Results are reported in inspection report 50-336/90-02.
A Region I specialist inspection of in-service testing of pumps and valves, and previously identified items was conducted between January 22-26, 1990.
Results are reported in inspection report 50-336/90-03.
On February 22, the licensee presented results of the November,1989 steam generator tube inspection and associated safety assessment to the NRC
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staff in Rockville, Maryland.
The inspection activities during this report period included 125 hours0.00145 days <br />0.0347 hours <br />2.066799e-4 weeks <br />4.75625e-5 months <br /> of inspection during normal activity working hours.
In addition, the review of plant operations was routinely conducted during periods of backshifts (evening shifts) and deep backshifts (weekend and midnight shifts).
Inspection coverage was provided for eleven hours during backshifts and three hours during deep backshift.
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3.0 Plant Operations
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3.1 Control Room Observations Control room instruments were observed for correlation between channels, proper functioning, and conformance with Technical Speci-
fications. Alarm conditions in effect and ala ms received in the control room were discussed with operators.
The inspector periodi-cally reviewed the night order log, tagout log, Plant Incident Report
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(PIR) log, key log, and bypass jumper log.
Each of the respective logs was discussed with operations department staff.
No inadequacies
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were noted.
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3.2 Plant Tours
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The inspector observed plant operations during regular and backshift tours of the following areas:
Control Room Vital Switchgear Room Diesel Generator Room Turbine Building Intake Structure Enclosure Building ESF Cubicles Loas and records were reviewed to ensure compliance with station procedures, to determine if entries were correctly made, and to
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verily correct communication and equipment status.
No inadequacies were noted.
3.3 Stand-by Readiness of Engineered Safety Features Systems and System Walkdown On January 30 and 31, th.ee engineered safety feature (ESF) systems were reviewed to verify system operability.
The systems reviewed were Facility I containment spray system, low pressure safety injec-tion system and auxiliary feedwater system. The review included proper positioning of major flowpath valves, proper operation of indication and controls, and visual inspection for proper lubrica-tion, cooling, and other conditions.
References used were:
-- Final Safety Analysis Report
-- Plant instrument and piping diagrams (P& ids)
25203-26015, 25203-26005, and 25203-26017
-- Station procedures (SP) 2606C-2, 2604L, and 2610C-2
-- Control room alarm book During preparation for the Facility I containment spray system walk-down, the inspector questioned the availability of the required boration flowpath from the containment spray system to the chemical and volume control system (CVCS) as prescribed in technical speci-fication action statement 3.1.2.2.b.
Specifically, valve line-up
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procedure SP 26060-2 changes 2 and 3 altered the position of valve i
2-05-028 from locked closed to required open.
Valve 2-05-028 is the refueling water storage tank supply to the CVCS system.
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inspector presented this discrepancy to the licensee to determine if
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the required boration flowpath was compromised between November 21,
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1989 (Change 2 to SP 2606C-2) to December 9, 1989 (Change 3 to SP i
26060-2). On January 30, during the containment spray system
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walkdown the inspector verified the correct position (open) of valve
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2-CS-028.
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On February 15, the licensee documented the results of the investi-gation per internal document MP-2-0-255. The licensee concluded that t
valve 2-CS-028 was open between the procedural revisions, and that the discrepancy between the required valve position was administra-tive based on the inadequate transcription of procedure valve line-up
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changes.
Inspector review of the licensee's investigation and i
chronology of events concluded that the required boration flowpath via valve 2-05-028 was not compromised; however, attention to detail in procedure valve line-up changes was deficient in this particular
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case. The inspector has no further questions in regards to this
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matter.
Inspector walkdown verified the ESF systems were in the appropriate standby mode with no inadequacies noted.
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3.4 Review of Plant Incident Reports (PIRs)
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Of approximately 130 plant incident reports (PIRs) that were developed during 1989, 11 were selected at random to determine if they were correctly determined to be reportable by the licensee as LERs.
The following provides the PIRs that were reviewed and the licensee's determination for reportability:
l PIR LICENSEE REPORTABILITY DETERMINATION 89-1 NO 89-16 NO 89-24 NO 89-27 N0*
89-34 N0 89-59 YES 89-71 NO 89-88 NO 89-99 NO l
89-111 NO
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- This was later reported in LER 89-004-02
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i Inspector review of the licensee's reportability determinations were
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acceptable except for PIR 89-24.
PIR 89-24 relates to an apparent technical specification violation.
Deviations from the technical
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specifications are reportable as an LER within 30 days after
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discovery of.the event in accordance with 10 CFR 50.73(a)(2)(1)(B).
While in Mode 6 on February 28, 1989, the "A" control room emergency air clean-up (CRAC) system was declared inoperable for greater than seven days. This is because the "A" diesel generator was out of service for maintenance. The normal offsite AC power was available and the control room ventilation was operating on offsite power.
t Technical specification 3/4.7.6.1 requires, with one control room emergency air clean-up system inoperable, (based on lack of emergency power source) restore the inoperable system to operable status within seven days or initiate and maintain operation of the remaining
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operable control room emergency air clean-up systems in the recir-
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culation mode.
Theplantwasoperatingwith"B"CRACnotinthe recirculation mode for 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> after
'A" CRAC was declared
inoperable. This was considered by the inspector as a condition prohibited by the technical specifications and thus required a 30-day LER.
No LER was submitted.
This is a violation (336/90-01-01).
After 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />, the "B" CRAC was initiated in recirculation mode of
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operation.
This resulted in both "A" and "B" CRAC systems in i
operation from offsite AC power.
The licensee did not consider this a violation since technical
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specification 3.8.1.2 allows one diesel generator to be out of
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service in mode 6 provided one offsite AC power source and one diesel generator are operable.
The licensee considered that this was an inconsistency in the technical specifications and that 3.8.1.2 was a controlling technical specification. Therefore, in the licensee's assessment this incident was not a violation of the technical
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specifications.
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The licensee considered that a revision to the technical specifica-tions was necessary but had not fully determined the appropriate wording to be applied. A proposed technical specification change was in preparation.
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A technicc1 specification is violated when the administrative controls no longer assure that the requisite LCOs intended to be in place for a given operating activity / mode are met.
That is, in this case, in order to operate in mode 5, the TS require:
(1) two CRAC systems operable with backup diesel generators (DGs);
(2) if (1) cannot be met because one CRAC system is inoperable for more than 7 days, then the second CRAC system with backup DG supply must be taken out of standby operations and put on recirculation (from passive to active service); (3) if cannot meet either (1) or (2), then no fuel movement or positive reactivity changes are allowed.
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For 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> on February 27-28, when the 'A' CRAC system was inoperable for more than 7 days and 'B' CRAC system was not in recirculation, Millstone 2 no longer had the assurance intended by the LCO that control room habitability would be assured from possible source terms. There was no prohibition placed by the licensee on plant activities at the time.
The fact that no core alterations or positive reactivity changes took place during the period of interest is a mitigating circumstance that makes the significance of the LCO violation minimal.
Inspector concerns on the above matters were discussed in a meeting with the Unit 2 director on March 2, 1990.
The licensee acknowledged the inspector's comments.
4.0 R,adiological Controls 4.1 Posting ard Control of Radiological Areas During plant tours, contaminated, high airborne radiation, and high radiation areas were reviewed with respect to boundary identifica-tion, locking requirements, and appropriate control points.
No inadequacies were noted.
Upon exiting the radiological control area (RCA), the inspector was found to be contaminated.
This was due to airborne particulates, which was exacerbated due to ventilation system maintenance in progress. A log maintained at the RCA indicates that personnel contamination is a common occurrence at Unit 2.
Further improvements in preventing contamination should be pursued.
Review of two routine containment entry RWPs (2-90-66 and 2-90-67)
indicated they were properly executed and had the required approvals.
The entries were well planned and executed.
5.0 Maintenance / Surveillance 5.1 Observation of Maintenance Activities The inspector observed and reviewed selected portions of preventive and corrective maintenance to verify compliance with regulations, use of administrative and maintenance procedures, compliance with codes and standards, proper QA/QC involvement, use of bypass jumpers and safety tags, personnel protection, and equipment alignment and retest. The following activity was observed:
r AWO M2-90-01269, B Charging Pump
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No inadequacies were identified.
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5.2 Observation of Surveillance Activities The inspector observed portions of surveillance tests to assess performance in accordance with approved procedures and Limiting Conditions of Operation, removal and restoration of equipment, and deficiency review and resolution.
The following tests were reviewed:
OP 26068, facility 2 core spray pump operability test, 2/26/90 j
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$P Quarterly !$1 testing of core spray system valves - facility
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2, 2/26/90
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Inservice inspection of B charging pump
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$3 Previously Identified Items 5.3.1 (Closed) Unresolved Item 89-16-02. Slow Start Time for "A" Emergency Diesel Generator Discussion with the licensee determined that a major replacement of components in the voltage / excitation circuitry has reduced the time required to obtain the rated voltage to an acceptable value.
Subsequent testing and routine surveillance testing has demonstrated satisfactory performance of the A emergency diesel generator.
5.3.2 (Closed) Unresolved Item 89-17-05. Licensee Acquisition of Design Specification and Control of Specification for Emergency Diesel Generator Voltage Control The design reconstitution of the saturable transformers was found effective. The licensee has taken extensive inplace measurements including windings, excitation, currents, dimensions, and location of
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hook up wires.
The original vendor (Basler Electric) has provided field support in this effort.
The licensee now has the documentation and drawings of the saturable transformers.
The EDG testing has
demonstrated the reliability of the system, this item is therefore closed.
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6.0 Engineerino/ Technical Support 6.1 NRC Bulletin 83-05. A$ME Nuclear Code Pumps and Spare
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parts Manufactured by Hayward Tyler Pump Company
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The bulletin discusses NRC findings pertaining to allegations that the Hayward Tyler Pump Company (HTPC) failed to effectively implement
its QA program during the manufacture of some ASME nuclear code pumps and spare parts. The discrepancies with the HTPC QA program occurred
between 1977 and 1981.
Northeast Utilities responded by letter dated
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August 17, 1983, which informed the NRC that three service water pumps manufactured by HTPC were installed before 1977 and therefore, only the actions of the bulletin related to spare parts were appli-cable. This is correct since the pumps were originally installed when the plant was licensed in 1975.
Bulletin items 2a, 2b, 2c, and 2d were addressed in NU letter dated August 17, 1983, and an internal enemorandum dated July 15, 1983.
Item 2a, which related to hand fitting of new pump parts and dimensionally checking of new rings and other parts was verified in the overhaul procedures of the service water pumps.
Item 2b relates to inservice testing of pumps in accordance with section 11 of the ASME code. During the most recent surveillance tests. (procedures $P 21104, 21103, and 21102) of November 7, 1989 January 8,1990 and January 8,1990, respectively, the pumps were checked for flow, pressure, support system flow (i.e. lube and chilled water), and vibration.
The recent surveillance test results were acceptable.
Item 2e relates to performance testing of the pumps.
Performance testing is done at normal surveillance intervals.
The latest per-formance testing was done on two pumps, February 26, 1990 and February 12, 1990. Before a service water pump is declared operational following any pump maintenance, baseline ISI pump testing is accomplished.
Pump motor current is routinely monitored in the control room, while pump packing leakage is checked as part of the
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operators' normal rounds.
Item 2d relates to hydrostatic testing of pressure boundary pumps parts.
Plant design change record (PDCR) 2-64-83 provided for the replacement of the pump columns for all three service wate'r pumps.
The original carbon steel columns were replaced with stainless steel columns, The PDCR required hydrostatic testing of the columns and thus the purchase order required hydrostatic testing.
The columns were replaced during the following outage.
All applicable actions specified by NRC Bulletin 83-05 were verified completed and thus the bulletin is closed.
6.2 NRC Bulletin 83-07. Apparently Fraudulent Products Sold by Ray Miller. Inc
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Bulletin 83-07, and Supplements 1 and 2 identified Millstone 2 as a user of Ray Miller products and requested certain actions.
Northeast Utilities responded by letters dated March 27, 1984, and June 21,
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1985, in which they indicated that only nine parts were furnished by
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Ray Miller for Millstone 2 and were installed on the reactor vessel head decontamination mixing tank.
The tank is not a OA category I or
radwaste QA piece of equipment and has never been used.
Review of i
internal memoranda on May 9, 1983 and on May 28, 1985 indicated that
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the reactor vessel head decontamination system was not a QA system, has never been used, and was disabled such that it could not be used and there are currently no plans for it ever being used. A review of the drawings of the system determined that the system was designed to operate at atmospheric pressure and therefore parts specified for use on tanks were overdesigned at 3000 psig.
The licensee's examination of the tank determined that six of the nine fittings were correctly marked and in good condition.
Three 1 1/2 inch half couplings on top of the tank were not accessible for inspection but were considered acceptable based on detail drawings and inspection of pipe runs to the tank. A review of detati drawings, internal memoranda and discussions with plant personnel verifies the closure of this bulletin.
6.3 Tem orary Instruction RI-86-03, General Electric HGA J
helays in Safety-Related Applications
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By letter dated March 31, 1983. General Electric informed Northeast Utilities that some seismic data for HGAll and HGA111 relays were found in error. As a result, Northeast Utilities implemented a plant design change record (PDCR 83-094) to replace all such relays employed in safety-related functions with seismically qualified Telemecanique (formerly Gould) J10 and J13 industrial control relays.
Fif ty-three relays located in the Millstone 2 4.16KV distribution system, high pressure safety injection system, containment spray system, charging system, service water system, boric acid system, containment air recirculation fan system, cor. trol room A/C system,
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steam air ejection system, and the containment and enclosure building air handling system were replaced by work orders from October 15, 1986 through December 5, 1986.
The PDCR and the work orders were reviewed to verify the closeout of this TI.
The AC relays were replaced on a one-for-one basis with seismically qualified, Class IE Telemecanique J10 AC industrial control relays and the DC relays were replaced on a one-for-one basis with seismically qualified Class II Telemecanique J13 DC industrial control relays. The work orders established the replacement and provided for testing for acceptance.
The requirements for this TI dated October 21, 1986, was to determine if GE type HGA relays were used in functions important to safety and to make note of potential safety consequences of maloperation.
It was determined that these types of relays were used in safety-related functions anc that they have all been replaced with acceptable relays.
Thus, this temporary instruction is closed.
6.4 Tem >orary_ Instruction No. RI-86-02. General Electric Type AK-2-25' Breakers The objective of this TI was to determine if the General Electric Type AK-F-2-25 breakers were used at Millstone 2 in functions that are important to safety and to review what preventive maintenance /
testing and operating experience is associated with these breakers.
Interviews with the plant personnel indicated that Millstone 2 does
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To verify
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this, a computer search of the PMMS did not identify any AK-F-2-25 f
breakers in the system since November 1987. A review of the 125 r
DC load center circuit breaker settings revealed no breakers of the
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concerned type.
In addition, the single line 125 volt DC and 125 volt AC vital systems drawings were reviewed to determine if any of
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the type AK-F-2-25 were in these systems. No GE AK-F-2-25 could be
located in any systems which were likely to have these type and size L
breakers. Therefore, this review determined that no GE AK-F-2-25 breakers are in safety-related functions at Millstone 2 and thus provides closeout of this TI.
LO Security
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7.1 Security Tours Selected aspects of site security including site access controls, personnel searches, personnel monitoring, placement of physical barriers, compensatory measures, guard force staffing, and response to alarms and degraded conditions, were verified to be proper during inspection tours.
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7.2 Illegal Substance Discovered Outside Protected Area
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On February 2,1990, the security department, during a vehicle search at the vehicle access point, discovered two marijuana butts in a
cigarette box.
The illegal substance did not enter the protected
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area.
The investigation revealed that one individual contractor admitted ownership and at no time did this person have unescorted l
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access to the facility.
The local police department confiscated the
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illegal substance and have a report on file.
The inspector considered the licensee security management response to this event to be appropriate.
8.0 $afety Assessment / Quality Verification 8.1 Committee Activities
The inspector attended meeting 2-90-13 of the Plant Operations Review
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Committee (PORC) on January 31. The inspector noted by observation
that committee administrative requirements were met for the meeting, and that the committee discharged its functions in accordance with regulatory requirements.
The inspector observed a thorough discussion of matters before the PORC and a good regard for safety in
the issues under consideration by the committee.
No inadequacies i
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9.0 Management Meetines Periodic meetings were held with station management to discuss inspection findings during the inspection period. A summary of findings was also discussed at the conclusion of the inspection. No proprietary information was covered within the scope of the inspection. No written material was given to the licensee during the inspection period.
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