IR 05000245/1985022
| ML20135G503 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 09/05/1985 |
| From: | Lipinski D, Mccabe E, Shedlosky J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20135G494 | List: |
| References | |
| TASK-***, TASK-TM 50-245-85-22, 50-336-85-27, IEB-79-06, IEB-79-06B, IEB-79-08, IEB-79-6, IEB-79-6B, IEB-79-8, NUDOCS 8509190318 | |
| Download: ML20135G503 (14) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report:
50-245/85-22; 50-336/85-27 Docket Nos:
50-245/50-336 License Nos. DPR-21; DPR-65
Licensee:
Northeast Nuclear Energy Company Facility:
Millstone Nuclear Power Station, Waterford, Connecticut i
Inspection at: Millstone Units 1 & 2 Dates:
July 23, 1985 through September 3, 1985 Inspectors:
00<. b %, b q lfl 95
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J John T. Shedlosky, Senior Resident Inspector Date FL M, 4 sir /rr David R. Lipinski, Resident Inspector Date Approved:
_ & kh 9 /5'/ P T E. C. McCabe, Chief, Reactor Projects Section 3B Date
Summary:
Routine NRC resident (130 hours0.0015 days <br />0.0361 hours <br />2.149471e-4 weeks <br />4.9465e-5 months <br />) inspection of plant operations, equip-ment alignment and readiness, radiation protection, physical security, fire pro-tection, design changes, surveillance.
The circumstances surrounding an automatic reactor scram which occurred at Unit 1 on August 13 were included along with the follow-up inspection of items identified during previous NRC inspections and in response to the TMI Action Plan. No violations or unacceptable conditions were identified.
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DETAILS 1.
Plant Status Unit 1 The reactor operated at full power, except for an automatic reactor scram on August 13, and had load reductions on August 9 and 10 due to off-site trans-mission line difficulties. The reactor scram, which was initiated by main steam line area high radiation, occurred after air was introduced into the feedwater system through a condensate demineralizer.
The reactor was again made critical at 2122 on August 13.
The turbine generator had been on line since 0944, August 4, 1984 (374 days) prior to that scram.
Unit 2 The reactor operated at full power except for load reductions in response to plant and off-site equipment problems.
Service water pump discharge strainers plugged during a storm on July 26.
A plant shutdown was initiated pursuant to Technical Specification LC0 action statements.
The service water system was restored the same day and the plant returned to full power on August 27.
The off-site transmission line problems on August 9 and 10 caused load reduc-tions to be made at Unit 2.
Power was also reduced in response to a dropped control rod on August 12.
2.
Reactor Scram - Unit 1 Millstone Unit 1 sustained a scram and main steam line isolation at 1207 on August 13.
The scram and isolation occurred as a result of main steam line area high radiation which has been attributed to a brief injection of air into the feedwater stream as a condensate demineralizer was placed into service without being fully filled.
Following the scram and isolation, operators manually initiated the isolation condenser to remove decay heat.
By 1217, after radiation levels were confirmed to be normal, the main condenser had been restored to service as the heat sink and the isolation condenser had been returned to standby readiness.
The resident inspector observed the post-scram transient from the control room and noted no unexpected pressures, levels, or temperatures.
No emergency core cooling system or safety-relief valve challenges occurred.
The operators did not attribute the trip to an air in-jection until the resident inspector made that correlation.
That did not af-fect the trip response, which was proper.
Normal licensee post-trip review was evaluated as sufficient to have identified the trip cause before restart.
Samples of reactor coolant taken shortly after the trip showed no evidence of significant fuel element failures.
All units below are micro-Curies per milli-litre (uCi/ml):
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Date/ Time Gross Activity Iodine 131 Iodine 134 (uCi/ml)
(uCi/ml)
(uCi/ml)
8-12/0800 0.081 0.00012 0.0157
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8-13/1315 0.154 not detected *
0.0010 8-14/1520 0.047 0.00012 0.0089 8-19/0745 0.084 0.00011 0.0167
- Peak overwhelmed by adjacent short-lived peak.
Process radiation monitor responses were evaluated by the inspector.
No evi-dence of unusual off-site release was observed.
Units below are milli-Roent-gens per hour (mR/hr) or micro-Curies per second (uCi/sec) or counts per second (C/sec).
Steam Line Condenser Off Gas Effluent Stock Area (mR/hr)
(mR/hr)
(uCi/sec) (C/sec) (uC1/sec)
Normal Full Power 300
8640
20 (Pre-scram)
Peak at/after scram 1800
10800
40 Post-Scram
10-2160
20 Background
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The effluent stack serves as a common elevated release point for all three Millstone Units.
Unit 1 condenser off gases are passed through large acti-vated charcoal adsorption beds prior to entry into the stack.
The inspector reviewed the operability and testing of these radiation monitors as discussed in paragraph 7.
The phenomenon of elevated main steam line radiation following inadvertent air addition has been observed at Millstone and at other Boiling Water Reac-tors. Air addition results in a brief rise in the number of Oxygen-16 target nuclei. A neutron bombardment reaction yields a proton and short-lived Ni-trogen-16.
In this specific event, feedwater and reactor coolant conductivity showed no perturbation and isotopic analysis of reactor coolant samples showed no unusual nuclides.
On this basis, the introduction of foreign material from the recently ultrasonically cleaned demineralizer bed is not indicated to be a potential contributor to the event.
The unit was taken critical at 2122 on August 13 and was synchronized to the grid at 2345 on August 13.
No unacceptable conditions or practices were noted.
3.
Dropped Control Element Assembly (CEA) - Unit 2 At 1415 on August 12, a Control Element Assembly (CEA) dropped from its full-out position of 180 steps to 143 steps.
Routine monthly surveillance test SP2620A "CEA Partial Movement" and SP2620B "CEA Group Deviation" were in pro-gress. The former test demonstrates the freedom of motion of CEA's by in-serting each CEA 10 steps then withdrawing the CEA to its previous position.
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The latter test demonstrates that CEA motion inhibit circuitry stops CEA motion when a CEA position deviates by 10 steps from the position of its group.
Several unsuccessful attempts to move CEA 67 were made before the CEA suddenly dropped from step 180 to step 143.
Troubleshooting revealed no inadequacy of the reactor manual control system nor of the CEA position in-dication systems.
Power was reduced to 69% of full power by 1510 and by 1530, CEA 67 had been realigned with its group (Group 3) at 181 steps.
The unit returned to full power at 0655 on August 13.
The inspector observed that the recovery was within the guidelines specified by Technical Specification Limiting Condition for Operation 3.1.3.1.E.
No unacceptable conditions or practices were observed.
4.
Unplanned Rapid Load Reductions - Units 1 and 2 On August 9 and 10, both Millstone 1 and 2 conducted rapid load reductions due to transmission line difficulties off-site.
Two of three 345,000 volt transmission lines linking the Millstone switchyard with the balance of the grid were lost at 1510 on August 9.
One line (#383) was lost due to tree impingement while the other line (#371) was lost due to a spurious overcurrent trip.
Rapid load reductions were accomplished on both units with no equipment failures or safety system actions.
By 1610 on August 9 all three 345,000 volt transmission lines had been restored to service and both units began increas-ing power.
Unit 1 reached full power at 2010 on August 9 and Unit 2 reached full power at 1437 on August 10.
At 1622 on August 10, a transmission line (#383) was opened for tree removal.
A second line (#348) also tripped.
By 1624, all three lines had been restored.
Both units were returned to full power by 1630 on August 10.
The inspector verified that Technical Specifica-tion Limiting Conditions for Operation 3.9.B.1 and 3.8.1.1 for units 1 and 2, respectively, were not violated.
No unacceptable conditions or practices were observed.
5.
Unusual Electric Power Demand August has been a period of.high electrical demand.
At 1500 on August 16, 1985, the Connecticut Valley Electrical Exchange (CONVEX) experienced its all-time high electric power demand of 5575 Megawatts.
CONVEX controls power in most of Connecticut and Western Massachusetts.
The Northeast Utilities portion of that load demand, approximately 78% or 4340 Megawatts, represented a new load peak for the utility.
No link between the high power demand and the offsite power losses was established.
6.
Three Mile Island (TMI) Action Plan Items The status of selected outstanding issues from the Three Mile Island Action Plan is discussed below.
For some items, hardware modifications remain to be made for other additional inspection effort may be required while others are resolved.
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I.A.I.3.2a Shift Manning Units 1 and 2 (Closed)
Both Millstone units have incorporated additional shift staffing re-quirements into their respective unit Technical Specifications, Section 6.
These include 2 Senior Reactor Operators, 2 Reactor Operators, 2 equipment operators (non-licensed) and a Shift Technical Advisor.
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l time restrictions of Generic Letter 82-21 have been incorporated into Administrative Control Procedure ACP 1.19 " Overtime Controls for Per-sonnel Working at the Operating Stations" which is applicable to both units.
Under this procedure, personnel may not be required to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> consecutively, or more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> of any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, or more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in 7 days, or more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period, and must have at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> rest between shifts.
The licensee
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has made provisions for waiver by senior supervisory personnel in extra-ordinary circumstances. This item is closed for both Millstone l'and 2.
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II.K.3.28 Qualification of ADS Accumulators
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Unit 1 (Closed)
The qualification of Automatic Depressurization System (ADS) accumulators was accepted by the NRC Office of Nuclear Reactor Regulation and is documented in a letter from J. A. Zwolinski to W.G. Counsil dated November 29, 1984 (serial LS05-84-11-033).
The basis of the acceptance includes the availability of a secondary supply system and surveillance testing of nitrogen supply integrity.
The inspector verified by physical in-spection the existence of the secondary nitrogen system as described in the above letter with the ability to charge nitrogen bottles from within the Turbine Building.
The inspector noted that a procedure is in place to place that system into service (Section 7.1 of OP311D Rev. 1 " Placing the Secondary N-2 Gas Supply System in Service"). The inspector reviewed
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the surveillance test procedure and confirmed that acceptance criteria are those specified in the above letter and that the restoration from the test includes leak checks of mechanical joints opened for testing.
Upon these bases, this item is closed for Millstone Unit 1.
Unit 2 (Closed)
This ite< is.only applicable to Boiling Water Reactors.
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II.K.3.18 Modification of ADS Logic j
Unit 1 (Closed)
The modification to Automatic Depressurization System (ADS) actuation
logic was considered for all Boiling Water Reactors to provide redundant protection against certain categories of small break loss of coolant ac-
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cidents.
It has been determined that these modifications provide addi-tional protection only for plants not having Isolation Condensers.
For plants using Isolation Condensers, these modifications were found to be unnecessary and the installation requirements were withdrawn.
The above summary is documented in a letter from J.A. Zwolinski to W.G. Counsil dated January 8,1985 (serial LS05-85-01-006).
Since Millstone Unit 1 incorporates an isolation condenser and the modifications were not made, this item is closed for Millstone Unit 1.
Unit 2 (Closed)
This item is only applicable to BWR plants.
d.
II.K.3.24 Space Cooling for HPCI/RCIC Unit 1 (Closed)
This item relates to High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) systems.
Millstone Unit 1 uses neither system.
Rather, a portion of the feedwater train takes the place of HPCI and the isolation condenser takes the place of the RCIC system.
Neither system operates in an enclosed space or uses steam turbines.
This item, relating to steam turbine space cooling, is closed for Millstone Unit 1.
Unit 2 (Closed)
This item is unique to BWR plants.
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II.K.1.2.
IE Bulletin 79-06/79-06B Unit 1 (Closed)
This Bulletin is unique to PWR plants.
Unit 2 (Closed)
Bulletins 79-06/79-068 related to certain immediate measures to be taken by owners of non-B&W Pressurized Water Reactors.
This Bulletin was re-viewed in Inspection Report 50-336/79-13 and is closed.
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II.K.1.4 IE Bulletin 79-08 Unit 1 (Closed)
Bulletin 79-08 related to certain immediate measures to be taken by owners of Boiling Water Reactors.
This Bulletin was reviewed in Inspec-tion Report 50-245/80-17 and is closed.
Unit 2 (Closed)
This Bulletin is unique to BWR plants.
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I.D.2 Plant Safety-Parameter Display Censole Units 1 and 2 (0 pen)
Plant Safety-Parameter Display Consoles are not yet installed at Mill-stone 1 and 2.
This is the subject of recent Orders issued by the NRC to both units.
This item will be inspected during system installation and testing.
This item is open for both units.
h.
II.K.3.1 Automatic PORV Isolation Unit 1 (Closed)
This item is unique to PWRs.
Unit 2 (Closed)
This item required various actions and studies to reduce the likelihood of a small break loss of coolant accident due to a stuck-open Power Operated Relief Valve (PORV).
A system to automatically sense such a condition and isolate the PORVs was initially envisioned.
Upon review of information and analyses submitted by the Combustion Engineering Group, the NRC Office of Nuclear Reactor Regulation found that the re-quirements of this item are met with existing PORV, safety valve, and reactor high pressure trip settings.
This decision is documented in a letter from J.R. Miller to W.G. Counsil dated September 23, 1983 (no serial number).
This letter includes the NRC Safety Analysis and the NRC contractor's Technical Evaluation Report.
The inspector observed that the licensee has not altered these key settings. This item is, therefore, closed for Millstone Unit 2.
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II.K.3.16 Challenges and Failures to Relief Valves Unit 1 (Closed)
This item required owners of Boiling Water Reactors to assess various actions and modifications to reduce the challenges and failures of relief valves.
Four recommendations were made relating to cycling of valves, reactor isolation on low water level, relief valve " simmer," and relief valve testing.
The licensee has incorporated a procedural requirement (EOP571 " Reactor Pressure Vessel Pressure Control") to manually blowdown reactor pressure to 900 psig should any relief valve begin cycling.
The low level isolation was found to already be at the recommended value (Technical Specification 3.2.1.A).
Relief valve " simmer" margins were found to be near maximum.
Relief valve testing requirements of the NRC staff's proposals were incorporated in Technical Specifications (3/4.6.E).
The NRC Office of Nuclear Reactor Regulation accepted these actions as completing the requirements of this item, as documented in a letter from
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W.A. Paulson to W.G. Counsil dated October 18, 1984 (serial LS05-84-10-019).
Tlie inspector verified that the procedural controls and Technical Specifications cited are actually in place.
This item is closed.
Unit 2 (Closed)
This item is unique to BWR plants.
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II.B.1 Reactor Coolant Systre Vents Unit 1 (Closed)
Requirements for the installatior of Reactor Coolant System (RCS) vents are based upon postulated post-Loss of Coolant Accidents (LOCA) condi-tions in which accumulated non-condensible gases could interfere with core cooling. Millstone Unit 1 is a Boiling Water Reactor and incor-porates remotely operated safety-relief valves near the high point of the RCS.
The inspector noted that advice to operators in the use of venting and blowdown using these valves is incorporated into several operating and emergency procedures.
Valve operability and inspection requirements are addressed in Technical Specifications.
This existing system has been found to be acceptable as an RCS vent and no separate vent system for the Isolation Condenser has been mandated by the NRC Office of Nuclear Reactor Regulation.
These findings are documented in a letter from D.M. Crutchfield to W.G. Counsil dated June 27, 1983 (serial LS05-83-06-057). The NRC Safety Evaluation Report and the con-tractors Technical Evaluation Report are enclosures to that letter.
This item is closed for Unit 1.
Unit 2 (0 pen)
Requirements for the installation of RCS high point vents are based upon postulated post-LOCA conditions in which accumulated gases could inter-fere with core cooling capacity.
A system of remotely operated vents was installed in the reactor vessel head vent piping and the pressurizer head vent piping during the 1980 refueling outage. Work in progress in those installations was observed by the Senior Resident Inspector and documented in inspection report 50-336/80-19.
The status of implementa-tion was reviewed again during early 1984.
The inspector found design specificat*ons and implementation to be acceptable.
System as-built drawings and Piping & Instrument Diagrams (P& ids) were found to have been updated to reflect the modification.
Operating procedures were found to be available and deemed acceptable.
Several items remain open regard-ing this system.
These include:
84-07-01 concerning control circuit vulnerability, 84-07-02 concerning valve position indication, 84-07-03 concerning In-Service Testing, and 85-11-01 concerning seismic supports for vent piping.
The RCS vent requirements for Unit 2 remain open pend-ing resolution of the above listed concerns.
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II.B.2 Plant Shielding Unit 1 (Closed)
This item was addressed in detail in Inspection Report 50-245/83-07.
That report, conducted by region-based specialist inspectors, was devoted entirely to the plant shielding aspects of TMI item II.B.2.
Report 50-245/83-07 closed this item with no issues remaining open or unresolved.
This item is closed for Unit 1.
Unit 2 (Closed)
This item was addressed in detail in Inspection Report 50-336/83-09.
That report, conducted by region-based specialist inspectors, was devoted entirely to plant shielding aspects of TMI item II.B.2.
This item was left open pending the installation of reach rods on four valves to permit
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manual operation from a low radiation area.
These valves are 2-51-400,
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2-SI-709, 2-CH-340, and 2-CH-440.
The inspector verified that the reach rods have been installed.
Record review confirmed that the work was completed on January 6, 1984 in accordance with Plant Design Change Re-quest (PDCR) 2-132-82 with management and quality assurance acceptances completed on February 16, 1984.
PDCR 2-132-82 is the design document reviewed and accepted by the inspectors in Inspection Report 83-09.
This item is now closed.
1.
Several items require licensee effort to prepare them for NRC review.
These are:
Item Topic Unit (s)
II.B.3 Post-Accident Sampling 1, 2 II.F.1 Accident Monitoring 1, 2 II.F.2 Instrumentation for Detection of Adequate
Core Cooling II.K.3.5 Auto Trip of RCPs
II.K.3.21 Auto Restart of LPCI/ CSS
II.K.3.25 Power on Pump Seals 1, 2 II.K.3.57 Manual Actuation of ADS
III.A.1.2 Emergency Support Facilities 1, 2 III.A.2 Emergency Preparedness-1, 2 III.D.3.4 Control Room Habitability 1, 2 These items will be addressed in a future report.
7.
Monthly Observation of Surveillance and Maintenance Unit 1
" Air Ejector Off-Gas Isolation Radiation Monitor Functional Test" per
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SP406E Revision 3.
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This surveillance is conducted to confirm that the main condenser off-gas stream will be isolated in the event of a release in excess of the limits of Environmental Technical Specification (Appendix B) 2.4.2.2K or in the event of monitor failure.
Correlation between radiation meters and the strip-chart recorder is also observed.
Channel 2 of the instru-
ment passed the surveillance test; Channel 1 failed the test, reading higher than allowable (the conservative direction).
The inspector ob-served that the failure was documented in accordance with the test pro-cedure and that corrective actions were initiated.
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Corrective Maintenance to Air Ejector Radiation Monitor Channel 1, per Automated Work Order M1-85-07104.
This activity was conducted to restore the affected radiation monitor
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to full operability following its failure of a routine functional test.
Repair and recalibration were accomplished in accordance with "Off Gas Radiation Monitor Drawer Calibration," SP406F. Revision 5.
The resistance box, (GE) variable voltage source (Transmation Snap Pack 1040), and digital voltmeter (Fluke 8050A) were all accounted in the licensee's i
Quality Assurance metrology program.
The instrument was successfully restored to service. There is no history of previous test failures for j
this instrument, which failed the surveillance initially because of a
shift in the zero point for an unknown reason.
No unacceptable condi-tions were observed.
" Turbine Stop Valve Testing" per OP314 Revision 12 " Turbine Generator"
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section 7.4.
This test confirms the freedom of motion of each turbine stop valve as well as confirming the actuation of Reactor Protective System relaying when a stop valve is closed.
Power was reduced to approximately 80% of full power to permit testing.
The inspector noted that the Reactor En-gineer was in the control room to monitor the power excursion.
No un-
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acceptable conditions were observed.
" Service Water Pipe Repairs"
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A leak which occurred in a weld of a service water pipe received a tem-parary repair on July 28.
The leak is located between a 36" X 30" re-ducer and a welded flange at the 30" end.
The repair made use of t
machined split rings and welded plate to encapsulate the entire reducer.
It was assembled mechanically and then welded in place.
The reducer,
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t which is located on the discharge side of the service water strainer, is scheduled for replacement during the October 1985 refueling / main-tenance outage.
This repair was documented and controlled as Job Order M18506185 and NCR 185-64.
A second leak in the discharge from the'"A" Service Water pump was re-paired on August 5.
The leak was down-stream from the pump discharge isolation valves. This leak was repaired with a clamp which was slipped down over the leaking area.
Pump discharge piping is located in the
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space above water level in the circulating water intake bays and is, therefore, subject to that environment.
This repair was documented and controlled by Job Order M18506485 and NCR 185-68.
The inspector examined the licensee's program for the inspection and re-placement of service water piping during the October 1985 refueling /
maintenance outage.
The licensee has identified all components known to be degraded.
Their replacement is planned pending material availa-bility.
The licensee has maintained an inspection program for volumetric examination of service water and emergency service water piping.
These will continue during the 1985 outage.
The inspector did find an extensive set of maintenance procedures, MP2721Q, Q1, Q2, and Q4, which addressed the application of various protective coatings, including those for internal coating of carbon steel pipe in sea water service.
The licensee has evaluated the service water system as operable, not-withstanding the leakage / degradation problem.
Leaks identified have been the result of pitting-type corrosion which has produced small holes.
Those holes have not materially changed system flow or capacity.
No general wall thinning or general corrosion has been identified by licen-
-see checks, and the pitting-type corrosion appears to be associated with flaws in the protective coating on the inside of the pipe.
Senior Resi-dent Inspector review and consultation with a Region I specialist iden-tified no concern about continued system operability.
Routine outage inspection will address the service water pipe work.
At present, the inspector has no further questions on this matter.
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ECCS Component Control Relay Terminal Connection.
An engineer found the leads disconnected from a GE CR120K20 relay located in a control room cabinet on July 30.
The relay is associated with the LPCI/ Containment Cooling heat exchanger bypass valve 1-LP-7A and is used to automatically hold the valve fully open for 60 seconds following an ECCS loss.of coolant accident actuation. This design feature is de-scribed in FSAR Section 6.2.4.2(5)(d), but valve position does not in-hibit LPCI flow.
The leads, which attach to " spade lugs" or " appliance terminals" of the relay, were reconnected.
At Millstone, only CR120 relays have these push on, friction fit connections.
Similar CR120 re-lays were inspected by the licensee and no other deficiencies were found.
The licensee four.d the disconnected wire while inspecting the CR120 re-lays which are scheduled for replacement during the 1985 refueling / main-tenance outage.
The inspector examined this and other CR120 relays, and found no additional terminal problems.
No work which caused removal of the leads in this case was identified by the licensee.
This appears to be an isolated occurrence which had no material potential to affect safety system ability to perform its design function.
The inspector had no further questions on this ite '
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Unit 2 Cleaning of the "A" and "C" Service Water Pump discharge strainers on
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July 26.
Inspection of "B" Containment Spray Pump on August 1.
This revealed that
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a piece of foreign material had passed between the pump wear rings. This had caused some galling of the rings, which were replaced.
8.
Engineered Safety Features (ESP) System Walk-Down - Unit 1 In addition to sampling inspection of safety-related systems and components during routine daily facility tours, the inspector conducted a detailed walk-down of the Isolation Condenser system. This walk-down included the condition of piping and valves in the steam supply and condensate return lines outside the drywell, makeup water lines from the fire fighting water and condensate transfer systems, and line break detection instrumentation. The inspector closely examined the grouting and embedment of piping supports for evidence of damage due to earlier water hammer events.
No damage was observed.
The inspector returned to examine the system following its restoration to standby readiness following its use on August 13.
No unacceptable conditions or practices were observed.
9.
Observation of Security Response Force Drill
On August 8, the inspector observed a Unit 1 Security Response Force training drill.
The drill was monitored and graded using a proceduralized check list.
Attributes in grading included strategy of the team leader, area search tech-niques, weapons handling, personnel apprehension and search, and general con-duct.
Licensee monitors rated the drill as successful.
The inspector found the drill to be realistic, well executed, and well graded. The inspector provided the licensee with no advance warning that he would observe such a drill.
10.
Loss of Service Water Header - Unit 2 Two of three service water pump discharge strainers clogged with debris during a storm on July 26.
Thisresultedinalossofoneoftwoplantservicewatqq supply headers. Unit 1 was not affected At 0538 the "B" service water pump was started to supply the "B" header, the
"C" pump was secured because of high differential pressure in its outlet strainer. The A pump continued to supply the A header.
Since there are three pumps available to supply two headers, plant operation was in compliance with the Technical Specifications. At 0830, one of two containment spray pumps was taken out of service f or a preventive maintenance inspection. This caused the plant to enter Technical Specifications Action Statement 3.6.2.1.a. which would allow operation for 30 days.
At 1140 the "A" Service Water pump strainer clogged and the pump was secured. The 8 pump continued to supply the B header with its strainer in a continuous blowdown mode.
The loss of the A service
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water header resulted in a loss of cooling to the A Emergency Diesel Generator and the "A" Reactor Building Closed Cooling Water (RBCCW) system.
Because of this, the plant entered action statements 3.6.2.1.b, 3.7.4.1, and 3.8.1.1.a.
The two service water headers were cross connected to supply cooling to both RBCCW systems at the RBCCW heat exchanger service water inlet and a plant shutdown was commenced in accordance with Technical Specification 3.0.3 at 1155. An Unusual Event was declared at that time and the NRC Operations Cen-
.ter was notified by the licensee using the ENS.
As a result of higher RBCCW temperature, containment temperature increased and exceeded 120 degrees F at 1325.
The plant then entered action statement 3.6.1.5.
Temperature was reduced to less than 120 degrees F at 1357 after flow was increased to the containment coolers and air recirculation fans were placed in fast speed.
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The requirement for a plant shutdown was removed following the successful
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completion of an operability surveillance test of the
"B" Containment Spray pump at 1558.
The Unusual Event was terminated at that time.
Reactor power
had been reduced to 43 percent.
The "A" service water header was restored at 1823 when the "A" pump was started after having the strainer manually cleaned. The "C" pump was started on the "B" header at 1825.
This allowed the "B" pump strainer to be disassembled for cleaning.
The reactor was re-i stored to full power at 1840, July 27.
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The inspectors observed the licensee's handling of this occurrence from the
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aspect of Technical Specification compliance, plant parameters and equipment availability, accuracy of reports made to the NRC Operations Center and the involvement of management personnel. There was generally good response and positive management involvement to the changing plant conditions caused by strainer plugging.
The strainer design is such that the supply of water for back washing is pro-
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i vided by the filtered water.
A severely clogged strainer results in a high differential pressure and insufficient pressure for backwashing.
The licensee j
normally operates the strainer in an automatic backwash cycle which is initi-
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ated by differential pressure.
Backwashing may be manually initiated at times of heavy fouling. This occurred on July 26.
However, at that time an unusu-ally heavy buildup of crushed shells and grass exceeded the backwash capacity of the strainers.
Although this was the first occasion in which a service water header was lost due to debris clogging a pump strainer, the licensee has initiated an engi-
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neering assignment to provide system modifications for higher reliability.
- The installation of strainer bypass valves is being considered.
There were no unacceptable conditions identified by the inspectors.
Since the licensee has initiated an improvement program and there is routine in-spection of this facet of facility operation, there is no open item resulting
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from this problem.
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4 11.
Review of Previously Inspected Items Items Common to Both Unit 1 and Unit 2
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50-245/84-15-03 & 50-336/84-15-03 (Closed) This unresolved item was pursued to clarify the conduct of " Fire Protection and Loss Prevention Audits." Cur-rent Technical Specifications require that such audits be conducted by "an outside firm experienced in fire protection and loss prevention." This type of audit is typically conducted at industrial facilities by insurance com-panies. The licensee's corporate staff (NUSCO) has been conducting the audits of the Millstane plants (NNECO).
The licensee had maintained that the cor-
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j porate staff is an "outside firm" with respect to the plant staffs.
In a letter dated August 12, 1985, the licensee proposed a revision to Technical Specifications clarifying the audit personnel requirements.
Resolution is in progress by the Office of Nuclear Reactor Regulation; therefore, this item is closed.
50-245/82-10-04 & 50-336/82-14-04 (Closed) This unresolved item was pursued to clarify long term record storage environmental requirements.
The licensee had.not been able to maintain the records vault within acceptable temperature and humidity limits so as to assure that radiographic, photographic, and paper records would be useable for the life of the plant.
The licensee has in-stalled a vault air conditioning system.
Testing of that system was completed on August 23 and the system is in operation.
This issue is closed.
- 50-245/25-15-67 & 50-336/25-15-67 (Closed) The surveys mandated by this Temporary Instruction (TI) were completed and documented in Inspection Report 50-245/85-19 & 50-336/85-25 paragraph 8.
This item is closed.
l 50-245/83-19-01 & 50-336/83-27-01 (Closed) This item was opened to resolve responsibility for the calibration and maintenance of storage ovens for low-hydrogen welding electrodes.
Such electrodes must be stored at elevated tem-peratures, as specified by their manufacturers, to prevent moisture (hydrogen)
pickup which could adversely effect weld quality.
The licensee has assigned responsibility for this activity to the Unit 2 Instrumentation and Control Department.
Procedure IC24290 was reviewed by the inspector and found to address the five ovens. These items are closed.
t Unit 2 Items
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50-336/83-09-01 (Closed) This item was opened to track the completion of
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Plant Design Change Request 2-132-82 which involved installation of reach rods on four manual valves to permit operation from an area of low radiation fol-lowing a postulated accident. The inspector verified completion of this de-sign change.
This item is closed.
12.
Exit Interview At periodic intervals during the inspection, meetings were held between lic-
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ensee site management concerning the inspection scope and findings.
No pro-prietary information was identified as being in the report findings.
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