IR 05000338/1982008
| ML20055A145 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 05/14/1982 |
| From: | Hardin A, Dante Johnson, Shymlock M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20055A103 | List: |
| References | |
| 50-338-82-08, 50-338-82-8, 50-339-82-08, 50-339-82-8, NUDOCS 8207150518 | |
| Download: ML20055A145 (13) | |
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UNITED STATES
i NUCLEAR REGULATORY COMMISSION
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c REGION 11
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o 101 MARIETTA ST., N.W.. SUITE 3100 ATLANTA, GEORGIA 30303 o
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Report flos. 50-338/82-08 and 50-339/82-08 Licensee: Virginia Electric and Power Company P. O. Box 26666 Richmond, VA 23261 Facility Name: North Anna Units 1 and 2 Docket Nos. 50-338 and 50-339 License Nos. NPF-4 and NPF-7 Inspection at North Anna site near flineral, Virginia Inspectors: O
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D. 'F. Johnson
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Date Signed
/9Cbud w s/w/sv il B. Shymlock Date Signed Approved by: hI 6/' V/ N fut h A. K. Hardin, Acting Chief, Division of Date Signed Project and Resident Programs SUMMARY Inspection on flarch 6 - April 5,1982 Areas Inspected This routine, inspection by the resident inspector involved 289 inspector-hours on site in the areas of followup of previous inspection findings, licensee event reports, previously identified items, post implementation review of NUREG-0737 l
items, licensee conditions, refueling activities, surveillance, maintenance activities, and plant operations.
Resul ts
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Of the nine areas inspected, four violations were identified.
A.
(Failure to adhere to Technical Specification limiting condition for operation and its j
associated action statement, paragraph 5.a.).
B.
Failure to adequately review a temporary procedure change, paragraph 5.a.).
C.
(Failure to follow procedures, paragraph 5.b. ).
D.
(Failure to adhere to Technical Specification requirements regarding high radiation areas, paragraph 5.b).
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8207150518 820707 PDR ADOCK 05000338 i
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DETAILS 1.
Persons Contacted Licensee Employees
- W. R. Cartwright, Station Manager
- E. W. Harrell, Assistant Station Manager
- J. A. Hanson, Superintendent - Technical Services
- J. R. Harper, Superintendent - Maintenance
- S. L. Harvey, Superintendent - Operations G. Paxton, Superintendent - Administrative Services J.11. Mosticone, Operations Coordinator R. A. Bergquist, Instrument Supervisor
- A. H. Stafford, Health Physics Supervisor J. P. Smith, Engineering Supervisor F. Terminella, Engineering Supervisor P. T. Knutsen, Engineering Supervisor J. R. Stratton, liechanical liaintenance Supervisor D. E. Thomas, Electrical Supervisor
- A. L. Hogg, Jr., Site QC, Manager
- F. P. fliller, QC Supervisor
- M.
E. Fellows, Staff Assistant K. A. Huffman, Clerk Other licensee employees contacted included eight technicians, six'
operators, and four mechanics.
- Attended exit interview 2.
Exit Interview The inspection scope and findings were summarized on April 5,1982, with
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those persons indicated in paragraph 1 above. The violations presented in i
paragraph 5 were discussed with station management at that time and acknowledged.
3.
Licensee Action on Previous Inspection Findings (Closed) Violation 339/81-28-01 Operations mode change without returning the bypass trip breakers to normal.
Periodic Test PT-36.1 Reactor Protection i
and ESF Logic Test for both units 1 and 2 were revised February 24, 1982.- A
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coament was added in the initial conditions.to inform Shift Supervisor that a mode change cannot be initiated during the performance of this procedure.
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Unresolved Items
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Unresolved items were not identified during this inspection.
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5.
Enforcement Actions a.
Failure to. adhere to Technical Specification limiting condition for operation and failure to adequately review a temporary change to a procedure.
On flarch 7,1982, with Unit 2 in tiode 3, and prior to cooling down and depressurizing for a refueling outage, the licensee decided to functionally test the automatic opening of the accumulator isolation valves at more than 2010 psig using actual reactor coolant system (RCS) pressure.
flormally the test of the accumulator valves is done by simulating an RCS pressure of 2010 psig while the actual RCS pressure, as required by Procedure flo. 2-PT-56.3, is established at more than 800 psig but less than 1000 psig.
(Step 2.2 of Procedure 2-PT-56.3).
In order to run the test a deviation to 2-PT-56.3 was required.
A deviation was processed in accordance with the licensees procedures including review by the Station fluclear Safety and Operating Committee and approval of the deviation by the Station Manager.
The test deviation allowed closing all three accumulator valves at an actual RCS pressure (as opposed to simulated) slightly below 2000 psig then raising actual RCS pressure to 2010 psig.
The test was successfully
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run in this manner.
Overlooked in the process of approving the deviation and running of the test was Technical Specification requirement 3.5.1 which requires that each RCS accumulator shall be operable with the isolation valve open when pressurizer pressure is more than 1000 psig and average temper-ature is more than 350 F.
An action statement associated with TS 3.5.1 allows only one accumulator to be inoperable due to a closed isolation valve.
Since all three valves were closed with actual RCS pressure near 2000 and remained closed to demonstrate the valves would auto-matically open when RCS pressure was raised to 2010, TS 3.5.1 and its associated action statement was not met.
The significance of the event is related to 1) the simultaneous closure of the isolation valves with RCS pressure above 1000 psig and temper-ature above 350 F violated TS 3.5.1.
This condition existed for about 2 minutes and during this time the valves would have reopened auto-matically on an increase in pressure above 2000 psig.
2) of more concern is the failure of the onsite safety reviu committee to recognize that approval of a deviation from Procedure 2-PT-56.3 would result in the inadvertent approval to deviate from Technical Specifi-
cation 3.5.1.
The failure to meet Technical Specification 3.5.1 is identified as violation (50-339/82-08-01).
The failure to adequately review the temporary change to Procedure 2-PT-56.3 is a violation of Technical Specification 6.5.1.6 (50-339/82-08-02).
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b.
Failure to follow procedures.
(1) Technical Specification 6.12.1 requires that a radiation monitoring device which continuously indicates the radiation dose rate in the area is required upon entry into a high radiation
area.
Technical Specification 6.8.1 requires that procedures be implemented and maintained. The licensee's Health Physics Procedure, Section 6, Exposure Control also requires a dose rate meter for entry into a high radiation area.
Entry into such areas with this monitoring device may be made only after the dose rate level in the area has 50en established and personnel have been made knowledgeable of them.
On January 10, 1982 an operator entered a high radiation area in order to check out a malfunctioning pressure indication. He did not notify health physics personnel to ascertain dose rate in the area nor did he have in his possession a radiation dose rate meter to measure area radiation.
On February 19, 1982, two engineers were taking data readings for a test in an area adjacent to the charging pumps which is a high radiation area. Neither individual had a radiation measuring device as required and in addition they admitted that they were not qualified in use of a dose rate meter.
and 6.8.1 (50-338/82-08-02 and 50-339/82-08-03)pecification 6.12.1 The above events are a violation of Technical S
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(2) Technical Specification 6.8.1 requires that procedures be imple-mented and maintained.
(a) On March 10, 1982 two valves (RC-125 RC-126) associated with the casing cooling system were specified by valve lineup procedures to oe closed by independent verification for the performance of a containment functional test.
The valves were erroneously left open which resulted in pumping approximately 20,000 gallons of water fran the casing cooling tank into the containment sump that overflowed onto the containment floor.
The cause of this event was personnel error and failure to follow procedures.
The operators performing the valve lineup signed and attested that the subject valves were closed when in fact they were open.
The operators relied upon unqual-ified personnel to inform them of valve position rather than actually checking them as required.
This is a violation of T.S. 6.8.1 (50-339/82-08-04)
(b) Ibintenance Procedure M.li. ADM-1.0 Section III requires approved maintenance procedures for safety-related systems.
thintenance Procedure 11MP-C-GP-1, step 5.3 requires specific approved of the Mechanical liaintenance Supervisor and Quality Control before proceding to disassemble a service water pump.
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On February 5,1982 Unit 2, service water pump 1A was disconnected, removed from service at the service water pump house and transferred to #1 track bay for further main-tenance.
The above activity was performed without an approved procedure.
The maintenance foreman in charge of the job arrived at the hold point for disassenbly of the pump.
He attempted to contact Quality Control personnel but was unable to reach them by phone on site.
Knowing the job was of high priority he proceeded with the disassemly of the pun.? without recciving proper written approval as required by procedures.
Items (b) and (c) are futher examples of personnel error and failure to follow procedures.
This is a violation of T.S. 6.8.1 (50-339/82-08-05).
6.
Plant Status Unit 1 During this inspection period the unit operated at or near capacity load except for the following:
On April 1 the reactor was automatically tripped from 99.4% power due to Power Range Hi Flux Rate Trip.
The reactor trip was caused by a short to ground while main.anance was being performed on the steam generator support RTD temperature monitors. This momentary short caused a voltage dip on the vital power supply panel 1-EP-CB-04A.
This voltage dip was detected by the Channel 1 Nuclear Instrument System control and instrument power causing the high flux rate and high flux setpoint trips. The power range channel N44 was previously placed in trip due to an inoperative detector. Therefore the 2 of 4 logic was completed for a reactor trip.
Unit 2 During this inspection period Unit 2 was shutdown (liarch 6) to conduct the first refueling outage. All of the fuel was removed from the reactor vessel by April 1.
Numerous items such as, special test, inservice inspections and plant modification per license conditions were started. These items will be identified and discussed over this outage period.
7.
Followup of Previously Identified Items (Closed) 338/78-42-04, Guidance for inverse multiplication startup plots.
1-0P-1.5 Unit Startup from Hot Standby Condition (Mode 3) to Startup Condition (flode 2) with Reactor Critical at >5% power and 1-0P-1.6 Plant Hot Startup both dated December 23, 1981 werE reviewed.
Both procedures gave specific instructions on using an inverse count rate ratio plot with an attached graph for data collection.
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(Closed) 338/78-42-05, Licensee to insure wind direction channel check performed adequately.
1 Log 6A Back Board Operator Log dated December 23, l
1981 gives specific direction on comparing upper to lower wind speed and
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direction to each other.
It also gives directions on how to determine if a detector is malfunctioning.
(0 pen) 338/82-08-01, Incorrect description on 1-Log-6A dated December 23, 1981. The last two log items on page 3 (Wind Speed Lower, Wind Speed (
Backup) are not correct.
They should read wind Direction Lower and Wind Direction Backup.
(Closed) 338/78-32-05, Develop policy so that technical staff and operators are consistent in mathematical calculations (specifically round-off error).
A memorandum was distributed describing an Engineering Study giving instructions on use of significant digits.
The specific problem was noted in 1-PT-23 Quadrant Power Tilt Ratio. This PT was revised flay 23, 1980, with instructions on use of significant digits.
(Closed) 338/81-13-03,339/81-10-03, Submittal of additional information per 10 CFR 50.48 and 10 CFR 50 Appendix R (Fire Protection).
The submittal package serial #310 dated Itay 19, 1981, was sent to NRR.
(Closed) 339/78-35-01, Drain valve on the residual heat removal relief valve, header to the pressurizer relief tank, was not shown on system print.
This was discovered during a hydro operation.
Valve operating print (FM-94A Rev 8) Residual Heat Removal System now shows this drain valve 2-RH-54.
(Closed) 338/81-11-03,339/81-07-01, Signoff of Design Changes.
Project Procedure llanual PPM-9, " Assignments and Responsibilities for the issuance and Sign-off of Design Changes / Field Changes" was issued September 21, 1981.
It specifically assigns control for the control copy of the design change to the assigned field project engineer.
The sign-off requirements for the Quality Control Inspector is also delineated.
(Closed) 338/78-32-02, Licensee improvements to chilled water system to control containment temperature. Apparent problem seemed to be the position of ring duct exhaust dampers and the vent seals around the ducts.
Addi-tional information is in IEIR 338/82-04.
(Closed) 339/81-16-15, Determine and correct cause of failure of six radios used by fire brigade. The reason for the failure was determined to be poor battery charge.
There were five new radios and chargers purchased and placed under the Fire !!arshall's control. There are three located in the staging area and two in his office.
They are inspected on a monthly basis and have been operational since January 15, 1982.
(Closed) 339/81-16-16, Procedures should specify that gas cylinders are to be secured in their storage or use locations.
Vepco's Accident Prevention Manual paragraph 204.12 and 204.15 address use, storage, and securing compressed gas cylinders.
(Closed) 338/81-13-04, 339/81-10-04, Followup renewal of SR0 license.
The SR0 license was issed for this individual April 9,1981.
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(Closed) 338/81-13-05,339/81-10-05, tiedical Evaluation of injured operators. Administrative Procedure Amt 29.4 gives specific direction on how this will be handled.
The procedure meets 10 CFR 55.11 and 10 CFR 55.41
requirements.
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8.
Licensee Event Report (LER) Followup i
The following LER's were reviewed and closed.
The inspector verified that -
reporting requirements had been met, causes had been identified, corrective
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actions appeared appropriate, generic applicability had been considered, and the LER forms were complete. Additionally, for tb4
'" orts identified by asterisk, a more detailed review was performed to verc, that the licensee
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had reviewed the event, corrective action had been takta, no unreviewed
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safety questions were involved, and violations of regulations or Technical J
Specification conditions had been identified.
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338/81-80 The response time for IJ emergency diesel undervoltage relay j
was nonconservative.
- 338/81-74 Delta T/TAVE Protection channel B was removed from service for high and/or erratic readings.
338/81-75 A containment isolation trip valve failed to close on a
phase ' A' isolation.
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- 338/81-76 Safety injection due to inadvertant steam dump valves opening.
i 338/81-77 Axial Flux Difference deviated greater than target band.
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338/81-87 Declared non-reportable.
338/81-88 Declared non-reportable.
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' A' Service water return header was isolated to repair pinhole leak.
338/82-07 Fire doors 571-7, A19.1, and A44-1 were declared inoperable.
l 338/82-09 Electric fire pump (1-FP-P-1 was removed from service to l
repair the other pump.
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- 338/82-10 Dose Equivalent iodine level exceeded T.S. limit following
Rx trip.
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j 338/82-11 Axial Flux Difference deviated greater than target band.
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- 338/82-12 Water level in Emergency Condensate Storage tank dropped
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below minimun volume.
339/81-58 The normal offsite electrical feed to 2H Emergency Bus was secured.
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339/81-61 C boric acid storage tank had a low boron concentration.
339/81-65 Power range channel N-41 bistable setpoint were found to be non-conservative.
339/81-67 Fire door S54-13 was inoperable.
339/81-69 High negative flux rate trip was found to exceed T.S.
setpoint.
- 339/81-71 Discovery of inoperative snubber on main steam line.
339/81-74 Failure to perform twelve hour surveillance on Radiation f onitoring Equipment Check (2-PT-37).
339/81-79 Delta Flux exceeds target flux difference.
339/81-86 Axial Flux Difference deviated from target band.
339/82-05 Fire door 571-16 not latched.
339/82-12 Fire doors 1180-2 not closed nor would A-80-2 latch properly.
9.
Three !!ile Island Task Items a.
Task item II.K.3.9 require licensee's to remove the rate input to the pressurizer power operated relief valves (PORV's).
The rate circuitry
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is part of the Westinghouse design PID (preoperational, integral and
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derivative) controller for one of the two ' installed PORV's.
The licensee has changed the rate time constant for ooth units to zero.
For further assurance the rate circuit is removed fron. service by removal of the JK jumper on circuit card C8-128.
s The licensee received Westinghouse Technical' Bulletin NSD-TB-81-12 entitled "Inadvertant PORV Opening" and has determined that the subject Foxboro Controller tiodel 62 is not used at North Anna Units 1 and 2 and that NSD-TB-81-12 is not applicable in that control function for the PORV's is done electronically in the process racks.
Based on the above task item IIK.3.9 is considered closed.
b.
Task Item II.E.4.2.5 requires that the containment pressure setpoint that initiates containment isolation for non essential penetrations should be reduced to the minimum compatible with normal operating condition.
The containment setpoint presure that initiates containment isolation for nonessential systems is 17 psia.
This setpoint is based on total system error of approximately 2.2 psi (3.65% of full scale) which consists of errors due to instrument accuracy / repeatability, instrument drift, temperature affects and calibration inaccuracies.
This error added to a maximum expected containment pressure of 14.8 psia 'provides the existing setpoint of 17 psia. This is the miniaw setpoint
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compatible with normal operating conditions and are less than the FSAR and Technical Specification requirements.
The containment is maintained subatmospheric when above 200 F.
It would be inappropriate to use a lower setpoint due to potentially severe operational compli-cations without a commensurated improvement in safety.
For example, loss of containment vacuum, in the absence of any potential release of radioactivity, would result in containment isolation. An orderly shutdown in such an incident would be complicated and delayed due to system isolations.
lne above provides sufficient information to conclude that the existing containment isolation pressure setpoint for North Anna, Units 1 and 2 meets the NUREG-0737 requirements.
This position was evaluated by f4RR and considered acceptable.
Based upon the above, NUREG-0737, task item II.E.4.2.5 is considered closed.
10.
Unit 2 License Conditions The following items relate to spr-ific license conditions some of which are required to be completed prior ming power operations.
These items will be followed during this o,
, and reported during this report and in Inspection Report 338/82-13 and 339/82-13.
License Condition 2.C(4)(b).
No later than Jbne 30, 1982, VEPC0 shall replace Rosemont pressure transmitters and differential pressure trans-mitters, and pressure transmitters and differential pressure transmitters from Barton lot I with suitably qualified devices.
This is identified as IFI 339/82-08-05.
Based on a review of Unit 2 per NUREG 0588, these specific transmitters may
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l not be qualified for the post accident environment.
During post accident l
environment, these transmitters may fail to perform their safety function.
The design change (DC) package has been approved and is being worked to conduct the transmitter change out with environmentally qualified trans-
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mitters.
The design change package is identified as DC-815-088, Replacement of d/p and Direct Pressure Transmitters.
This design change package will be completed prior to resuming power operation following this refueling.
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License Condition 2.C(4)(c); No later than June 30, 1982, the wide-range
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resistance temperature detectors (RTD) for the reactor coolant system shall l
be qu311fied for radiation exposure for the 40-year plant life and appro-
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priate exposure condition due to design basis accidents.
Pending completion of such qualification and acceptance by the Commission, VEPC0 shall replace i
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cach of these detectors at each refueling outage.
This is identified as IFI (
339/82-08-06.
However several items must be resolved with RTD vendor product before complete qualification can be confirmed.
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In the interim VEPC0 hay sutanted a lelt'er to NRR dath Feh, aryh4,19d2
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serial No.105, which request' approval of the fohing aftp.natceplan.
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The Electric Power Resnrch Ins 6tute (EPRI) has developed a method for q
RTD time response.,The method is called Loop Current Ste._
measuring (LCSR).The licensee plan is tUte$t the peactor coolant system Response the LCSR method each quarter. This will narrow and wide range RTD'
using qualified RTD is av511able.
be conductv each quarter until a They do not intend to replace the RTD during this, fueling outagh License Condition 2.C(4)(e); "By ho # dte'r than June 30, 1982, all
z safety-related electrical equipment dn 'thefacility shell be qualified in
" Guidelines for Evaluating Environmental qualificatichs of Class IE S[
accordance with the provisions of: Division of,0peratircj Reactors
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Electrical Equipment in 0perating Reacto s" (56R Guidelines); or NUREG-0588,
" Interim Staff Position on Environmental qualification of Safety-Related Electrical Equipment," Decem6er 197).' Copies of these documents are attached to Order far flodification of License NPF-7 dated November 7,1980.
This is identified as' IFI 339/82-08-07.
Tie licensee is submitting an Operating Licensee Amendment.
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The addition of the secondary overlod(trotection devices to all normally
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energized power and control circuits thalenter containnent will provide adequate protection of the mechanicallintegrity of the electrical
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pene tra tions.
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y This secondary protection will be, either 'a circuit breaker, a set of fuses,
or relays in series with the above noted circuits.
The DC package lists the i
specific circuits.
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This DC package was approve 25% cam ~plete at the ecd of this inspection
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License Condition 2.c(10) No later than October 11,1980, VEPC0 shall submit,
a design for the backup overcurrent protection system for containment i
electrical penetrations for Commission review and approval.
The backup
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system shall be installed and operational prior to resuming power operation <3 following the first refueling outage. This is identified as IFI
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339/82-08-08. This modification identified as DC 81-21B Secondary Pro-
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tection of Electrical Penetrations is provided 1.0 protect the electric 61
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penetration assemblies.
It will allow these penetration assemblies to I
withstand without loss of mechanical integrity, the short circuit current if the primary overload circuit protection device fails.
License Condition 2.c(12)(c) Prior to startup following the first refueling outage:
Items 1 and 5:
Diverse Power Supply for Th and Tc. This is identified as IFI 339/82-08-09. With the completion of a review of equip-ment supplied with power from vital, semi-vital or 125 VOC buses per IE Bulletin 79-27 the following was noted. The power sources for the wide range cold leg temperature (TC) loops (T-1410, T-1420, and T-1430) were supplied from Vital Bus 1-11.
Also that the power sources for the wide range hot leg temperature (TH) loops (T-1413, T-1423 and T-1433) were supplied from Vital Bus 1.I.
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The proposed modification will change the circuitry such that the associated hot leg temperature loop and cold leg temperature loop of each Reactor Coolant loop will be supplied by the same power source.
However it will not be the same as loop 2 and 3.
The DC package 80-S52 Diversifying RCS Wide Range Temperature Loops (Hot and Cold Legs) is approximately 75% complete at the end of this inspection period.
License Condition 2.c(15)(a); VEPC0 shall submit the details of the inspection program for control rod guide thimble tube wall wear for Commission approval.
This is identified as IFI 339/82-08-10.
The licensee submitted to flRR the guide thimble surveillance program conducted at Salem Unit 1 as part of a cooperative owner's group agreement.
The letter sent to flRR was dated fiarch 4,1982 serial flo.124.
License Condition 2.c(15)(b), VEPC0 shall install inspection ports in the steam generators.
This is identified as IFI 339/82-08-11.
Discussions with NRR and the licensee are continuing at this time and the current status is not resolved.
License Condition 2.c(15)(c); VEPC0 shall remove and inspect the recir-culation spray pumps inside containment and replace pump bearings if nccessa ry.
A similar inspection shall be performed at least once every five years thereaf ter.
This is identified as IFI 339/82-08-12.
The A inside recirculating spray pump has been disassambled and the bearing replaced.
This pump is awaiting optical alignment.
This job is approximately 50%
complete at the end of this inspection period.
Licen e Condition 2.c(15)(d) VEPC0 shall install leak test connections on the RHR isolation valves.
This is identified as IFI 339/82-08-13.
The design change package 81-S40 RHR Isolation Valves Leak Rate Test Connection requires that test connections be added.
These test connections are 3/4" stainless steel valves that will be added to the existing valve body drain r,i ppl e. This will be accomplished on fiOV-2700 and f10V-2701 supply isolation valves and MOV-2720A and it0V-2720B return isolation valves.
This will allow pressurization between the disk and allow leak rate testing of these valves per T.S.
This job is approximately 80% complete at the end of this inspection period.
License Condition 2.c(15)(e); VEPC0 shall demonstrate by test the backup depressurization capability of the PORV's using the same shutdown procedure as described in VEPCO's procedure 2-0P-3.2 dated 7/23/80.
This is identi-fied as 339/82-08-14.
This test has been completed and a report is being prepared.
When this report is complete it will be reviewed.
License Condition 2.c(15)(f) VEPC0 shall submit for Commission approval, the results of the tests applicable to florth Anna Power Station, Unit 2, of a study concerning mixing of added borated water and cooldown under natural circulation conditions.
This is identified as IFI 339/82-08-15.
The licensee conducted this test on liarch 7,1982.
The test was monitored by the resident inspector.
A write up including data on this test was submitted by the licensee to flRR dated April 5,1982 Serial flo. 200.
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License Condition 2.c(15)(g); VEPC0 shall retest all engineered safety features reset control actions to verify proper reset action.
This is identified as IFI 339/82-08-16.
This test was partially completed during the performance of 2-PT-83.5 Emergency Bus Blackout and SI Functional Test.
This PT will be conducted prior to startup.
License Condition 2.c(15)(h)(1); complete a formal training program for all the mechanical and electrical maintenance and quality control personnel, including supervisors, who are responsible for the maintenance and avail-ability of the diesel generators.
The depth and quality of this training program shall be at least equivalent to that of training programs normally conducted by major diesel engine manufacturers.
This is identified as IFI 339/82-08-17.
A letter from Colt Industries dated April 8,1981 to Vepco was reviewed.
It identified that fourteen Vepco Electric ltaintenance and Mechanical Itaintenance personnel attended and satisfactorily completed a training course on maintenance of the 38 TD 8-1/8 engine.
License Condition 2.c(15)(h)(2); The ac prelube pump shall be modified to DC power operation and shall be installed in the system to operate in parallel with the engine driven lube puup.
The prelube pumps shall be provided with manual start.
In an automatic or manual start, the prelube pump shall be operated only during the engine cranking cycle or until a satisfactory lube oil pressure is established in the engine main lube oil distribution header.
This is identified as IFI 339/82-08-18.
The licensee and flRR are discussing this item based on the licensee letter of liarch 18, 1982 serial #141.
The status of this item has not been resolved.
License Condition 2.c(15)(h)(3); the diesel generator operating procedures shall be modified to require loading the engine up to 50 to 75 percent of full load for one hour after eight hours of continuous no load operation.
This is identified as IFI 339/82-08-19. This is has been completed and will be reviewed during the next inspection period.
License Condition 2.c(15)(h)(4); The day tank overflow line shall be rerouted to return excess fuel to the seven day fuel oil storage tank.
This is identified as IFI 339/82-08-20.
The licensee is requesting an amendment to this requirement in letters to llRR dated February 17, 1982 serial #079 and Itarch 18, 1982 serial #141.
They have requested that the day tank overflow line back to the seven day storage tank be deleted.
They are requesting the addition of a dedicated day tank high level alann in the control room. Also the addition of two pressure switches in the drain line on the bottom of the day tank.
One of these switches will provide a redundant stop signal to the lead transfer pump when the pressure indicates a near-full condition in the day tank.
The second pressure switch will provide a similar signal to the back-up transfer pump.
The dedicated day tank high level alann has been installed for both diesel day tanks and has been tested. This design change is identified as81-03B.
License Condition 2.c(15)(h)(5); Each seven day fuel oil storage tank shall be provided with a seismic Category I, tornado missile, and flood protected emergency fill line.
Each fill line shall have a shut-off valve, a strainer, and a truck fill connection consisting of a hose coupling with cap
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and chain. This is identified as IFI 339/82-08-21.
The addition of an emergency fill connection for each of the underground storage tanks has been completed.
Further followup will be conducted during the next inspection pe riod.
This design package is identified as81-502.
License Condition 2.c(15)(h)(6); With respect to vibration of Instruments
and Controls, Vepco shall either provide test results and results of analyses which qualify the engine skid mounted control cubicles for the severe vibrational stress that will be encountered during engine operation, of floor mount the skid mounted panels and control equipment presently furnished with the diesel generators.
This is identified as IFI 339/82-08-22.
The control panels and relay panels which are currently
mounted on the diesel generator skid are to be moved to floor mounts.
the l
gage and relay panels will be floor mounted in seismic supports, without l
moving the panels from their present locations.
This design change is identified as81-504 and is approve 60% complete at the end of this i
reporting period.
11.
Plant Operation Numerous containment enteries were made during the current unit 2 refueling ou tage. These entries were made to review work in progress, overall housekeeping and safety during work activities, assure adherence to licensee Health Physics Policies.
On a regular basis radiation work procedures (RUP's) were reviewed and the specific work activity was monitored to assure the activities were being conducted per the RUP's.
Radiation protection instruments were verified operable and calibration / check frequencies were being met.
The inspectors kept informed on a daily basis of overall status of both units and of any significant safety matter related to plant operations.
Discussion were held with plant management and various members of the operations staff on a regular basis.
Selected portions of daily operating logs and operating data sheets were reviewed daily during this report period.
The inspectors conducted various plant tours and made frequent visits to the control room. Observations included: witnessing work activities in progress, status of operating and standby safety systems and equipment, confirming valve positions, instrument readings and recordings, annunciator alarms, housekeeping and vital area controls.
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