IR 05000423/1989003
| ML20247C844 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 05/03/1989 |
| From: | Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20247C835 | List: |
| References | |
| 50-423-89-03, 50-423-89-3, NUDOCS 8905250063 | |
| Download: ML20247C844 (18) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
50-423/89-03 Docket No.
50-423 License No.
NPF-49 Licensee: Northeast Nuclear Energy Company P.O. Box 270 Hartford, CT 06101-0270 Facility Name: Millstone Nuclear Power Station, Unit 3 Inspection At: Waterford, Connecticut Inspection Conducted: February 28 - April 4,1989 Reporting Inspector: G. S. Barber, Resident Inspector Inspectors: W. J. Raymond, Senior Resident Inspector G. S. Barber, Resident Inspector J. E. Carrasco, Reactor Engineer Approved by: M O d M. h s/7/6S E. C. McCabe, Chief', Reactor Projects Section IB Date Inspection Summary: Inspection on 2/28/89 - 4/4/89 Areas Inspected: Plant operations, status of previous inspection findings, fuel receipt and inspection, generic concerns about steam generator tube plugs, ESF system walkdown, Plant Incident Reports (PIRs), Licensee Event Reports (LERs), maintenance, and surveillance. The inspection involved 112 inspection hours, of which seven were backshift hours, including four deep backshift hours.
Results: No unsafe plant conditions were identified.
Seven items were closed: four unresolved items, one Temporary Instruction (TI), and two previous viola-tions. One unresolved item (UNR 89-03-01) was opened on the need for a direct arming indication for the cold overpressure protection system (COPS).
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t . . . TABLE OF CONTENTS PAGE 1.0 Persons Contacted....................................................
2.0 S umma ry o f Fa c i l i ty Ac t i v i t i e s....................................... 1-
- 3.0 Status of Previous Inspection Findings...............................
3.1 (Closed) UNR 87-24-02, Root Cause and Corrective Actions for Sodium Contamination of the' Reactor Coolant System............
3.2 (Closed) UNR 88-08-01, EQ Calculation Discrepancies for Rockbestos Cables for Containment High Range Radiation Monitors......................................................
3.3 (Closed) UNR 85-54-03 Structural Steel Sampling Program.........
3.4' (Closed) TI 2515/93, " Inspection for Verification of Quality Assurance - Request Regarding Diesel Generator Fuel Oil - Multiplant Action Item A-15", Millstone Units 2 and 3.........
?3.5 ' (Closed) (UNR 88-03-01) COPS Operability with Less than the ' Required Number of Inputs or Loops............................
3.6 (Closed) (NC3 88-03-02) Violation Due to Operation without the Required Overpressure Protection Systems......................
3.7 '(Closed) (NC4 88-03-03) Failure.to Report the Unavailability of Required Overpressure Protection Systems......................
4.0 Review.of Facility Activities........................................
4.1 ' Fuel Receipt and Inspection (60705).............................
4.2 Generic Concerns about Steam Generator Tube Plugs (71707).......
5.0 Plant Operational Status Reviews.....................................
5.1 Review of Plant Incident Reports (PIRs)......................... -11 5.2 ESF System Walkdown (71710).....................................
6.0 Review of Licensee Event Reports ( LERs)..............................
7.0 Maintenance..........................................................
8.0 Surve111ance.........................................................
l 9.0 Management Meetings..................................................
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.__. _ _ _ _ _ _ _ _ __ - . . . r . - . DETAILS i 1.0 Persons Contacted ! Inspection findings were discussed periodically with the below personnel.
j S. Scace, Station Superintendent C. Clement, Unit Superintendent, Unit 3 M. Gentry, Operations Supervisor R. Rothgeb, Maintenance Supervisor K. Burton, Staff Assistant to Unit Superintendent J. Harris, Engineering Supervisor D. McDaniel, Reactor Engineer R. Satchatello, Health Physics Supervisor M. Pearso, Operations Assistant 2.0 Summary of Facility Activities The plant began the inspection period at 96% power.
This slightly below full power operation was because both moisture separator reheaters (MSRs) and the "A" Heater Drain Pump (HDP) were out-of-service.
Power was lowered to 84% at 11:45 a.m., March 1, to effect repairs on a moisture separator drain tank pump.
The pump was repaired and power was returned to 96% by 9:20 p.m.
The "B" MSR drain puop tripped due to a ruptured diaphragm on the "B" MSR Drain Tank emergency dump valve. This necessitated a power reduction to 88% at 1:14 p.m., March 5.
Troubleshooting was completed and power was returned to 96% by 8:56 p.m. that day.
Power level remained at 96% until 9:10 p.m. March 11, when it was lowered to 40% to correct a service water leak per In-Service Test IST 3-88-20.
The leak was repaired. During power escalation, the high level dump valve for the 1A Feedwater Heater (FWH) failed open. The power escalation was stopped and the 1A FWH was isolated.
The power increase resumed.
Full power was achieved at 8:14 a.m., March 13.
Power remained at 100% until 8:55 p.m., March 15 when a seal leak de-veloped on the "A" Turbine Driven Feed Pump (TDFP).
Power was lowered to 95%. The Motor-Driven FP was started, the TDFP was secured, and power was returned to 100%. The "1A" FWH and the "A" HDP were started after repair and installation work was completed.
Plant power remained at 100% through-out the balance of the inspection period.
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- 3.0 Status of Previous Inspection Findings
3.1 -{ Closed) UNR 87-24-02, Root Cause and Corrective Actions for - I Sodium Contamination of the Reactor Coolant System ' i A 7:00 a.m., November 18, 1987 a chemistry sample identified high l Sodium (Na) concentration in the refueling water storage tank (RWST) j and the reactor coolant system.
The plant was in the refueling mode i at the time, with the Reactor Cavity filled with borated water.
Valves between the Chemical Addition Tank (CAT), and RWST were found to be leaking.
The difference in elevation head with the RWST drained initiated the leakage.
Normally, the RWST and CAT are kept at approximately the same levels. All four interconnecting valves were closed and the leakage was stopped.
After the event, RCS and RWST Na concentrations were reduced by the letdown demineralizers and a temporary, skid-mounted ion exchange system. The parallel motor-operated valves (MOVs) (3QSS*MOV29A/B) were lapped and the wedge for an in-line manual valve (3QSS*V29) was machined to reduce seat leakage.
The following procedure changes were made: OP 3305 (Spent Fuel Pit Cooling and Purification) was revised to re-quire isolating the CAT Tank prior to flooding the cavity and restor-ing it to service after the cavity inventory has been pumped back to j sa RWST.
l i OP 3309 (Quench Spray System) has been revised to incorporate new
responses to the CAT Low-Level alarm.
These responses are as fol-j lows: ' I ' a.
If in Mode 6, verify that the CAT manual discharge valves are closed.
b.
If the leakage path cannot be otherwise identified, initiate sampling of the RWST and RCS for sodium content.
i i The inspector concluded that the licensee actions adequately ad-dressed the root cause for the November 18 event.
This item is closed.
Although the licensee actions were comprehensive, some CAT leakage ] still exists.
The sodium hydroxide solution's specific gravity ) generates a larger than expected differential pressure (DP) than a j water-to-water interface. The licensee is considering further design modifications to reduce leakage, i l
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3.2 (Closed) UNR 50-423/88-08-01, EQ Calculation Discrepancies-for Rockbestos Cables for Containment High Range Radiation Monitors The attachment to this report provides a chronological history and explanation of-qualification (EQ) reports related to the environ-mental qualification (EQ) of RSS-_6-104 Coaxial (Coax) Cable construc-tion using Radiation Cross-linked Polyethylene (KXL-100) dielectric.
Sorrento Electronics (SE) issued a 10 CFR 21 report to the licensee regarding the use of RSS-6-104 Coax in their containment high range radiation monitors.
SE indicated that unsatisfactory insulation
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resistance would result if the RSS-6-104 coax was heated above 350 degrees F.
The licensee performed a calculation (3-ENG-094, Revision 1) to determine the maximum expected temperature. The maximum tem-perature reached per the licensee's calculation was 264 degrees F.
The inspector reviewed the calculation assumption validity, use of test data applicable to Millstone 3, use of the same samples in each qualification report, and proper data translation and use in the cal-culation.
Two discrepancies were noted. One was the report used to validate environmental qualification and the second was a failure to inter-polate between two data points.
The licensee has since revised the calculation by calculating the radiation level error with the inter-polated data and has provided an updated calculation to justify operation past the 40 year design life.
This item is closed.
3.3 (Closed) UNR 85-54-03 Structural Steel Sampling Program The steel load verification program for Millstone 3 was undertaken to determine if the estimated loads used for the original design of the structural steel were sufficient to account for the new final recon-ciled supported loads.
During inspection 85-54 on the steel load verification program, four findings resulted in unresolved item 85-54-03.
During.the week of March 13, 1989, a region-based specialist inspector verified that the four findings from the previous inspec-tion in this area were properly addressed by the licensee with the , following details: (1) The first finding concerns attachments which were either not considered in the Teledyne analysis or were recently added to structural steel members. The inspector verified the incorpora-tion of the following structural members shown below in the Teledyne analysis technical report TR-6288-1, Revision 1: The control building (area 4), beam number S20AG4 (shown on page 31 of Teledyne Report) The ESF building (area 2), beam number T7G2G1 (shown on page 20 of Teledyne Report) l
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1 s l The ESF building (area 1), beam number T2G20K (shown on page 15 - ' of Teledyne Report) , t
The ESF building (area 1), beam number T4G30H (shown on page 15 of Teledyne Report) The Teledyne Report shows that these members are within stress design margins.
(2) The second finding was that the assumption made in the analysis with regard to loads were not cicarly stated.
Based on the inspector's review, the report clearly states that the signi-ficant loads were increased by 15% to account for minor loads not incorporated in the overall analysis. This appears on page 9 of Teledyne technical report TR-0288-1.
(3) Teledyne analysis number 6288-3 was undertaken to determine the adequacy of the original design considering current loads and load combinations.
Loads from attached piping, conduit, duct and cable trays were considered. The area under consideration is bounded by columns 54 and 55 and by columns A.4 and C at the control building elevation 47 ft.-6 inches. The inspector verified the following important aspects of the Teledyne an-alysis: Method of Analysis - The GTSTRUDL code was used for the analysis to obtain beam reactions for the check of the connections and to perform the AISC code check.
For the analysis of the 6 main structural beams (W24 x 68), the SRSS method was used as de-scribed in Appendix I of the analysis.
Stresses resulting from the three axis of seismic loads were combined by SRSS with th-support loads and then added absolutely to the live and dead load to produce maximum stresses.
Tables on sheet number A3 and A4 of the Teledyne analysis sum-marized the most highly stressed members with the allowables and factor of safety for normal and shear stresses, respectively.
The inspector found these results acceptable. The inspector also spot checked key design calculations for steel connections.
These calculations demonstrated that the original design con-tinues to be adequate for the latest revised loads imposed upon the structural members.
Therefore, the structural steel analysis for the framing of the control building (area 4) is complete and adequate.
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(4) The steel load verification program was based on a statistical sample. The licensee's acceptance criteria for the sampling plan clearly identified the statistical confidence level of 95/2.5 which means that there is a 95 percent probability that ~1ess'than 2.5 percent of the population will be overstressed.
This is based on a total random sample of 268 beams with no overstress conditions; this is presented on page 40 of the ' Teledyne Report Number TR-6288-1 Revision 1.
' All findings of previous inspections of this area have been resolved.
This item is closed.
3.4 (Closed) TI 2515/93, " Inspection for Verification of Quality ~ Assurance - Request Regarding Diesel Generator Fuel Oil - Multi: plant Action Item A-15," Millstone Units 2 and 3 . The completion of ';I 2515/93 required that the inspector verify that the licensee has included die.wl generator fuel' oil in its quality-assuran::e program.
Revision II to the licensee's Quality Assurance Topical Report (NVQAP), Appendix A, states the following: "The following systems, structures and components of a nuclear power plant, including their foundations and supports, are de-signated as Category I.
The pertinent quality assurance re-quiramants of Appendix B to 10 CFR Part 50, should be applied, as a minimum, to all quality activities affecting the safety function of these systems, structures, and components as listed below and to other itas and services specifically identified by NU in each FSAR addressing Section 3.2.1 of NRC Regulatory Guide 1.70."
Emergency diesel generator fuel oil is listed in Appendix A under the-haading "Consumables." Therefore, the licensee has met tne require-ments of Multiplant Action Item A-15 for Millstone Units 2 and 3 re-garding inclusion of emergency diesel generator fuel oil in their quality assurance program.
This item is closed.
3.5 (Closed) (NC3 88-03-01) COPS Operability with Less than the Required Number of Inputs or Loops On April 12, 1988 the NRC issued a Notice of Violation and Proposed Imposition of Civil Penalty for unavailability of the cold overpres-sure protection system (COPS).
During NRC review of this event, it was found that no procedural guidance existed to specify the use of COPS with less than all the inputs or with isolated loops. An un-resolved item was opened to track resolution of this NRC concern.
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OP 3301 I, Arming the Cold Overpressure Protection System (COPS), Step 6.1 specifies that COPS will not be used with more than one loop isolated.
Steps 7.1 and 7.2 document the procedure to be used for arming COPS for 4 and 3 loop operation, respectively. OPS Form 33011-1, COPS Arming checklist, specifies that a satisfactory channel check must be performed on the Wide Range (WR) T-hot and T-cold and WR RCS Pressure inputs.
Inoperable inputs would make the channel check unsatisfactory. Thus, adequate assurance exists that COPS will be operable for the mentioned conditions.
This item is closed.
3.6 (Closed) (NC3 88-03-07) Violation Oue to Operation without the Required Overpressure Protection Systems On April 12, 1988 the NRC Staff issued a Notice of Violation and Proposed Imposition of Civil Penalty to the Northeast Nuclear Energy Company (NNECO). This action was the result of an inspection con-ducted on January 19-29, 1988 at Millstone Unit 3 to review an event which occurred on January 19, 1988 involving an increase in 'feactor pressure while the res;.or was in cold shutdown.
NRC inspectors reviewed the circumstances associated with a violation of Technical Specification Limiting Condition for Operation 3.4.9.3.
This violation involved a failure to provide proper overpressure protection.
The NRC expressed concern with several deficiencies in control of operations that were exhibited during this event.
Specifically, there was a need for (1) better control of the configuration of equipment, (2) better planning of activities that could affect that control, (3) improved procedures for performing those activities, and (4) improved training of personnel performing those activities.
To enhance the control of operations, the licensee instituted the following changes.
Configuration Control The licensee initiated a review of General Operating Procedures for other cases where specific directions to remove safety-related equip-ment frorn or place safety-related equipment in operation were not governed by a specific system procedure.
Such procedures were re-vised as appropriate.
, Planning of Activities Maintenance dealing with the solid state protection system (SSPS) was reviewed by the licensee to ensure that system interactions were identified and properly described in procedures.
That effort pro-vided Operating and Maintenance personnel specific information on system interrelationships when removing control systems from service.
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Improved Procedures Procedures were reviewed by the licensee to ensure overpressure pro-tection was available when required by Technical Specifications.
Procedures were modified or developed to ensure startup or deener-gization of solid state protection systems were adequate and provided the necessary warnings of system interactions.
Personnel Training The policy was restated that only qualified technicians, trained in solid state protection systems, will work on solid state protection systems. This policy was reemphasized in operator and maintenance training.
These were general activities that the licensee committed to and met.
The following is a list of specific commitments: a.
A procedure for arming COPS (OR3301I) was implemented. This procedure clearly identified the need for SSPS as a required support system and was therefore required by the Technical Specification definition of OPERABLE. No work on SSPS is now permitted while taking credit for the affected train of COPS.
b.
The annunciator response procedure for SSPS TROUBLE, a control board annunciation, now includes the effects on COPS operabil-ity.
c.
The control operators' rounds now monitor SSPS for COPS oper-ability.
d.
The licensee has clarified and expanded its previous policy, so that only qualified technicians may deenergize all or part of vendor-supplied equipment.
e.
The equipment lineup for declaring COPS operable is indepen-dently verified.
f.
The lessons learned have been discussed with licensed operators and instrument technicians.
g.
The Technical Specifications have been reviewed regarding SSPS input to COPS.
The SSPS is a part of COPS and is included by the Technical Specification definition of OPERABLE. Procedures now require SSPS to be completely operable when taking credit for the affected train of COP _ - _ - _. . - _ _ - - . ,. , - . g h.
Specifics of the event and all new operating procedures and changes to existing operating procedures were evaluated for trainir.g impact. Appropriate changes were incorporated in ac-cordance with the training program modification guidance pro-vided in the Nuclear Training Manual. With regard to the In-strumentation and Control Technician SSPS training course, les-sons learned material was developed and added to the course prior to its next delivery.
1.
An SSPS procedure was developed to specifically place SSPS in.
operation. All other complex logic panels were reviewed for required procedure changes.
j.
A steam dump calibration procedure was implemented.
k.
A procedure was developed for response to low temperature over-pressure protection. This procedure builds on the successful operator response to the transient initiating this event.
1.
The General Operating Procedures were reviewed to ensure that all safety-related systems were made operable by a system pro-cedure with appropriate prerequisites.
The inspector reviewed the lesson plans that were written to provide operators and technicians feedback on the event.
The lesson plans were complete and accurate and emphasized the event from the proper perspective. The procedures that were written provided details needed by technicians to establish the prerequisites, provided operators with the details needed to arm COPS, and provided needed direction for anomalous conditions.
This item is closed.
Licensee review of a direct COPS arming indication concluded that the additional safety benefit was minimal. However, based on discussions with the licensee and existing regulatory guidance on the subject, it appears that directly observable indications should be provided to confirm the blocking or resetting of safety-related functions.
The licensee has agreed to reconsider the need for this indication.
This item is unresolved pending further licensee evaluation of the neces-sity for additional COPS indication (UNR 89-03-01).
3.7 (Closed) (NC4 88-03-03) Failure to Report the Unavailability of Required Overpressure Protection Systems NRC review of the cold overpressure event showed that the licensee failed to make the required report. After further consideration, the licensee reclassified the event as reportable under 10 CFR 50.73 (a)(2)(vii). The licensee also stated that the event should have been reported via the ENS per 10 CFR 50.72. The licensee reempha-sized the need for careful review of reporting criteria to their ! _ _ _ - _ _ _ _. _ - _
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staff.
Reporting of this event was reviewed separately as a part of review of the issued Licensee Event Report (88-05). This item is closed.
4.0 Review of Facility Activities 4.1 Fuel Receipt and Inspection (60705) The inspector observed fuel receipt on March 1.
Six shipping con-tainers each containing two fuel assemblies were off-loaded.
The licenree conducted the receipt and inspection in accordance with OP3211A, New Fuel Receipt and Inspection. -The inspector noted a discrepancy while reviewing the fuel handling crane checklist. The calibration expiration date listed in the checklist for the crane (OP Form 3211A-7) was February 15.
The last calibration date listed in PMMS (Production Maintenance Management System) was November 30, 1988 with a scheduled calibration interval of 18 months.
The checklist was corrected.
Preparations for fuel load were continued. A pro-cedure change was submitted to list only the calibration date and calibration frequency.
The fuel transport truck arrived on site at 6:30 a.m., March 1.
Radwaste Health Physics performed contamination and radiation surveys of the shipping containers.
No contamination was found and radiation levels were very close to background.
Security inspected the truck at the Vehicle Access Point and it was admitted to the site.
The inspector observed the licensee recording seal numbers and indepen-dently checked the seal status; none were found tripped. The in-spector observed that the licensee had a licensed SRO (Senior Reactor Operator) supervising the off-loading.
The crane operators used due care and caution while uprighting the fuel. The fuel was lifted and stored in the new fuel storage racks.
Lifting weight limits were observed.
Each assembly was set squarely in its storage location.
No inadequacies were noted.
4.2 Defective Steam Generator (SG) Tube Plug The following inspection was performed at Millstone 2 and certain aspects (procurement, inspection and engineering resolution) may be applicable to Millstone 3 during future tube plugging activities.
The following information is provided as background.
The licensee informed the Millstone 2 inspector on March 21 of a sig-nificant deficiency identified during testing on March 20 in one of four tube plugs installed in the #2 SG (See Inspection Report 50-336/89-05). The licensee initiated a review of his steam generator plugs in response to the February 25 steam generator tube leak at the North Anna facility.
Representatives from the licensee's corporate engineering organization visited North Anna to observe the stress _- - _ __-__-_ ___
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corrosion cracking in the tube plug identified there.. As part of thec evaluation of the Westinghouse mechanical tube plug defects and the-potential impact on Millstone 2 steam generators, the licensee selected 4 plugs installed in Millstone 2 during the 1986-1988 time period for removal and examination.
Based on information from the plug vendor, Westinghouse, the licensee determined that plugs manufactured in the 1984 time period from heat nos. 3962,.3279 and 3513 were susceptible to intergrannual stress corrosion cracking.
The first time suspect plugs could have been used at Millstone 2 was in 1985, but none from the suspect lots were used in 1985.
The licensee selected three plugs installed in 1986 and one installed in 1988 for examination. The four plugs selected were also chosen on the basis of evidence of leakage (dripping or staining on the tube sheet) noted during visual inspections inside the hot leg water box (See Millstone 2 Plant Incident Report 89-27).
The licensee repaired suspect plugs in the Millstone 2 SGs.
NRC re-view of the repair and engineering evaluations is documented in In-spection Report 50-336/89-05.
The use of plugs was reviewed for Millstone 3.
There are a total of 10 welded and 14 mechanical plugs instelled in the four Millstone 3 SGs.
None of these are from the suspect heat numbers.
Six welded plugs were originally installed by the vendor, four welded plugs were installed in December 1984, 10 mechanical plugs were installed in June 1985, and four mechanical plugs were installed in November 1987.
There is no immediate safety concern for defective plugs being in-stalled in the Millstone 3 SGs.
5.0 Plant Operational Status Reviews The inspector reviewed plant operations from the control room and reviewed the operational status of plant safety systems to verify safe operation of the plant in accordance with the technical specifications and plant operating procedures. Actions taken to meet technical specification re-quirements when equipment was inoperable were reviewed to verify the limiting conditions for operations were met.
Plant logs and control room indicators were reviewed to identify changes in plant operational status since the last review and to verify that changes in the status of plant equipment were properly communicated in the logs and records.
Control room instruments were observed for correlation between channels, proper functioning and conformance with technical specifications. Alarm condi-tions in effect were reviewed with control room operators to verify prorar response to off-normal conditions and to verify operators were knowledge-able of plant status. Operators were found to be cognizant of control room indications and plant status.
Control room manning and shift staff-ing were reviewed and compared to technical specification requirements.
No inadequacies were identified. The following specific activities were also addressed.
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5.1 Review of Plant Incident Reports The plant incident reports (PIRs) listed below were reviewed during the inspection period to (i) determine the significance of the events; (ii) review the licensee's evaluation of the events; (iii) verify the licensee's response and corrective actions were proper; and, (iv) verify that the licensee reported the events in l.
accordance with applicable requirements, if required. The PIRs re-viewed were: 35-89 dated 2/24/89, 36-89 dated 2/25/89, 37-89 dated 3/2/89, 38-89 dated 3/4/89, 39-89 dated 3/6/89, 40-89 dated 3/9/89, 41-89 dated 3/12/89, 42-89 dated 3/15/89.
The following items were noted by the inspector: PIR 35-89, Missed Fire Detector Surveillance, dated 2/24/89, docu-mented that only 22 of 28 required fire detectors were being tested per Technical Specification 3.3.3.7 Table 3.3-11.
The root cause and corrective action will be reviewed when the Licensee Event Report is issued.
This is a potentially significant example of a surveillance program inadequacy.
The licensee reviewed all surveillance proce-dures in response to the last SALP but failed to detect this inade-quacy.
Inspector review of the surveillance program will continue.
PIRs 37-89, dated 3/2/89, 43-89, dated 3/16/89, 44-89, dated 3/16/89, and 46-89, dated 3/17/89, all document instances when fire doors were either blocked open or opened without permission. A violation in Inspection Report 50-423/88-23 identified open fire doors as a generic problem.
During the most recent two monthly inspections, eight instances were noted where fire doors were opened without the proper compensation. The circumstances for these eight events and the current four events are slightly different, but the number of these events indicates a lack of understanding or awareness of fire door requirements.
The inspector questioned the adequacy of tha cor-rective actions in light of previous history.
Corrective action ade-quacy will be reviewed in future inspections.
5.2 ESF System Walkdown (71710) The inspector walked down accessible portions of the Turbine Driven Auxiliary Feedwater (TDAFW) System. The purpose of the walkdown was to verify the operational status of the system and to ensure that conditions that existed would not preclude successful operation.
The inspector verified the DWST (Demineralized Water Storage Tank) suction valves (the TS source) were locked open at the tank and in the TDAFW pump room.
Redundant suction valves and pump discharge valves were also properly positioned. Mechanical snubbers for the turbine exhaust were properly aligned and attached to their seismic supports.
Proper oil levels were observed in the lube oil sump with _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - _ _ _ _ _ _ _ _ - - _ _ _ _ -
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l , l-valves being aligned to provide the necessary lubrication on turbine start.
No ignition sources or flammable materials were noted in the area except for certain housekeeping items that will be addressed later.
Calibrations of skid mounted instrumentation was current. Power was available to remotely controlled valves.
The following items were ., l noted as deficient and turned over to the licensee for correction: Housekeeping was good above the turbine grating but poor below -- the grating, with loose insulation, paper, tags, etc. being noted.
A ladder, two chairs, and three pipe wraps were noted in the -- northeast corner. A headset and rope was noted in the southeast corner.
A pipe cap was missing on the pump casing suction vent.
-- -- A chain for a locked closed steam valve was found excessively rusty and was degraded (3 MSS-V914).
-- A " reject" tag was noted on a pressure switch isolation line for TDAFW pump suction pressure.
It was below deck level attached to the line.
In addition, twa " temporary support" stickers were attached to a support structure for a lube oil cooler line.
A steam drain line below deck level, if opened, would blow steam -- directly on safety-related electrical conduit.
-- Two safety-related conduit runs for the Target Rock discharge solenoid valves come within 15 inches of each other on the north wall. The licensee will verify that this is consistent with 10 CFR 50 Appendix R fire protection requirements.
6.0 Review of Licensee Event Reports (LERs) Licensee Event Reports (LERs) submitted during the report period were reviewed to assess LER accuracy, the adequacy of corrective actions, com-pliance with 10 CFR 50.73 reporting requirements, and to determine if there were generic implications or if further information was required.
Selected corrective actions were reviewed for implementation and thorough-ness. The LERs reviewed were: LER 89-003-00, Unidentified Fire Seal due to Inadequate Design Re- -- view. On February 7, 1989 at approximately 1230 hours, with the plant at 100% power (Mode 1), an unidentified temporarily sealed fire penetration was discovered between the "A" and "B" Train Containment Recirculation Spray System (RSS) cubicles located in the Engineered Safety Features (ESF) building.
The unidentified fire penetration ... _ _ _ _ _ _ _ _ _ - - _ _ - - _ - _ _ - _ _ _ _ _ _ _ _ _.
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was declared inoperable,'the associated fire' detectors were verified operable and an hourly fire watch was immediately established in the affected area. The root'cause of this event was personnel error.
The unidentified fire penetration was a 3.25" construction penetra-tion covered by Uni-strut on both sides,_with less than a 1/4" edge visible. The subject penetration was not documented on the applic-able drawings due to inadequate design review.
Based on an inspec-tion of the wall for other camouflaged penetrations with none found, this is considered an isolated event.
No inadequacies were noted.
This licensee-identified item was.evalu-ated as having low safety significance, being acceptably corrected, and not due to a previous corrective action inadequacy (LII 89-03-02).
-- LER 89-004-00, Inoperable Fire Detection due to Operator Error. On February 16,.1989 at approximately 2330 hours, at 0% power, 133 de-grees and 14.7-psia, the Shift Supervisor (SS) observed that fire detection in the 3 foot 8 inch and 24 foot 6 inch elevations of the ' containment Building had been inoperable for approximately 10 hours.
A similar event occurred on February 27, 1989 at 0950 hours, at 100% power, when the SS was informed that fire detection for the "B" Train Emergency Diesel Generator fuel oil vault had been unavailable for-approximately 1.5 hours. The Control Room Operator (CO) had ac-knowledged the alarm indication at.approximately 0800 but did not reset the false alarm signal.
The root cause of_both events was per-sonnel error.. Presence of an indicated fire alarm masked possible.
future alarms resulting in the applicable _ Fire Zone being inoperable.
Miscommunications between Operations Department personnel resulted in the loss of fire detection for the first event. -In the second event, the C0 failed to evaluate an alarm condition.
Immediate corrective action for both events was to establish an hourly fire watch. All on-shift personnel have been briefed to reinforce the importance of prompt evaluation and response to all Fire System alarms. No inade-quacies were noted.
This licensee-identified item was evaluated as l having low safety significance, being acceptably reported and cor-l rected, and not due to a previous corrective action inadequacy (LII 89-03-03).
LER 89-005-00, Inadvertent Safety Injection Due to Personnel Error by -- Switch Mispositioning. On February 17, 1989, at 2023 hours at 0% power, 14.7 psia and 132 degrees, an inadvertent safety injection (SI), with flow to the core,_ occurred on Train B.
The Reactor Cool-ant System had been at reduced inventory for maintenance. After troubleshooting the Solid State Protection System, the Instrument and Control technician recommended to the Supervising Control Operator (SCO) that he " reset the SI block." The SCO placed the switch in the RESET direction, allowing the pressurizer low pressure SI signal to ! , - - - _ - '
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l l be processed. The charging system was restored to normal operation and the vessel level was verified to be stable within 20 minutes of , ! the SI actuation. A containment isolation occurred, initiated by the l SI actuation.
Root cause of the SI was personnel error.
The SCO l placed the switch in RESET instead of the BLOCK position.
One of the
480V load centers, bus 32X, deenergized 2 seconds after the SI actu-ation due to a defective circuit breaker.
The circuit breaker was replaced. As action to prevent recurrence, all operators have been counseled on the importance of being cognizant of their actions when manipulating equipment. The SCO responsible for this error was in-dividually counseled. No inadequacies were noted.
LER 89-006-J, Missed Fire Detector Surveillance on Six. Fire Detec- -- tors due tc Administrative Error, documented on February 24, 1989 at 1445, while in Mode 1, 96% Power,.584 degrees F, 2250 psia. A review performed by Operations Department personnel on Surveillance Proce-dure SP 3641D.3 revealed a discrepancy between the procedure and Technical Specification Table 3.3-11 requirements.
The surveillance procedure form, SP 36410.3-3, listed 22 smoke detectors.
Upon in-vestigation, it was found that the Surveillance Procedure was in error and did not identify six (6) additional smoke detectors which were located within the Main Control Board pr.nels. The 6 detectors were installed via a late construction design change prior to initial startup. The Technical Specification requirement of verifying system operability every six months had not been performed for these detec-tors. No immediate actions were required of plant operators, due to the control room being continuously manned.
The root cause of the event was administrative error. The surveillance procedure had not been updated to comply with the final system design and the Technical Specifications.
The Surveillance procedure has been revised to in-corporate the 6 missing detectors and detector testing was satis-factor 11y completed on February 28, 1989. A comprehensive review has been performed on the Fire Detection and Control System and no other reportable discrepancies were noted.
No inadequacies were noted.
This licensee-identified item was evaluated as having low safety sig-nificance, being acceptably corrected, and not due to a previous cor-rective action inadequacy (LII 89-03-04).
7.0 Maintenance The inspector observed and reviewed selected portions of preventive and corrective maintenance to verify compliance with regulations, use of ad-ministrative and maintenance procedures, compliance with codes and stand-ards, proper QA/QC involvement, use of bypass jumpers and safety tags, personnel protection, and equipment alignment and retest. The following activities were included: _ _ _ _ - - - _ _ _ - _ _ _ _ _
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l -- Reactor Plant Component Cooling Pump Work, dated 3/5/89 -- Diesel Generator Air Compressor Repair, dated 3/20/89 -- RWST to Safety Injection Pump Valve Work, dated 3/29/89 No inadequacies were identified.
1 8.0 Surveillance l l The inspector observed portions of surveillance tests to assess perform- ! ance 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: -- New Fuel Handling Crane Limit Switches, dated 2/28/89 -- SI Accumulator Isolation Electrical Verification, dated 3/5/89 -- RCS Leakage to RCP Seals, dated 3/28/89 No inadequacies were noted.
. 9.0 Management Meetings 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.
On March 16, 1989, at a licensee-requested management meeting in NRC Region I, Northeast Utilities made a presentation to the NRC on the per-formance of Millstone 1 and Millstone 3.
The presentation materials are appended to Millstone 1 Inspection Report 50-245/89-04.
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-- . _ _ _ _ _ _ _ -.- , - . _ -. NRC REPORT 50-423/89-03 ATTACHMENT A ROCKBESTOS C0AX CABLE QUALIFICATION AND ISOLATION RESISTANCE VALUES GA/Sorrento HRRM for MP3 1.
Rockbestos Test Report QR-2806 (4/23/82) is the original qualification test for RSS-6-104 Coaxial Cable using preliminary aging data points for determining the thermal aging parameters for the test plan.
Insulation resistance.(IR) data in this test is not representative of the actual co-axial cable IR because of the influence of test chamber extension leads and terminations on the recorded values. The primary dielectric material for these cables is identified as Radiation Cross-linked Modified Poly-olefin (KXL-100).
2.
Supplement 3 to QR-2806-(12/7/82) provides additional data and information regarding actual tnermal aging results. These values demonstrate that the cables tested in QR-2806 actually had been aged far in excess of 40 years.
3.
Sorrento/Rockbestos Test Report QR-6810 (2/21/86) provides a report on actual IR versus temperature values for RSS-6-104 Coaxial Cable (KXL-100 dielectric). The data is from a test performed for engineering informa-I tion only, and was done based on the premise that IR degradation and re-covery is a function of temperature only.
This premise had been verified for Polyolefin and other wire insulation materials by internal Rockbestos testing during the 1982-84 period.
4.
Rockbestos Test Report QR-6802 (3/12/86) provides a retest qualification
of the RSS-6-104 Coaxial Cable construction using Radiation Cross-linked Polyethylene (KXL-100) dielectric.
(Note: Polyethylene is a specific type of the Polyolefin class of material.) This test demonstrates qualifica-tion and reinforces the conclusions of the previous QR-2806 report. Ther-mal aging in this test was performed based on new Arrehenius data points experimentally obtained in 1985.
IR data contained in this report again , does not reflect actual coax cable values because extension leads and terminations were used in the test chamber.
5.
Rockbestos Test Report QR-7804 (1/27/88) is a recent test which most clearly demonstrates that IR degradation and recovery of Radiation Cross-linked Polyethylene insulation is a function of temperature only.
The ) i dielectric tested in this case was Rockbestos KXL-760G, which uses the ! same base material as the KXL-100 formulation used in the Coaxial Cable ' construction.
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