IR 05000373/1982052
| ML20028F471 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 01/18/1983 |
| From: | Richard Anderson, Guldemond W, Madison A, Raglin K, Walker R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20028F448 | List: |
| References | |
| 50-373-82-52, 50-374-82-20, IEB-82-03, IEB-82-3, NUDOCS 8302010377 | |
| Download: ML20028F471 (13) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Report No. 50-373/82-52(DPRP); 50-374/82-20(DPRP)
Docket No. 50-373; 50-374 License No. NPF-11, CPPR-100 Licensee: Commonwealth Edison Company Post Office Box 767 Chicago, IL 60690 Facility Name: LaSalle County Station, Units 1 and I Inspection At: LaSalle Site, Marseilles, Il Inspection Conducted: November 1-5, 8-10, 12, 14-19, 22-24, 29, 30, 1982
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' tis /e3 Inspectors: W Gul emond
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Bb is n TREWw
'4 e/es Thb p Ma/as Approved By
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/16/93 Reactor Projects Section 2A Inspection Summary i
Inspection on Movember 1-5, 8-10, 12, 14-19, 22-24, 29, 30, 1982 (Report No. 50-373/82-52(DPRP); 50-374/82-20(DPRP))
Areas Inspected: Routine, unannounced resident inspection of icensee Actions on Previous Inspection Findings; Operational Safety Verification; Monthly Surveillance Observation; Licensee Event Reports Followup; IE Bulletin Followup; IE Information Notice Followup; Preoperational Test Witnessin ; Startup Test Witnessing; Part 21 Report Followup; Cold Weather Preparation ; Plant Trips /
Sa - ty System Challenges; and Independant Inspection Effe :. The inspection involved 197 inspector-hours on-site includin 35 inspector-hours on-site during off-shifts.
Results: Of the twelve areas inspected, no 1; ems of noncompliance were identi-fled in eleven areas. One item of noacompilance was identified in the remaining area (Paragraph 3: Failure to follow Radiation Pratection Procedures.)
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DETAILS 1.
Persons Contacted R. Holyoak, LaSalle Project Manager
- G. J. Diederich, Superintendent
- R. D. Bishop, Administrative and Support Services Assistant Superintendent J. G. Marshall, Operating Engineer
- J. C. Renwick, Technical Staff Supervisor
- R. Kyrouac, Quality Assurance Supervisor The inspectors also talked with and interviewed members of the operations, maintenance, health physina, and instrument and control sections.
- Denotes personnel attending exit interviews.
2.
Foliowup on Previously Identified Items (Closed) Condition 32 to Operating License NPF-11: This license conditica (0 pen Item 81-00-136) requires that prior to November 1, 1982, the licensee shall complete a test and shall submit its evaluation of the results which confirm the capability of the vacuum breaker valves to withstand the opening and closing forces a. ociated with pool swell.
In consultation with the
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Office of Nuclear Reactor Regulation on November 2, 1982, the inspector verified that the report was submitted as required.
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(Closed) Noncompliance Item (373/82-41-02): This item of noncompliancs documented a case in which a reactor operator trainee was observed moving control rods caring a reactor startup in the absence of direct supervision i
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as required by 10 CFR Part 55.
T5e licensee's corrective actions, received in a letter veted November 5, 1982, have baen reviewed and are deemed acceptable.
(Closed) Open Item (373/81-00-123): This open iter tracks Condition 2.C(29) of Operating License NPF-11 which requires an independent review of loop C of the Unit 1 RHR System. This item was verified closed on
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June 23, 1982. Documentatier. of the closure was inadvertently omitted from an earlier inspection report.
(Closed) Open Ites (373/81-00-108): This open ite' tracks Cor.c stion
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2.C(10)(B) of Operating License NPF-11 which requires that intermed -te Range Monitor C-51-K-601 A/H and two inch air operated g4obe valve C11-F011 be replaced if so indicated by the results of requali;tcation testing. On November 22, 1982, the licensee supplied documentation to the inspector that the Intermedt.te Range Monitor had been succes's-fully tc ted and that C11-F011 had been replaced with cm appropriately i
qualified valve.
(0 pen) Open Item (373/82-49-01): This open item tracks a 10 CFR 50.55(e)
report concerning holddown bolt over*'rquing on Emergency Core Cooling System (ECCS) pumps. The inspectors witnessed portions of the recheck
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of Unit 1 ECCS pumps on November 18 and verified proper actions taken to ensure correct torque. This item will remain open pending receipt and approval of the licensee's thirty day report.
No items of noncompliance or deviations were identified.
3.
Operational Safety Verification The inspectors observed control room operations, reviewed applicable logs, and conducted discussions with plant operators during the month of November 1982. The inspectors verified the operability of selected emergency systems, reviewed tagout records, and verified proper return to service of affected components. Tours of Unit 1 and Unit 2 reactor buildings and turbine buildings were conducted to observe the control of plant equipment conditions. fire hazards, fluid leaks, and excessive
vibrations and to verify that maintanance requests had been expeditiously initiated and resolved for equipment in need of maintenance.
a.
The following Fire Protection Deficiencies were noted and reported to station fire protection personnel:
(1) The ring was pulled on fire extinguisher #328 located outside the Control Rod Drive (CRD) hydraulic pump room.
(2) A wheelbarrel of oily rags and paper was left unattended for an extended period of time in Unit 2 turbine hall.
(3) Fire extinguisher #59 in Unit 2 turbine hall has not been inspected for three of the last four months.
(4) Fire siren F1-31 in the Unit i rsactor building was clogged with rags.
(5) Fire door #452 in Unit I reactor building was blocked open
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All of these deficiencies were immediately corrected.
b.
The inspector made the following observations during a Unit i reacter building cours on November 15 and 17, 1982:
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t 1) On November 15, 1982, a small puddle of water was on-the floor
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i of the "A" RHR pump room. The source of the puddle could not j
be located. The Rad / Chem foreman was informed about the puddle.
A licensee technician, sent to investigate, could not find the
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puddle. The inspector observed the puddle again on' November 17, 1982. The source of the puddle was determined to be a small flange leak on the suppression pool cleanup line in the vicinity of the recirculation stop valve.
(2) On November 15, 1982, the Reactor Core Isolation Cooling (RCICJ System rupture diaphragm vent line was dripping water oato the
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floor of the RCIC corner room. The licensee determined that the water was not contaminated. On November 17, 1982, the inspector verified that the leak had been stopped and the water cleaned up.
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(3) On November 15, 1982, a pile of used anti contamination clothing was found in the annulus area outside the CRD pump room. The Rad / Chem foreman was informed of the clothing. The clothing was removed.
(4) On November 15 and 17, 1982, the latching and securing (dogging) mechanism on the water tight door for the RCIC room was observed to travel past the full closed position and unlatch the door.
(5) On November 17, 1982, considerable lagging debris was observed scattered on the floor of the RCIC room, in the vicinity of the RCIC turbine.
(6) Two High Pressure Core Spray System valves and one instrument valve were observed to have packing leaks.
None of the above conditions were identified by Work Request tags.
The inspector brought each of the conditions to the attention of the Unit 1 Operating Engineer. The Operating Engireer acknowledged the concerns and indicated appropriate corrective actions would be taken.
c.
At 3:30 A.M. on November 18, the Reactor Water Cleanup System isolated in response to a high flow indication. At the time, the train "A" regenerative heat exchangers were in use.
This train was isolated and the system was recovered with train "B" heat exchangers. At 5:05 A.M. high temperature and differential temperature alarms were received for the room containing the "B" train heat exchangers. An individual sent to investigate these alarms reported that the room was full of steam. The entire reactor water cleanup system was manually isolated.
It had been determined that the problem with the "A" train heat exchangers was a leaking relief valve. The problem with the "B" train was a pipe leak on one of the three re-generative heat exchangers. The licensee returned the Reactor Water Cleanup System to service using the "A" train heat exchangers while effecting repairs to the "B" train. No neasurable activity was released as a result of the leaks.
d.
The inspector, by observation and direct interview, verified that the physical security plan was being implemented in accordance with the station security plan, and that radiation protection controls were being implemented except as noted below.
Liquid radwaste discharge records for discharges conducted while the blowdown radiation and flow monitors were out of service were reviewed. Saaples and calculations were made in accordance w.th LCP-140-7, form 1005A. All rerjuired entries were made. Double
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checks of the valve lineup for liquid releases were made in accord-ance with Action 100 of Technical Specifications, Table 3.3.7.10-1.
Double sets of analyses and calculations were made in accordanco with Action 100 of Technical Specifications, Table 3.3.7-10-1.
No discrepancies were noted associated with the discharge and/or records.
The inspector found a security related door unsecured. The licensee was notified and proper compensatory actions were taken.
It was determined that the problem occurred subsequent to a required check.
The inspector found a partially full can of soda, an empty large chewing gum package, and an apple core bettreen doors #449 and #450, an airlock from the refuel floor to the auxiliary building roof.
The inspector also notad evidence of smoking and eating throughout the Unit 1 reactor building and stopped several personnel from chewing gum and tobacco in various controlled areas. This is con-sidered a breakdown in the radiation control program and an item of noncompliance.
(373/82-52-01)
e.
During a tour of the Unit i reactor building on November 18, the Resident Inspector noted several locally mounted instruments for the Low Pressure Core Spray System (LPCS) and the Reactor Core Isolation Cooling System (RCIC) without calibration stickers or with calibration dates that appeared past due. Review of the calibration-2 cords showed that three of the instruments were overdue for calibration.
These astruments were the RCIC turbine local steam supply pressure indication, the RCIC pump local dis-charge pressure indication, and the RCIC pump discharge pressure switch which provides a permissive signal to the RCIC minimum flow recirculation valve. The Unit 1 Operating Engineer, (OE) was immediately appraised of these findings and inspector concerns over the potential impact an the operability of the RCIC syctem. The OE immediately contacted the Instrument Maintenance Department and had the three instruments calibrated.
The Resident Inspectors investigated the situation to determine why the instruments were past due for calibrarion and to establish the operability impact on the RCIC system.
Procedure LAP-300-6, "LaSalle County Station Instrument Surveillance Program," controls periodic instrument calibration. This procedure required a weekly listing of instruments past due for calibration.
This weekly listing indicated that all three instruments were non-safety related.
Per this list and calibration records, the pressure switch was due for calibration on January 24, 1982.
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censee had desigaated the instruments as not relatea to Technical Specifications or Limiting Cenditions for Operation, in accordance with LAP-lna-6, the past due calibration condition was judged by the licen; e as acceptable.
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A review of the Technical Specif.ications by the Resident Inspector revealed that none of the instruments which were past due for cali-bration, were explicitly addressed. Thus, from the standpoint of calibration control, the licensee was within prescribed procedural bounds with this situation.
The following determinations were made with respect to RCIC oper-ability. The pressure gages which were overdue for calibration provide local indication only. They were not used in either system operation or testing. Thus, they did not impact system operability.
The pressure switch, on the other hand, carried some implications of operability. This switch provided a permissive signal to the minimum flow recirculation valve to allow that valve to open under low flow conditions with RCIC pump discharge pressure at or above 125 psig. The pressure permissive was required to preclude inad-vertent flow through the RCIC pamp recirculation Ifne from the con--
densate storage tank to the suppression pool.
If the setpoint on the pressure switch was excessively low, the recirculation valve could open prematurely with the pump running, diverting flow from the reactor vessel to suppression pool thereby reducing flow to the reactor vessel to less than specified by design.
The srrveillance requirements for the RCIC system sppeared to be adequate to preclude such an occurrence. Specifically, periodic valve lineup checks are performed on the system; minimum flow recirculation valve operation is monitored during periodic system i
test demonstrations, and most significantly, RCIC flow capability.
is perjodically demonstrated undar conditions that would reveal impaired operation of the minimum flow recirculation valve. There-fore, even though the pressure switch was overdue for calibration,-
system operability was not impairid.
In response to the inspector's concerns on overdue calibrations, the Plant Superintendent convened a meeting with staff members on November 22 to review the length and content of the overdue-for-calibration list. The conclusion reached by the licensee.was that ncne of the items remaining on the list was of safety or major operational sinaificance. The inspectors are currently reviewing this list independently of the licensee.-
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No additional items of noncompliance or deviations were identified.
4.
Monthly Surveillance Observation The inspector observed the following Technical Specifications required surveillance testing.
LIS-NR-04 APRM Rod Block and Scram Functional Check LIS-NR-04 APRM Gain Adjustment Functional Test LIS-NB-01 Reactor Vessel Low Water Level Scram and Primary Containment Isolation Calibration and Functional Test
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LIS-hR-02 Intermediate I.ange Monitor Rod Block and Reactor Scram Functional Test The inspector verified that the testing was performed in accordance with approved procedures, that test instrumentation was calibrated, that limit-ing conditions for operation were met, that removal and restoration of the affected components were accomp'ished, that test results conformed with Technical Specifications and y;ocedure requirements, that test re-sults were reviewed by personnel other than the individual directing the test, and that any deficiencies identified during the cesting were properly reviewed and resolved by appropriate management personnel.
No items of noncompliance or deviations were identified.
5.
Licensee Event Reports Followup Through direct observations, discussions with licensee personnel, and review of records, the following Event Reports were reviewed to determine that reportability requirements were fulfilled, immediate corrective action was acccmplished, and corrective action to prevent recurr;.cc had been accomplished in accordance with Technical Erecifications.
373/82-112/03L-0 Out of Specification Reactor Low Water Level Scram Setpoint 373/82-133/03L-0 Ground Fault In RHR Service Water Pump 373/82-117/03L-0 Failed Unit 1 Division 1 Battery Charger 373/82-107/03L-0 Scram Due To Low CY Tank Level *
373/82-94/03L-0 Recirc Pump Seal Failure 373/82-96/03L-0 Failure Of RCIC Testable Check Valve To Indicate Closed 373/82-132/03L-0 Failed RHR Sample Pump 373/82-120/03L-0 Loss Of Off-Gas Hydrogen Analyzer Channel 373/82-123/03L-0 Missing Snubber Lord Pin 373/82-122/03L-0 Incperable Lake Blowdown Flow Instrument 373/82-128/03L-0 Inoperab'e SBGT Flow Transmitter 373/82-125/03L-0 Failed Fuel Pool Ventilation Radiatior. Monitor
- LER submitted late due to reclassification as reportable on followup reviews.
LER 82-119/03L-0, Docket 50-373, documents an event in which the mode switch was taken from startup to run with one train of the Control Room Emergency Filtration System inoperable. This was in violation of Technical Specification 3.0.4 which stated that entry into an operational condition shall not be made unless the limiting conditions for operation are met without reliance on provisions contained in the Action Statements.
The licensee identified this item as a noncompliance. As such, and in accordance with the NRC Enforcement Policy, a Notice of Violation was not issued.
This is the second occurrence of this type. The first occurred on June 6, 1982 when the unit was taken from Mode 4 to Mode 3 with inoperable reactor coolant leakage detection systems.
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In following 2p on this particular LER, the inspector reviewed the licensee's procedures, instructions, and checklints for control of mo'e changes with the results as noted below.
Technical Specific.*. ions required normal unit startup Procedure LGP #1-1 which required the completion of Master or Minimum Startup Checklists (LGP 1-S1 and LGP 1-S2 respectively) as a prerequisite to piacing the mode switch to startup.
These checklists were reviewed for content, format, and implementation cuidance.
'he content was deemed comprehensive.
Some format problems were noted.
Specifically, LGP 1-Si noted that Technical Specification requir-ments were indicated by a *. However, the following procedure steps were not so indicated:
A-6, 10, 11, 12; C.6, 14, 13, 16; F.11.
Numerous steps in LGP 1-S2 have the same deficiency. With respect to impicmentation, no guidance was provided on the time frame for comple-tion of these checklists. Thus, reasonable assurance is not provided to assure that plant conditions will not change between the time the checklist was performed and commencement of startup. However, if the appropriate checklists are completed in a timely fashion, adequate assurance is provided to assure that Technical Specificaticn require-ments for going to Mode 2 would be satisfied.
With respect to going from Mode 2 to Mode 1, tbs following comments tre applicable. Technical Specification 3.2.1 required that Average Planar Linear Heat Generation Rs*es (APLHGR) not exceed specified limits when thermal power was greater than or equal to 25% of rated thermal power.
Per the La Salle Unit 1 Techncial Specification, Surveillance Requirements 4.2.1, APLHGR values were to be verified as follows when the plant is greater than or equal to 25% power:
(a) at least once per 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s: (b)
within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of a 15% pnwer increase; (c) initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a limiting control rod pattern.
LGP 1-1 raquires that APLHGR be checked within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of reaching each 15% power increment starting at 15% power. Thus, conformance with APLHGR limits need not be verified prior to entering operating conditions l
where the limits are applicable. The same commente apply to the Average i
Power Range Monitor (APRM) flow biased simulated thermal power-upscale scram trip setpoint, the flow biased simulated thermal power-upscale control rod block trip setpoint, the minimum critical power ratio limits, and the j
linear heat generation rate limit.
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The inspector could locate no procedural provisions governing changes between Mode 5 and Mode 4 or Mode 4 and Mode 3.
Based on these findings, the licensee's procedural controls for mode changes appear to be caly marginally adequate. The findings were dis-cussed with the licensee during an exit meeting on December 3, 1982.
This item remains open (373/82-52-02).
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6.
IE Bulletin Fsllowup The inspector verified that IE Bulletin 82-03, Revision 1 was received by the licensee and reviewed for information purposes. This bulletin requires no further action of LaSalle at this time.
No items of noncompliance or deviations were identified.
7.
IE Informat ion Notice Followup (Closed) IE Information Notice 82-42: This Information Notice documents the discovery of defects in Panasonic Model 801 and Model 802 thermo-luminascent dosimeters. The inspector determined that these dosimeters are not enc.,loyed at LaSalle.
No items of noncompliance or deviations were identified.
8.
Unit 2 Preoperational Testing Unit 2 Preoperational Test activities as noted below were monitored during the inspaction period:
a.
PT-LP-201, Low Pressure Core Spraj (LPCS) System Valve Logic Testing:
The inspector reviewed the procedure and found all prerequisites to be met with the exception of a few which had an officially approved waiver signed and filed within the procedure.
LSU 500-1, Revision 6, Attachment A, " Jumpers, Lifted Leads, and Relay Blocks used in Test PT-LP-201" was a form filed within the master test copy. This form was used to keep track of circuit modifications, the panel in which the modific&tions were made, the test procedure step for which the modifications were required, and the test prouedure step which returns the circuit to normal.
This was an effective means of keeping control over circuit modifications for testing.
b.
During testing ot' the LPCS test line isolation valve on November 5, 1982, the valve position indication in the control room did not agree with local valve indication.
A Discrepancy Report was initi-ated by the Test Engineer. Part of the testing for the LPCS test line valve involved using the armed pushbutton to manually give a system initiation signal. Thi action affects other valves as well.
The LPCS injection valve properly opened as a result cf this action.
This injection velve was left open ell weekend even though the testing did not continue past 4:00 P.M.
(The manual LPCS valve downstream of injection valve was tagged closed, however).
c.
During testing of the LPCS pump breaker on November 8, 1982, the Divisia,n 1 diesel generator was inadvertently started when a techn;uian apparently installed a jumper which was intended to prevent the diesel start, in the wrong location. The problem was attributed to poor labelling at the jumper installation location.
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d.
The inspectors also monitored shift. turnover of the Unit 2 reactor operators during the LPCS pump breaker test. Based on the turnover, the inspector concluded that the off going operator was not fully aware of the nature and scope of the testing activities. This was -
brought to the attention of the Unit 2 Operating Engineur and the Technical Staff Supervisor.
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On November 13, the inspector monitored the performance of portions of Unit 2 High Pressure Core Spray System Testing (PT-HP-201). The testing consisted of operating the high pressure core spray pump with the pump taking suction from and discharging to the Unit 2 suppress!.on pool. The testing was conducted in accordance with the approved pr -
cedure. The inspector noted continuous involvement of the Unit 2 Reactor Operater. No discrepancies were identified.
No items of noncompliance or deviations were identified.
9.
Unit 1 Startup Testing On November 17, the inspector monitored the performance of the generator load reject portion of test procedure STP-27. The test was performed in accordance with the procedure, the required data was obtained, the initial conditions and prerequisites were satisfied, and the test pari.-
1:1 pants were adequately briefed in their duties prior to test perform-ance. The test simulated a generator load rejection from between 20 and 25% thermal power by opening the circuit breakers isolating the generatcr from the electrical distribution network. As expected, this resulted in a turbine trip. The transient was wholly accommodated by the operation of four turbine bypass valves. Reactor vessel water level varied less than two inches during the transient.
On November 19, the inspector monitored Step I of STP-23.
During this portion of the test, step changes in reactor vessel water level demand were introduced and transient plant response data was obtained. The test was performed in accordance with the procedure, the required data was obtained, the initial conditions and prerequisites were satisfied, and the test participants were adequately briefed in their duties prior to test performance.
No items of noncompliance er deviations were identified.
10.
Part 21 Report Followup On October 29, 1982, I T T Barton submitted a 10 CFR, Part 21 Report identifying a potential defect in their Model;s 763 and 764 electronic transmitters. The inspector determined that, based on discussions with licensee personnel, these model transmitters are not employed at LaSalle.
Thus, this item is considered closed.
No items of noncompliance or deviations were identified.
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11.
Cold Weather Pgmaration Thc inspector reviewed the licensee's procedures for cold weather preparatloa to ascertain their effectiveness. The inspector noted that
no one procedure directed personnel to verify heat tracing had been replaced following maintenance on a system. However, the licensee felt l
that the Equipment Out-Of-Service Procedure does satisfy this concern
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by requiring a system to be returned to normal. This is considered an open item (373/82-52-03).
The licensee had recently completed the required surveillance and the inspector reviewed this for proper documentation.
No items of noncompliance or deviations were identified.
12.
Plant Trips / Safety System Challenges At 4:52 P.M. on November 9, Unit 1 experienced a reactor vessel low water level automatic scram. Vessel IcVel reached the low level scram point following a loss of feedwater flow induced by a malfunction in the controller for the "A" turbine d:iven feedwater pump. The resulting level and steam pressure fluctuations caused the feedwater heater string to isolate cutting off feedsater to the reactor vessel.
All systems functioned normally in response to the event with the excep-tion of the recirculation pumps. There exists a non-safety related logic which causes the recirculation pumps to shift from fast speed to slow speed when total feedwater flow falls below 30% of rated flow. During this event, this circuitry tripped the recirculation pumps off fast speed as designed. However, neither pump started on slow spec =1 as designed.
This problem has occurred before, but has been limited to the "B" recir-culation pump only. The licenses is reviewing the subject logic in concert with the NSSS vendor in t. effort to resolve the problem. The inspector is following this issue as an open item (373/32-52-04).
No ECCS systems were challenged during this event. The reactor war returned to criticality at 9:50 P.M. on November 10.
At approximately 1:50 P.M. on November 15, Unit 1 experienced an automatic
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l reactor scram. The scram occurred following a low condenser vacuum turbine i
trip. Condenser vacuum was lost wLen both opereting circu'ating water pumps simultaneously tripped. All systems functioned normally. No ECCS systems were actuated.
Investigation revealed that the circulating water pumps tripped when the circulating watcr bay high level trip switch was inadvertently bumped, causing it and the pumpe ?n trip. The reactor was taken critical at 6:05 A.M. on November 16.
At 2:19 P.M. on November 16, Unit #1 was manually scrammed in anticipa-tion of an automatic scram. At the time of the scram, the unit was at approximately 500 psig with turbine warming it progress. The scenario leading to the scram proceeded as follows. Feed water te the reactor was being supplied by the condensate and condersate booster pumps. The I
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flow rate was being manually controlled by use of the manual isolation valve upstream of the feedwater regulating valve due to excessive leakage through the feedwater regulating valve. A Turbine Driven Feedwater Pump (TDRFWP) was operating at minimum speed, with its discharge valve shut, recirculating back to the condenser.
In accordance with the normal startup procedures, the operator opened the discharge valve on the TDkFWP in preparation for placing it on the line. This created a flow path from the condensate booster pumps to the reactor vesuel which bypassed the feedwater regulating valve and its upstream manual isolation valve. The flow of water through this bypass path caused a rapid upward excursion in reactor vessel level and a positive reactivity insertion.
As 1cvel and power began to rise, the operator tripped the TDRFWP. However, the discharge valve, having a seal in contact in its positioning circuit, continued to open maintaining the bypass path. An APRM half scram was received.
In order to terminate the reactivity excursion, the operator manually tripped the reactor. He then shut the Main Steam Isolation Valves (MSIV's) to prevent water from entering the main steam lines.
Level stabilized at 140 inches. No ECCS systems were actuated.
Ir order to orevent recurrence of this type event, the licensee is parsuing procedure changes and system modifications.
In the interim, administrative controls have been establi:hed to require that the Motor Driven Feedwater Femp (MDRFWP) be started and operated pumping through the feedwater regulating valve until such time as plant pressure reaches the shutoff head of the condensate booster pumps. This would effectively eliminate the bypass flow path. The subject of procedure changes and system modifications sill be tracked as an open item (373/82-52-05).
Unit 1 experienced an automatic scram at 8:31 A.M. on November 21.
The scram occurred during surveillauto testing on the MSIV's.
The surveill-ance procedure called for trip;1.ig one of the two MSIV position inputs to the RPS on the "a' MSIV, performing a slow closure on the "B" MSIV, and verifying that a half scram occurs. For reasons yet to be determined, the second position input to RPS from the "A" MSIV was in the tripped condition prior to test performance. Thus, when "B" MSIV was slow closed i
a full scram resulted. The licensen is modifying the surveillance pro-cedure to require that the status of all position relays be verified pric" to performing the surveillanca test.
System nodifications are being considered to annunciate the status of the position relays. This item remains open (373/82-52-06).
All systems functioned normally on the scram. No ECCS systems were
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actuated. The reactor was returned to criticality at 9:00 P.M. on
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November 21.
I No items of noncomplience or deviations were identified.
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13.
Independent Inspection Effort (
On Novemb6e 17, the inspectors received Generic Letter 82-23, dated October 30, 1982. This letter identified an inconsistency between l
requirements ot' 10 CFR 73.40(d) and Standard Technical Specifications
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for performing audits of Safeguards Contingency Plans. The former required an annual independent review and audit of the contingency plans. The latter required a biannual independent review and audit of contingency plans.
In order to ascertain if this apparent inconsistency was applicable to LaSalle, the inspector reviewed the security audit requirements of the licensee's Technical Specifications and Secuilty Plan.
Technical Specification 6.1.G.1.b.6 required that biannual audits would be per-formed in accordrnce with the Commonwealth Edison Company Quality Assurance Program at the responsibility of the Manager of Quality Assurance of the Facility Security Plan and Implementing Procedures.
The Facility Security Plan, which incorporated the contingency plans, contained a requirement for a comprehensive independent annual review and audit of the Facility Security Plan and its implementation. These findings were discussed with a representative of the NRC Region III Security and Safeguards Branch. Pursuant to the discussion, the con-trolling requirements were determined to be those contained in the Facility Security Plan. The audit requirements contained in the Technical Specifications were in addition to those contained in Facility Security Plan. Thus, the inconsistency addressed in Generic Letter was deemed to be not applicable to LaSalle.
No items of noncompliance or deviations were identified.
14.
Exit Interview The inspector met with licensee representatives (denoted in Paragraph 1) throughout the month and at the conclusion of the inspection period and summarized the scope and findings of the inspection activities.
The licensee acknowledged these findings.
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