IR 05000245/1979009
| ML19253B523 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 08/22/1979 |
| From: | Mccabe E, Shedlosky J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML19253B519 | List: |
| References | |
| 50-245-79-09, 50-245-79-9, 50-336-79-09, 50-336-79-9, NUDOCS 7910160358 | |
| Download: ML19253B523 (18) | |
Text
.
.
.
U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT Region I Report No. 50-245/79-09, 50-336/79-09 Docket No. 50-245, 50-336 License No. DPR-21, DPR-65 Priority:
Category:
C
--
Licensee: Northeast Nuclear Energy Company Post Office Box 270 Hartford, Connecticut 06101 Facility Name: Millstone Nuclear Station, Units I & II Inspection at: Waterford, Connecticut 06385 Inspection condu ed:
arch 17 through April 30, 1979 Inspectors:
um ror s
f.'T.' Sh ojky,ResidentInspector date signed Approved by:
f C hMs, h e /22/ 7'i E. C. McCabe, Chief, Reactor Projects date signed Section No. 2 Inspection Summary:
March 17 - April 30,1979 (Combined Inspection Report 50-245/79-09 and 50-336/79-09 Routine, onsite regular, backshift and weekend inspection by the Resident Inspector (28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br /> Unit 1; 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> Unit 2). Areas, inspected included accessible portions of the Unit i reactor, turbine, radioactive waste and intake buildings; the Unit 2 primary containment, auxiliary, turbine and intake buildings; radiation protec-tion; physical security; fire protection; plant operating records; surveillance testing; maintenance and core alterations.
A briefing was conducted by the NRC for Unit 2 operations department personnel and plant management concerning the nuclear incident at Three Mile Island plant.
Findings: one item of noncompliance
~
(failure to, comply with the requirements of a Radiation Work Permit).
- -,
.
_,,
m
,,y 7010160 3 5 2
.
DETAILS 1.
Persons Contacted The below listed technical and supervisory level personnel were among those contacted:
J. M. Black, Superintendent, Unit 3 P. Callaghan, Unit 1 Maintenance Supervisor F. Dacimo, Station QC Supervisor E. C. Farrell, Superintendent, Unit 2 M. Griffin, Station Security Supervisor H. Haynes, Unit 2 Instrumentation and Control Supervisor R. Herbert, Superintendent, Unit 1 J. Kelly, Unit 2 Operations Supervisor E. J. Mroczka, Superintendent, Plant Services J. F. Opeka, Station Superintendent R. Place, Unit 2 Maintenance Supervisor P. Przekop, Unit 1 Engineering Supervisor W. Romberg, Unit 1 Operations Supervisor S. Scace, Unit 2 Engineering Supervisor F. Teeple, Unit 1 Instrumentation and Control Supervisor Additional NRC Personnel W. J. Raymond, Reactor Inspector, Region I B. A. Wilson, Operator Licensing Branch, Office of Nuclear Reactor Regulation 2.
Review of Plant Operations - Plant Inspections The inspector reviewed plant operations through direct inspection and observation during routine power operations and a scheduled refueling outage of Unit 2.
Inspections were made of the accessible portions of the Unit 1 control room, reactor, turbine and radioactive waste buildings, the intake structure and the Unit 2 control room, primary containment, auxiliary and turbine buildings, the condensate polishing area and the intake structure.
During this inspection, activities in progress were normal plant power operations and surveillance testing of Unit 1 and a refueling outage of Unit 2.
The inspector observed operations in the control room including shift turnovers and back shift activities.
Inspections were made on weekends and holidays.
-.
,,
Il i
%
a.
Instrumentation Control room process instruments were observed for correlation between channels and for conformance with Technical Specification requirements.
No unacceptable conditions were identified.
b.
Annunciator Alarms The inspector observed various alarm conditions which had been received and acknowledged.
These conoitions were discussed with shift personnel who were knowledgeable of the alarms and actions required.
During plant inspections, the inspector observed the condition of equipment associated with various alarms.
No unacceptable conditions were identified.
c.
Shift Manning The operating shifts were observed to be staffed to meet the operating requirements of Technical Specifications, Section 6, both to the number and type of licenses.
Control room and shift manning were observed to be in conformance with Technical Specifications and site administrative procedures.
d.
Radiation Protection Controls Radiation protection control areas were inspected.
Radiation Work Permits in use were reviewed, and compliance with those documents, as to protective clothing and required monitoring instruments, was inspected.
There were no unacceptable conditions identified.
e.
Plant Housekeeoing Controls Storage of material and components was observed with respect to preven-tion of fire and safety hazards.
Plant housekeeping was evaluated with respect to controlling the spread of surface and airborne contami-nation.
There were no unacceptable conditions identified.
f.
Fire Protection / Prevention The inspector examined the condition of selected pieces of fire fighting equipment.
Combustible materials were being controlled and were not found near vital areas.
Selected cable penetrations were examined and, thus, fire barriers were found intact.
Cable trays were clear of debris.
g.
Control of Equipment
--
g.1 i
.
.
During plant inspections, selected equipment under safety tag control were examined.
Equipment conditions were consistent with information in plant control logs.
h.
Instrument Channels Instrument channel checks were reviewed on routine logs.
An independent comparison was made of selected instruments.
No unacceptable conditions were identified.
i.
Equipment Lineups The inspector examined the breaker position on all switchgear and motor control centers in accessible portions of the plant. Equipment conditions were found in conformance with Technical Specifications and operating procedure requirements.
j.
Techinical Specifications for Refueling - Mode 6, (Unit 2)
On April 10, the inspector verified, through inspection of plant equipment, Technical Specifications required for Mode 6, Refueling activities.
Those Specifications included 3.1.2.1 - Reactivity Control, Boration Systems - Flow Paths; 3.1.2.3 - Charging Pump; 3.1.2.5 -
Boric Acid Pumps; 3.1.2.7 - Borated Water Sources; 3.4.2 - Reactor Coolant System, Safety Valves; 3.8.1.2 - Electrical Power Systems; 3.8.2.2 - AC Distribution; 3.8.2.4 - DC Distribution; 3.9.1 -Boron Concentration; 3.9.2 - Refueling Instrumentation; 3.9.3 - Decay Time; 3.9.4 - Containment Penetrations; 3.9.5 -Communications; 3.9.6 -
Containment Crane Operability; 3.9.8 - Coolant Circulation; 3.9.9 -
Concainment Radiation Monitoring; 3.9.10 - Containment Purge Valve Isolation; 3.9.11 - Water Level - Reactor Vessel; 3.9.12 - Water Level -Storage Pool; 3.9.13 - Storage Pool Radiation Monitoring; 3.9.14 - Storage Area Ventilation.
No unacceptable conditions were identified.
3.
Review of Plant Operations - Logs and Records During the inspection period, the resident inspector reviewed operating logs and records covering the inspection time period against Technical Specifications and administrative procedure requirements.
Included in the review were:
Shift Supervisor's Log 3/17 to 4/30/79
-
- -
,
l
%
Plant Incident Reports 3/17 to 4/30/79
-
Jumper and Lifted Leads Log all active entries Maintenance Requests and Job Orders -
all active entries Safety Tag Log all active entries
-
Plant Recorder Traces daily during control room
-
surveillance Plant Process Computer Printed daily during control room
-
Output surveillance Key Control Leg 3/17 to 4/30/79
-
Several entries in these logs were tne subject of additional review and discussion with licensee personnel.
No unacceptable conditions were identified.
4.
Briefing on Description of Circumst6nces and Preliminary Chronology of the Three Mile Island Accident by NRC Personnel for Unit 2 Licensed Operators A review of the description of circumstances described in Enclosure 1 of IE Bulletin 7905 and the preliminary chronology of the Three Mile Island Unit 2, March 25, 1979 accident was conducted in three presentations conducted April 20 and 21.
All Unit 2 operations department licensed operators attended.
One person with a reactor operators license assigned to Unit 2 engineering did not attend the presentation; however, he did view pre-recorded video tapes and will be briefed by the resident inspector.
This review was directed toward understanding:
(1) the extreme seriousness and consequences of the simultaneous blocking of both auxiliary feedwater trains at the Three Mile Island Unit 2 plant and other actions taken during the early phases of the accident; (2) the apparent operational errors which led to the eventual core damage; (3) that the potential exists, under certain accident or transient conditions, to have a water level in the pressurizer simultaneously with the reactor vessel not full of water; and (4) the necessity to systematically analyze plant conditions and parameters and take appropriate corrective action.
5.
Control of Work in Radiation Areas - Unit 2 a.
The inspector observed the implementation of radiation protection controls.
These were reviewed against the requirements of 10 CFR 20, the operating license Technical Specifications, station administrative control and operating procedures.
The following Radiation Work Permits (RWP) were reviewed on the indicated days:
March 20
--
.,.
,
l 9,
..
790639 Auxiliary Building LLRT 790569 Health Physics Department Blanket RWP 790636 Containment - Trash and Laundry Pickup 790672 Containment - Tool Crib 790674 Containment - Reft 91 Hoist Modification 790668 Containment - Housekeeping 790682 Containment - Move Reactor Head Bracket, Install Hydroset 790673 Containment - Install Cavity Seal 790683 Containment - Rework Ex-Core Detector Flanges March 27 790997 Containment - Tool Crib 790998 Containment - Electrical Penetration Tests 791010 Containment - Instrument Calibration 791011 Containment - Instrument Calibration 791012 Containment - Instrument Calibration 791022 Containment - Inspection 791024 Containment - Decontaminate Electrical Penetration 791014 Containment - Refuel Machine Auxiliary Hoist Work April 9 791686 Containment - Decontaminate West Walkway (14' 6")
791687 Containment - Decontaminate and Change Herculite Between Elevator and Stairway (22')
791674 Containment - Decontaminate Scaffolding (22')
791664 Containment - Construction Liasion Supervision Neutron Shield Storage 791681 Containment - Tool Crib 791662 Centainment - Tool Crib 791600 Containment - Torque Steam Generator Manways 791667 Containment - Neutron Shield Storage 791679 Containment - Welding and Grinding RCPC Platform 791663 Containment - Electrical Penetration Testing 791675 Containment - CEA Movement & Eddy Current Testing April 11 791767 Spent Fuel Handling Area - Sleeving Irradiated Fuel 791784 Spent Fuel Handling Area - Sleeving Irradiated Fuel 791669 Spent Fuel Handling Area - Fuel Pool Verification and Inspection
.
1" i-
791770 Spent Fuel Handling Area - Move Fuel Inspection Basket onto Inspection Elevator 791627 Spent Fuel Handling Area - Fuel Movement and Inspection b.
On April 5, two workers, supporting the installation of steam generator number 2 primary manways, were found with facial contamination.
The work was controlled by RWP791479. The contamination was found when leaving the control area. Contamination levels were about 200,000 DPM and were decontaminated to 3,000 DPM.
Whole body counting indicated an initial internal deposition of 2424 nanocurie C060
-
Worker A 4116 nanocurie C058 Worker A
-
1772 nanocurie C060 Worker B
-
2942 nanocurie C058 Worker B
-
located on the lower face, neck and upper chest.
Licensee evaluation of this and subsequent whole body counting results indicated the following quarterly lung and whole body exposures:
Worker A - Quarterly lung - 1102 millirem Quarterly whole body - 33 millirem Worker 8 - Quarterly lung - 906 millirea Quarterly whole body - 27 millirem and a resulting 155 hours0.00179 days <br />0.0431 hours <br />2.562831e-4 weeks <br />5.89775e-5 months <br /> at 10CFR20 maximum permissible concentrations.
These estimated exposures will be reviewed during future health physics inspections.
The inspector interviewed the workers involved and health physics technician covering the work.
The contamination is believed to have occurred while preparing the steam generator manway diaphrams by hand for welding.
Some contamination may have occurred while removing protective clothing.
Survey results of the diaphrams indicated 2.5 Rem / hour contact with a closed Beta window, and over 5 Rem / hour with an open Beta window.
RWP 791479 included as required protective clothing a fresh air supplied hood and lapel air sampler.
There was some confusion in the work area as to the requirements of the RWP Controlled Area and a tent set up inside that controlled area.
Contrary to the require.ments of the Technical Specifications Section 6, the HP Operating procedure 2903 and RWP 791479, neither the Health Physics Technician nor the two workers wore fresh air supplied hoods.
Addition-
~ ~
. g _,
$
'
s
_y
.
ally, one of the workers was not provided a lapel air sampler.
This is identified as an Item of Noncompliance.
The inspector reviewed the licensee's corrective actions including their investigation, analysis of exposures, assignment of MPC hours, and the continued implementation of work control procedures at Units 1 and 2.
The inspector had no additional questions at this time.
6.
Loss of Flow in Reactor Shutdown Cooling System and Subsequent Reactor Coolant System Heatup into Mode 4 (Hot Shutdown) Unit 2 Unit 2 commenced its refueling outage on March 10.
Prior to commencing core alterations, the reactor coolant system (RCS) was drained down to the midplane of the reactor vessel hot leg.
The steam generator primary manways were removed for the eddy current testing of tubes in accordance with Technical Specification 4.4.5.0.
The plant was in cold shutdown Mode 5.
The Shutdown Cooling System (SDC) a s in service with one Low Pressure Safety Injection (LPSI) pump; RCS temperature wcs being maintained at 150 F.
The licensee experienced difficulty in achieving the normal SDC system flow rates.
This was attributed to the reduced net position suction head of the LPSI pump due to the lower reactor vessel level.
Reactor vessel level at the midplane of hot leg is normally sufficient to allow SDC system flow, but not high enough to overflow into the steam generator plenums.
In this case. LPSI pump lost suction pressure and started cavitation.
Attempts to regain flow were made by venting and transfer to an alternate pump.
At about 190 F RCS temperature primary :ontainment integrity was verifiad and personnel were cleared from the containment.
The LPSI pump norma; suction from the Refueling Water Storage Tank (RWST) was opened to provide pumping water to the pump suction. This restored flow.
Peak RCS temperature was 208 F.
The plant had an unplanned transition into Mode 4, Hot Standby.
Opening the flow path to the LPSI pump from the RWST resulted in additional water being added to the RCS and a spill of an estimated 15,000 gallons of water into the primary containment pumps.
The inspector reviewed selected Technical Specification requirements for Mode 4 and Mode 5 operation and verified that equipment operability require-ments were met.
The licensee's investigation into this incident disclosed the need to remove an air pocket in a H P noint of the SDC System suction line.
That high point collected noncondensable gases, and this condition was aggravated when the RCS was drained for steam generator entry.
The licensee correction actions included the evacuation (vacuum primping) of the SDC System high point.
OP2310, Shutdown Cooling, Revision 3, Change ?., dated April 2,
- -,
,,
h
%
l
.
1979 increased the scope of the operator's actions in the event SDC is lost.
These actions include the evacuation of the SDC suction line high point.
While reestablishing SDC flow by adding water from the RWST, RCS temperature cooled from 206* F to 70 as RWST water was added. Temperature then increased to 176 F when mixing occurred.
The licensee analysis of this avent indicated that fracture tcughness design basis limitations were not downgraded.
7.
Core Thermal Power Exceeded Rated Thermal Power In addition to the heat balance calculations performed by the plant process computer, the licensee has d:.veloped a correlation between turbine first stage pressure and feedwater flow.
This provides a more accurate means of determining feedwater flow than the feed flow instruments.
On March 30, with the reactor core near the end of life, extraction steam was removed from both high pressure feedwater heaters.
The increased core inlet subcooling allowed thermal power to be increased to 100% of Rated Thermal Power (2011 MW). At 0300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br /> on March 31, the reactor was brought to 2003 MW (t) as calculated by the process computer.
The shift reactor engineer assumed that the turbine first stage pressure feedwater flow correlation was not valid with extraction steam removed from the high pressure feedwater heaters.
On April 4, dering a review of engineering data, the licensee determined that the turbine first stage pressure correlation was valid.
The calculated peak reactor power on March 31 was 2045 MW (t) or 101.69%
of rated power.
The licensee had lowered power on the morning of April 2 to insure that the reactor was below 2011 MW (t) by any calcula-tional method.
Additionally, the licensee briefed responsible personnel of this event.
The inspector had no additional questions at this time.
8.
ECCS_ Flow Verification Surveillance On April 7, the inspector witnessed portions of the ECCS Flow Verification Surveillance test.
That testing required by Technical Specification 4.5.2.f was performed in accordance with SP2604Q, Revision 1, Change 1, dated February 2, 1979.
The procedure was accepted by the plant safety committee during meeting 7911.
The inspector reviewed the procedure content, including the stated acceptance criteria, against the Technical Specification require-ments.
The inspector verified the fulfillment of the test prerequisites and initial conditions.
High pressure safety injection pump flow verifica-
-.
n i.,
1,
tion (Procedure sections 7.1 and 7.2) was observed from the control room and test data was gathered independently.
The inspector noted that the criteria for readjustment of flow balancing valves stated in sections 7.1.7, 7.1.14, 7.2.5, and 7.2.13 is more restrictive that the section 2.1 Test Acceptance Criteria and Technical Specification 4.5.2f.
He was informed that the added conservatism was to allow using control room linear flow indications.
The test was performed using a more accurate computer digital display.
The licensee's resolution as to the procedural requirements for readjustment of flow balancing valves is identified as an open item to be followed up during a future inspection (336/79-09-01).
9.
Sleeved Control Element Assembly (CEA) Guide Tube Inspection Program (Unit 2)
The Sleeved CEA Guide Tube Inspection Program was submitted by the licensee to the NRC i? letters dated December 8, 1978, January 24, 1979, and March 29, 1979.
T x inspector observed the establishment of prerequisites and the performance of eddy current testing of sleeved CEA guide tubes on April 8 and 9.
The testing was being conducted in accordance with 00000-ESS-134, Eddy Current Test Procedure for Sleeved CEA Guide Tubes in Fuel Assemblies, Revision 0, dated March 21, 1979.
This document was found to have implemented the commitments contained in the documents referenced above.
The licensee reported the results of the inspection program by letter to the NRC dated April 17.
The test results stated in that correspondence do not contradict data obtained by the inspector.
The inspector had no additional questions at this time.
10.
Auxiliary Feedwater Pump Turbine Drive Governor (Unit 2)
The inspector verified that the licensee has taken action on the recommenda-tions.Sade by the Auxiliary Feedwater Pump Turbine vendor concerning the required bleed down of hydraulic fluid from the governor's speed setting cylinder.
The vendor had identified a set of circumstances which may allow the turbine to trip on overspeed if it is restarted within 30 minutes after shutdown.
The inspector found that the licensee had implemented the vendor's recommen-dations.
OP2322, Auxiliary Feedwater System, Revision 4, dated May 11, 1978, was reviewed.
Stated precautions include if the " turbine is to be restarted within 30 minutes of a prior turf % ~rhutdown, caution must be taken in opening t'1e combined steam stop k ens 3 ring that the turbine comes on the governor at the minimum 991
,
- -
(n
\\
w
s
These caut. ins are implemented in procedural steps 7.2.2, 3 and 4.
The procedure also requires that on shutdown the Auxiliary Feedwater turbine speed is reduced to minimum (1400 RPM) prior to closing the combined steam stop.
The speed control and combined steam stop are controlled remotely from the control room or locally in the Turbine Building Steam Driven Auxiliary Feed Pump Room.
There were no unacceptable conditions identified.
11.
Statu of Previously Identified Open Items The in oector reviewed the status of items which had been called out during previous NRC inspections as requiring followup.
The following items, taken from the referenced inspection rep,rts, are considered to be closed.
Unit 1 76-73-03 Spent Fuel Pool Filter Checklist Revir, ion and Panel Labeling.
The inspector reviewed the operating procedure in use at the filter control panel and had no additional questions.
76-33-02 Nuclear Review Board Member Listing Site files of NRB membership and meeting minutes have been updated.
76-33-06 FSAR Organization Update.
There is no requirement to keep the FSAR updated.
77-31-02 LPCI/ESW Heat Exchanger Examinations.
This work has been accomplished during the last two refueling outages. Repairs have been made to the service water, water boxes during the 1979 refueling outage.
77-31-03 Stack Flow Instrument.
Flow instruments have been added to plant exhaust paths during the 1979 refueling outage.
This allows remote monitoring of total stack flow and replacement of local differential pressure instrumentation.
77-31-04 Electrical Cable in Drywell near IC4.
During the 1978 refueling outage, the cable near IC4 was replaced in relocated conduit.
The isolation condenser condensate return line was insulated and additional drywell cooling installed.
There were no failures of cabling in the drywell through the 1979 refueling outage.
77-31-06 IRM Calibration - Calibrations are performed using a standard current source as required by specification 4.1.A, table 4.1.2.
- -
,
y, w
a
.
.
77-31-07 Review of Actions taken because of NRC Circular 7713. This was reported in inspection 50-245/79-02.
78-07-03 & 04 Failure to test all Refueling Interlocks.
Sur-veillance procedures were updated to test interlocks which stop the platform if more than one control rod is not fully inserted and the mode switch is in refuel. Also, interlocks for the rod block with all rods in, platform over the core and mode switch in startup as tested.
During the 1979 refueling outage, the licensee discovered that these updated procedures did not verify refuel bridge position switch operation. The surveillance procedures were again changed to functionally test these components by driving the bridge over the core.
78-17-03 The Plant PORC performed a review of the status of LPCI and Core Spray piping NDT work based on NUSCO recommendations.
78-17-08 & 09 The status and documentation of the class 2 and 3 inservice inspection program and results was the subject of reports on the program.
78-19-01 Failure to perform required surveillance of APLHGR due to demanding the wrong computer option.
Data reviewed by the inspector documented the recording of this data daily during reactor operation at greater than 25% rated thermal power.
78-34-01 Revise daily surveillance log to include reviewer signoff.
The operations department daily surveillance log has been revised to include the required review by the 1600-2400 Shift Supervisor.
78-34-02 Revise procedure to include vacuum breaker testing. SP626.2, Manual Operation of Relief Valves when Reactor is at Low Pressure has been revised to require that SP632.4, Suppression Chamber Drywell Vacuum Breaker Exercise, be performed to meet the require-ments of Technical Specification 4.7.A.5.
78-34-04 Failure to follow diesel surveillance procedure. SP668.1 data sheet 668.11 has been revised to include required diesel generator loading of 2665 to 2700 KW. The inspector reviewed the data from completed testing and found that the diesel met this acceptance criteria.
78-34-06 Acceptance criteria to be added to a surveillance procedure.
SP625.4, Emergency Condensate Transfer Pump Operational Readiness test has been revised to include acceptance criteria for pump
~
-,.
I
~
.
.
head and flow of greater than 93 psig pump discharge pressure and less than or equal to 104 psig pump different;al pressure repsectively.
Unit 2 77-03-04 Measures established to assure that purchased material conforms to-procurement-documents.
The MEPC, Revision 5, dated December 15, 1978, has been revised to include the commitments and positions of NNECO letter dated May 18, 'r977 and NRC letter dated June 21, 1977.
77-13-01 Licensee completion of corrective actions to reduce oil leak on RCPS and provide protective screens.
RCP junction boxes have been equipped with protective screens.
77-13-02 Resolution of Conflicting Requirements for TS 3.5.1.
These conflicts were resolved by Technical Specification Amendment 36 issued in January 1979.
77-25-03 Review and evaluate the basis and calcuations for TSP Requirements considering the allowable Boric Acid Inventory.
TSP inventory requirements were updated in Technical Specification Amendment 36.
77-25-04 Revision of ACP-QA-2.06 concerning control of Red Tags and Shift Supervisor Determination of Bypass / Jumper Status.
ACP-QA-2.2 has been revised (Rev. 10 dated October 20, 1978) to reference ACP-QA-2.06.
77-25-05 Evaluate and report design changes effected by bypass / jumper control, pursuant to 10 CFR 50.59(B), and implement appropriate administrative controls.
Turbine Coast Down Circuit has never been placed in service.
It was evaluated by the licensee, who decided not to place it in service.
The circuit was removed in accordance with a plant design change request.
If the turbine coast down circuit is desired in the future, it must be processed through another plant design change request.
77-25-06 Convert existing bypass / jumper controls to permanent design changes and complement appropriate administrative controls to define the allowable period of jumper control.
The licensee has completed a review and disposition of jumpers.
The licensee's position is that jumpers may not substitute for design changes.
~"
7o
.
.
The revision of jumpers implemented is a normal inspection item.
The inspector had no additional questions at this time.
77-25-07 Submit LER describing inoperability of LPSI pump during corrective maintenance.
An LER number 7752, was submitted.
77-25-08 Review Technical Specification 3.2.3 to be consistent with capability for measuring azimuthal power tilt. The Technical Specifications revision 32, dated May 17, 1978, have deleted this requirement for all modes except mode 1.
77-25-09 Correct Boric Acid leakage problems during refueling outage and clean areas affected by leakage.
This area was inspected during the 1979 refueling oatage.
The inspector had no additional questions at this time.
77-29-01 Procedure changes and modification as referenced in PORC 77-119 to be completed during-December 1977 outage.
Plant Design Change Request PDCR-2-209-77 to remove Emergency Diesel Generator automatic start on Reserve Station Service Transformer (RSST) Lockout, or RSST Audion.
This eliminated items described in FSAR sections 8.3.3.1 c, d, e, & f.
The Emergency Diesel Generator auto start signals are SIAS, and emergency bus undervoltage as required in the Technical Specifications.
The inspector had no additional questions at this time.
77-31-07 Procedure for Pressurizer Spray Valve Replacement. Pressurizer spray valves were replaced during the 1978 refueling outage under plant maintenance procedures.
77-31-08 Procedure for retest of charging pumps Pulsation dampers were installed during the 1978 outage.
This work included vibration monitoring.
Maintenance procedure MP2703C5 includes instructions covering these pulsation dampers.
77-31-09 Startup Testing Procedure Review.
NRC review of 1978 startup testing has been completed.
This review is a routine inspection item for the 1979 outage.
78-05-01 Failure to follow procedure ACPQA3.02 with regard to interim changes to procedures.
The licensee committed to the review of ACPQA3.02 paragraph 6.7.1.5 requirements with responsible personnel, the review of all controlled copies of station procedures and an audit of this action by the Quality Assurance organization.
This information was promulgated by the Station Superintendent in a memoranda MP-1104 dated March 9, 1978.
- -.
.,
s s
!
y
,
.
78-06-03 Revision to Technical Specification 4.5.2.C.3 on minimum amounts of TSP to maintain proper pH.
This was cleared by Technical Specification Amendment 36.
78-06-04 Procedure 2506 Loss of Coolant Change to reflect shutdown of second diesel.
OP2506, Loss of Coolant Revision 3, dated June 20, 1978, and OP 2503, Electrical Emergency, Loss of Normal Power Revision 3, dated December 21, 1978, requires the transfer of loads and the securing of one emergency diesel generator to meet the design loading and day tank capacities stated in the FSAR.
78-06-07 Review of Analysis of Charcoal Filters in EBFS LER 78-01. This item was inspected and closed in NRC inspection report 50-336/79-01.
78-06-09 Modification of PE0 Log Sheets.
Observations for snubber deteriora-tion are made during Plant Equipment Operator rounds.
Millstone Unit 2 has a program to implement Technical Specification require-ments for snubber surveillance.
The licensee does not intend to restate the requirements for visual inspection for implementation on a daily basis.
The inspector had no additional questions.
78-06-11 Calibration of Maintenance Test Instruments.
The vibration instrument in question (MM2101) is not used for Class 2 and 3 ISI programs.
The Fast Fourier Transformer (FFT) spectral analysis equipment used for ISI work is covered under the program described in ACPQA9.04, Control and Calibration of Measuring and Test Equipment.
The instrument in question (MM2101) may not be used to evaluate the performance of Category I equipment unless included in the ACPQA9.04 program as required by ANSI N 45.2.4, Section 2.3.
The resident inspector has observed the use by mechanics of the MM2101 during post repair checkout of safety related equipment.
However, in each case, the data from the FFT spectral analysis equipment was used to certify equipment condition.
78-07-02 Design reviews of documentation not adequate, failure to implement QA and QC requirements.
The licensee has had Job Orders R-70457, R-70457A and R-70457B reviewed by Engineering and Quality Assurance.
The licensee instructed the Engineering and Quality Assurance organizations of the necessity to insure that all phases of design and inspection have been prepared prior to the initiation
-.
,,
,
s
.
-,
-
.
of work.
ACP-QA-304 Design Change Control and the Plant Design Change Request Checkoff List 305 have been revised to prominently state these requirements.
78-09-01 Qualification Records for Equipment Used to Make Modifications.
The tools and procedures used for CEA guide tube and sleeve expansion was reviewed during inspection 50-336/79-02.
There were no problems identified in this area.
78-14-01 Determination of accuracy of remote indicators for ECCS motor operated valves and establishment of appropriate acceptance criteria to assure correct open position.
The licensee verified the proper position of the ECCS flow balancing valves to meet the Technical Specifications required flow.
The motor operator drive gear was scribed and indexed to allow valve position to be veri-fied visually.
The motor operator limit switches were tested to demonstrate repeatable opening of the valve to the correct throttle position.
The resident inspector observed ECCS performance testing which verified proper flow balance.
78-18-02 Completion of Review and Approval of Startup Test Procedures 789 and 7810.
NRC review of 1978 startup testing has been completed.
This review is a routine inspection item for the 1979 outage.
78-30-01 Failure to update a surveillance procedure to reflect a plant modification.
The licensee's evaluation of this noncompliance had identified two problems.
First, the failure to revise a procedure.
The appropriate Plant Design Change Request (PDCR)
which added air operators to the valves in question was closed out without revising SP2601A and 8.
Changes have been implemented to include a Plant Design Change Request checkoff list.
The licensee has reviewed all of the 1978 refueling outage.
The second problem involved operations personnel use of procedures without updating them.
The operations department personnel were made aware of this occurrence and were reminded of the need to be alert to identify required procedure changes and initiate the resultant change.
The inspector reviewed SP2601A, Borated Water Sources and Flow Path Verfication and Boric Acid Pump Operability Test; SP2601B, Borated Water Source Flow Path Verification Monthly; and OP2304C, Boric Acid System and found the referenced changes implemented.
ACP-QA3.04, Design Change Control, Revision 4, dated 2/23/79, was reviewed by SORC at meeting 797 paragraph 6.10.
Followup includes the requirements that the person coordinating the design change
" ensure that all affected Station Procedures are updated to
- _.
,,
,
I
~
_
-
.
reflect the design changes.
When the design change requires a change to the surveillance procedure, this change should be implemented within one month of completion of the design change or before the next surveillance interval falls due, whichever is sooner." Also to " submit any necessary FSAR Change Requests..."
SF 305 PDCR Checklist includes the requirements for the review for the need of procedure update, FSAR changes, ISI program changes, and MEPL changes.
78-30-03 Failure to include two contai'
t isolation valves on a containment integrity verification procedure. SP2005A, Verifying Containment Integrity includes hydrogen sample stop valves 2-AC-46 and 2-AC-51.
78-31-01 Failure to document maintenance checkout and return to service on maintenance requests.
ACP-QA-2.03, Performing NonCategory I Work, Revision 3, dated February 13, 1979, paragraph 5.4.
pesponsi-
.
bilities require that the Shift Supervisor or Senior Control Operator is responsible for the " completion of required surveil-lance for redundant safety related equipment and for performance of the functional check following maintenance".
Also, paragraph 6.5 completion requires in part that "when the MR returns to the Control Room for the fu ctional check, the Safety Tags will be lifted and the check cr '-ted...When the work for a particular job is complete and the tags are cleared but the functional check is not to be accomplished immediately, the ege'pment is to be tagged to prevent operation until the functional check is completed...".
78-31-02 Record retention requirements for maintenance requests used for safety related activities.
ACP-QA-7.03, Performing NonCategory I Work, Revision 3, dated February 1, 1979, paragraph 6.5 and 6.7 requires that the QC copy of t.ie Maintenance Request is retained by the records facility and it must be found legible prior to acceptance by the records facility.
78-31-03 Failure to incorporate internal parts that perform safety related functions of pumps, valves and emergency power supplies in the MEPL.
The inspector reviewed MEPL, Revision 5, dated December 14, 1978, and concluded that equipment components as pistons, cylinders, liners, bearings, etc. h;d been included.
78-37-01 Failure to pressure test personnel air lock per Technical Specifi-cation 4.6.1.3.6.
The licensee has reviewed this occurrence with the responsible supervisors and has reminded them of the importance of full compliance with the Technical Specifications.
A department
- -
7-n ~
,
.
.
status board Dacks up computer print outs for tests with a frequency of quarterly or less.
Independent reviews of the computer print out test control lists were done by a station engineer.
The test control lists were reviewed and updated to remove miscellaneous tests.
These are tracked by outage schedules and procedure coverage.
.,
.,.,
I
.
/
/