IR 05000315/1982004
| ML20054B123 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 03/31/1982 |
| From: | Dubry N, Hayes D, Swanson E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20054B118 | List: |
| References | |
| 50-315-82-04, 50-315-82-4, 50-316-82-04, 50-316-82-4, NUDOCS 8204160334 | |
| Download: ML20054B123 (13) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Reports No. 50-315/82-04(DPRP); 50-316/82-04(DPRP)
Docket Nos. 50-315; 50-316 Licenses No. DPR-58; DPR-74 Licensee: American Electric Power Service Corporation Indiana and Michigan Power Company 2 Broadway New York, NY 10004 Facility Name: Donald C. Cook Nuclear Power Plants, Units 1 and 2 Inspection At: Donald C. Cook Site, Bridgman, MI Inspection Conducted: February 1 through March 13, 1982 Inspectors: '.R.
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N. E. DuBr
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Approved By: D a
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Reactor Projec s Section ib
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Inspection Summary Inspection on February 1 through March 13, 1982 (Reports No. 50-315/82-04(DPRP);
50-316/82-04(DPRP))
Areas Inspected: Routine, onsite regular and backshift inspection conducted by two resident inspectors. Areas inspected included: Followup on Previous Inspection Findings, Operational Safety Verification, Inspection During long Term Shutdown, Surveillance Observation, Maintenance Observations, Onsite Review Committee, Unit 1 Blackout, Plant Trips, Licensee Event Reports Followup, NUREG-0737 Task Action Plan, and Maintenance Outage. The inspection involved a total of 299 inspector-hour onsite by two NRC inspectors including 58 inspector-hours onsite during off-suift hours.
Results: Of the ten areas inspected no items of noncompliance or deviations were identified in eight areas. Three items of noncompliance were identified in two areas.
(Failure to comply with test procedures during hydrostatic testing, Paragraph 5; failure to assure proper document control, Paragraph 5; and failure to maintain adequate control, documentation, and evaluation of test equipment procurement and calibration, Paragraph 6.)
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B204160334 820401 PDR ADOCK 05000315 G
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DETAILS 1.
Persons Contacted
- W.
Smith, Jr., Plant Manager
- B.
Svensson, Assistant Plant Manager
- E. Townley, Assistant Plant Manager
- E.
Smarrella, Technical Superintendent
- J.
Stietzel, Quality Assurance Supervisor
- K. Baker, Operations Superintendent R. Dudding, Maintenance Superintendent A. Blind, Performance Supervising Engineer T. Kriesel, Environmental Supervisor R. Begor, Staff Assistant J. Wojcik, Chemistry Supervisor
- Denotes those present at exit interviews.
The inspectors also contacted a number of licensee and contract employees and informally interviewed operations, technical, and main-tenance personnel during this inspection period.
2.
Followup on Previous Inspection Findings
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(Closed) Noncompliance item (315/81-03-04; 316/81-03-03):
Reviews and revisions of plant procedures were not properly controlled or conducteJ. The inspector reviewed the corrective actions being taken by the licensee and found the following:
a.
A brief summary explaining cancellations has been incorporated into effected procedures and instructions.
b.
Instruction and procedure reviews are being done in the opera-ti.ons department. The tickler control system, referred to in
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the licensee's June 26, 1981, letter (AEP:NRC:00576), has been set up but the inspector was informed it will not be activated until May 1982.
c.
The June 26, 1981, letter stated that review of all OHP 4024.XXX.XXX series was expected to be complete by August 30, 1981. Also stated in the same letter was that the remainder of procedures l
past due would be reviewed by December 31, 1981. During this i
review effort the inspector noted that most of the work had been
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accomplished on and by the site, with the following exception.
Procedures (OHP 4024 series) committed to have been reviewed by August 30, 1981, are still outstanding. These procedures
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l are being initially reviewed by AEPSC. The inspection
Refer to IE Inspection Reports No. 50-315/81-03(DPRP); 50-316/81-03(DPRP);
50-315/81-05(DPRP); 50-316/81-05(DPRP)
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revealed that AEPSC did not start work in this area until September 1981, which was after the commitment date of August 30, 1981. The inspector found that AEPSC had been informed by the site that unless review efforts being done in New York were expedited the December 31, 1981, revised commitment date to review all past due procedures would also not be met.
In a January 25, 1982, letter (AEP:NRC:0625A)
A1:PSC extended their commitment date to May 1, 1982, to ccoplete the annunciator procedure reviews.
Tb inspector found that Procedure No. OHP 4024.202.XXX did not exist.
Subsequent to informing plant personnel this, the missing procedure, was written by AEPSC and forwarded to the site by the end of the reporting period.
In addition the inspector also reviewed the status of maintenance and technical department procedures (refer to IE Inspection Reports No. 50-315/81-03(DPRP); 50-316/81-03(DPRP); 50-315/81-05(DPRP), and 50-316/81-05(DPRP)) to ascertain the status of reviews, tickler systems, and incorporation into the review process per the AEPSC letter of January 25, 1982.
The above noncompliance is being closed. However, the annunciator review will remain an open item pending the licensee's fulfillment of the May 1, 1982, commitment.
(315/82-04-01; 316/82-04-01)
(Closed) Noncompliance item (315/81-03-03): Failure to submit a special report whithin the required time. The special report (SI-15)
was submitted on February 6, 1981. This report adequately described the sequence of events and analyzed the occurrence of October 11, 1980.
A tickler system has been set up by the licensee and future submittals of special reports will utilize the licensee event report format to prevent recurrence of this nature.
(Closed) Noncompliiance item (50-315/81-01-02): Failure to report as required. A 10 CFR 50.72 report was not made when the plant entered an action statement requiring plant shutdown. Operations Department Supervisors have demonstrated an understanding of report requirements.
(Closed) Noncompliance items (50-315/81-05-02; 50-316/81-05-04):
Failure to report as required. A 10 CFR 50.72 report was not made for an unplanned radioactive release.
Recent reports reflect diligence in t
ascertaining reportability thereby demonstrating the effectiveness of the licensee's corrective actions.
(Closed) Noncompliance item (50-315/81-13-02): Failure to report as required. A licensee Event Report 50-315/81-004/03L-0 was submitted on April 9, 1981, and the requirements for initiating and upgrading condition reports were discussed with the personnel involved.
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3.
Maintenance Outages Unit 1 On January 31, 1982, a Unit I high vibration turbine trip caused a reactor trip. The turbine trip was a result of a blade / bucket loss on the first stage of the high pressure turbine. This resulted in a five week maintenance outage while repairs were made to the turbine.
During this time the licensee also conducted other surveillance and maintenance activities associated withthe nuclear plant as detailed throughout this report. The reactor was started up on March 3, 1982, and tied to the grid on March 4, 1982.
Unit 2 On March 10, 1982, Unit 2 was reduced in power from 100% in an attempt to control increasing temperatures on No. 23 Reactor Coolant Pump (RCP) Motor. On March 11, 1982, Unit 2 was removed from service and conducted a normal shutdown from 75% power because of increasing No. 23 RCP motor temperatures and evidence of primary system leakage from the No. 23 RCP seals. Due to the shutdown a primary coolant dose equivalent iodine-131 concentration greater than the technical specification limits of 1 pCi/gm was experienced from about 3:00 p.m.
EST on March 11, 1982, to 5:00 a.m. EST on March 12, 1982. During l
this time period, while doing degassing operations, an unplanned l
gaseous release occurred. The peak release was 41.4% of the instan-taneous limits and an average of 2.8% of the integrated release limits.
The integrated release was calculated from 12:00 noon on March 11, 1982, to 5:00 a.m. EST on March 12, 1982, with the total release of 94.7 Ci noble gas.
This outage is predicted to be two weeks with repair of the No. 23 RCP and Ice Condenser surveillances being the major planned work.
These activities will be covered in a subsequent report.
4.
Operational Safety Verification-Inspection During Long Term Shutdown The inspectors observed control room operations, reviewed applicable i
logs and conducted discussions with control room operators during the period of February 1 through March 13, 1982. The inspectors verified the operability of selected emergency systems, reviewed tagout records, and verified proper return to service of affected components. The inspectors also reviewed surveillance tests required during the Unit 1 shutdown, verified tagouts and records used during the outage, and checked applicable containment integrity requirements. Tours of the Unit I containment, auxiliary building, turbine building, and screen-house accessible areas, including exterior areas were made to assess equipment conditions, plant conditions, and radiological controls and safety. Verification that maintenance requests had been initiated for equipment in need of maintenance was also made.
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The inspector observed plant housekeeping / cleanliness conditions including potential fire hazards and fluid leaks; and, verified implementation of radiation protection controls. During the inspec-tion period the inspector walked down accessible portions of the Unit 1 Safety Injection accummulator system, the high head injection of the Unit 1 ECCS, and the Unit 1 and 2 Essential service water and component cooling water systems to verify operability.
By observation and interview the inspectors verified that the station security plan had been implemented and also witnessed portions of the radioactive waste system controls associated with radwaste processing.
These reviews and observations were conducted to verify that facility operations and maintenance efforts were in conformance with the re-
quirements established under Technical Specifications, 10 CFR, 49 CFR, and administrative procedures.
5.
Maintenance Observations Station maintenance activities of safety related systems and com-ponents listed below were observed / reviewed to ascertain that they
were conducted in accordance with approved procedures, regulatory guides, and industry codes or standards, and in conformance with Technical Specifications.
The following items were considered during this review:
the limiting conditions for operation were met while the components or systems were removed from service; approvals were obtained before initiating the work; activities were accomplished using approved procedures and were inspected as applicable; functional testing and/or calibrations were performed before returning components or systems to service; quality control records were maintained; activities were accomplished by qualified personnel; parts and materials used were certified; and radiological and fire prevention controls were implemented.
The following maintenance activities were observed / reviewed:
Unit 1 12 THP 6030 IMP.085 Repair and calibration of ILA-131 Accumulator Level Transmitter
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1 MHP 5021.003.001 Maintenance and Repair of the "E" Centrifugal Charging Pump Mechanical Seal RFC 12-2497 Installation of Emergency Leak Off Line For The Unit 1 Centrifugal Charging Pumps 12 MHP 5050 SPC.005 Hydro of The "W" Centrifugal Charging Pump Emergency Leak Off Line 12 MHP 5050 SPC.005 Replacement and Hydro of ESW Outlet Piping From the "E" Component Cooling Water Heat Exchanger
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Following the completion of maintenance / design change on the Unit 2 ESW system and the Unit 1 Centrifugal Charging System the inspector noted two items of noncompliance.
The first item was failure to comply with Technical Specifica-tions 6.8.1 in that the setpoints and equipment delineated in the Hydrostatic Test Procedure 12 MHP 5050 SPC.005, were not adhered to.
The inspector observed that a safety relief valve was not employed in the test rig and that the test pressure had been exceeded by 25%
on the Essential Service Water System before the operators referred to the test procedure for the required test pressure.
(316/82-04-02)
The second item was the failure to assure proper document control per 10 CFR 50, Appendix B, Criterion VI, and D. C. Cook Procedure No. PMI 5040, for the change to the Unit 1 centrifugal charging system (RFC 12-2497). The licensee failed to update the system "0P" drawings in a timely manner following the partial completion of the RFC. The inspector observed that the Unit 1 portion of this RFC was completed in mid February 1982, and that some of the valves effected by this change were unlabled and others moved to different system and physical locations. This resulted in confusion to the operators conducting the Emergency Core Cooling System valve alignment start up checks on March 2, 1982, because neither they nor the control room had revised
"0P" drawings reflecting the "as-built" system. The licensee returned Unit 1 to operating status on March 3, 1982, even though the "0P" drawings were still not available to the operators. The inspector noted that the licensee had revised "0P" drawings placed in the cont-ol room on March 5, 1982. Although the system was properly aligned, this is a concern since the use of out-of-date operating drawings has the potential to result in an actual occurrence.
(315/82-04-02)
The two events discussed above are violations as set forth in the Appendix.
6.
Surveillance Observations The inspectors observed Technical Specifications required surveillance testing " Ice condenser Lower Inlet Doors" (Unit 1), ** 12 THP 4030 STP.207, and verified that testing was performed in accordance with adequate procedures, that limiting conditions for operation were met, that removal and restoration of the affected components were accomplished, that test results conformed with Technical Specifications and procedure require-ments and were reviewed by personnel other than the individual directing the test.
The inspectors also noted that any deficiencies identified during the testing were properly reviewed and resolved by appropriate management personnel.
The inspectors also witnessed / reviewed portions of the following test activities:
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Unit 1 12 THP 6010 RAD.590 Calibration of 1R-33 Gland Seal Exhaust Vent Monitor 1 OHP 4030 STP.005 Emergency Core Cooling System Operability Test 12 MHP 4030 STP.023 250 Volt "N" Train Batteries 12 THP SP.034 CO System Test
Unit 2 2 THP 6030 IMP.252 Feed Flow Transmitter Protection Set I 12 THP 6040 PER.020 Thermal Power Measurement Calorimetric 2 OHP 4030 STP.029 Reactor Thermal Power Determination-Short Form 12 THP SP.034 CO System Test
During the inspection effort in this area the inspector identified an area of concern which has to do with the qualification of test equipment used to confirm technical specification adherence and the safe operation of the plant. The regulations state in part"...ac-tivities affecting quality shall be prescribed by documented instructions, procedures, or drawings and shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been accomplished." The regulations also state that measures shall be established to assure that purchased material, equipment, and services shall be assessed for control of quality to be consistant with the licensee's program. The licensee's internal procedure (PMI 3010), Plant Procurement Control states in part..."that originators of purchase requisitions for materials,
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equipment, and/or services which are for nuclear grade application are responsible for selecting only those companies which are on the qualified suppliers list (QSL).
It was noted that the test equipment used to check the calibration of components used in the Thermal Power Measurement, Ice Condenoer Door surveillance, and containment purge rates were either purchased from or serviced by vendors not on the QSL.
Engineering / Performance Test Procedure (PMI 6040) directs that test equipment shall be identified and controls established to clearly identify the calibration status of the test equipment.
Procedure No. 12 THP 6030 IMP.001, " Test Equipment Calibration Program for the D. C. Cook Nuclear Plant" states in part"...that if the standard is found to be out of calibration during a periodic test a documented evaluation (as-found-data), will be made of the validity of all plant equipment tests performed with this equipment since its last acceptable calibration." The procedure is also explicit as to the handling of test equipment when it is found to be out of calibration more than once in succession.
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Contrary to the above, it was noted by the inspector that some test equipment being calibrated (serviced) by offsite vendors had no
"as-found" documentation or information as to whether repairs or adjustments were made to the equipment during its calibration.
It is therefore impossible for the licensee to make the evaluations discussed above.
The items discussed above are set forth in the Appendix as violations of the regulations and plant procedures and are combined as one item because they are related.
(50-315/82-04-03; 50-316/82-04-03)
7.
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 accomplished, and corrective action to prevent recurrence had been accomplished in accordance with technical specifications.
Unit 1 RO 79-055/03X-1 Penetration Fire Barrier Seals Inoperable (Revision)
R0 81-001/03X-1 Boron Injection Tank Concentration low (Revision)
R0 81-006/03L-0 Ground on IMO-910, In Flowpath from RWST to Centrifugal Charging Pumps R0 81-006/03X-1 Ground on IM0-910, In Flowpath from RWST to Centrifugal Charging Pumps (Revision)
R0 81-026/99X-1 RCP Shim Found Missing (Revision)
R0 81-033/01T-0 Seismic Inadequacies in Switchgear Cabinets R0 81-036/03L-0 Inoperable Emergency Diesel-Empty Day Tank R0 81-037/03L-0 Channel 4 Power Range OP and OT Delta T Tripped R0 81-038/03L-0 Diesel Generator Inoperable to Replace Fuel Rack Taper Pin R0 81-042/04L-0 SG Blowdown With Blowdown Sampling Isolated R0 81-043/03L-0 Containment Pressure Hi-Hi Actuation Logic Failed R0 81-049/03L-0 Loss of Alternate Reserve 69/4 KV Source R0 81-050/03L-0 Acceleration Recorder Removed From It's Concrete Pad R0 81-054/03L-0 OP-0T Delta T Channel 3 Failure RO 81-060/03L-0 RCS Temperature Reading Low Due To Low Current Condition R0 81-062/03L-0 Excessive RCS Leak Rate R0 81-063/03L-0 RTD Failure R0 81-064/03L-0 NESW Return Isolation Valve Inoperable R0 81-065/03L-0 R-11 and R-12 Inoperable R0 82-004/03L-0 Seismic Monitor Inoperable 80 82-005/03L-0 650 Ft. Airlock Failed Leak Rate Test RO 82-006/03L-0 Inoperable ESW Valve
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Unit 2 RO 81-XX-05 Inadvertent SI During Mode 4 Solid State Protection System Testing R0 81-037/03X-1 Containment Pressure Hi-Hi Setpoint Drift (Revision)
R0 81-039/01T-0 Seismic Inadequacies in Switchgear Cabinets R0 81-045/03L-0 Diesel Generator Inoperable to Replace Fuel Rack Taper Pin RO 81-046/03L-0 Containment Pressure High Bistable Drifted High R0 81-047/03L-0 Surveillances Not Done On Redundant RWST Level R0 81-048/01T-0 Charging Check Valve Bonnet Studs Degraded RO 81-056/03L-0 Missed Surveillance on N-Train Battery R0 81-065/04L-0 High Condenser Delta-T R0 81-068/03L-0 S/G Stopvalve Dump Valve Inoperable R0 81-069/03L-0 S/G Stopvalve Dump Valve Inoperable R0 81-070/03L-0 S/G Stopvalve Dump Valve Inoperable R0 81-071/03L-0 Power Supply Breaker to APDMS Failed R0 81-073/03L-0 Reactor Coolant System Temperature And Pressure Reading Low Due To Low Current Condition R0 81-074/03L-0 Solid State Protection System Logic Card Failure R0 82-002/01T-0 Essential Service Water Valve Surveillance Missed R0 82-006/03L-0 R-11 And R-12 Inoperable (Closed) LER (50-316/81-058/03L-0):
The normal letdown relief was lifting 370 psig below the setpoint. When reviewing this licensee event it was noted that the cause description did not agree with the valve repair data.
Rather than a loose lock nut, the valve had a ruptured bellows and pitted valve disc which needed replacement.
This was brought to the licensee's attention and he has committed to issuing a revised event report.
(316/82-04-04)
8.
Unit 1 Blackout At 9:19 a.m. on February 18, 1982, a " blackout" of Unit 1 occurred.
While in backfeed, the "K" breaker opened when the disconnects for
"K-1" was opened.
The Diesel Generators started and were manually loaded.
Station loads were then transferred manually to normal reserve transformers. The cause of the event was found to be an improperly wired socket in the Relay Rack Power Supply. The socket wiring has been corrected and the licensee is confirming that this has corrected the cause. The unit was shutdown for maintenance at the time of the event.
9.
Plant Trips Following the Unit 2 reactor trips on loss of vacuum from 89% percent power on February 22, 1982, the inspector determined the status of the reactor and safety systems by observation of the control room indicators and discussions with licensee personnel concerning para-meter and emergency system status.
Based on past experience it was expected that the primary system dose equivalent iodine Technical Specification of 1 pCi/gm would be exceeded.
The primary coolant went to a 2.33 pCi/gm peak Dose Equivalent Iodine and the licensee
was able to return to within Technical Specification limits by an accelerated RCS cleanup effort. There was no ECCS actuation, all systems functioned normally and there were no associated releases.
The inspector verified the establishment of proper communications and reviewed corrective action taken by the licensee. The plant was returned to operation on February 23, 1982.
The Unit I tripped on March 3, 1982, as a result of the No. 2 turbine stop valve being manually closed while turbine impulse chamber indicated 10% turbine power during a startup. There was no ECCS actuation, all systems functioned normally, and there were no releases. The licensee commenced another startup and the reactor was taken critical at 1910 and entered Mode 1 (power operation) on March 4, 1982, at 12:45 a.m. EST.
The Unit I tripped on March 5, 1982, from a low-low No. 1 Steam Generator level caused by a problem with the Turbine Control System.
The inspector ascertained the status of the reactor and safety systems by discussions with licensee personnel concerning plant para-meters, emergency system status and reactor coolant chemiatc; The inspector reviewed the licensee's actions and is following the licensee's efforts to identify the cause of the trip and the sequence of events associated with the trip.
(50-315/82-04-06). The Unit was returned to operation on March 5, 1982, at 6:38 p.m. EST.
10.
Onsite Review Committee The inspector examined the onsite review functions conducted during the month of February to verify conformance with technical specifications and other regulatory requirements. This review included:
review group membership and qualifications, meeting frequency, quorum requirements, committee activities including reviews of proposed technical specification changes, noncompliance items, corrective action, proposed facility and procedure changes, proposed tests and experiments conducted per 10 CFR 50.59, and others required by technical specifications.
Plant Manager Instruction PMI 1040, " Plant Nuclear Safety Review Committee" was issued July 28, 1981, as a result of noncompliance items contained in Reports No. 50-315/81-03(DPRP); 50-316/81-03(DPRP).
The new instruction sets up two subcommittees to perform detailed reviews of event reports and design changes prior to full committee review.
A revised safety committee review checklist is being prepared and will be reviewed as well as the functioning of the two subcommittees in future inspections.
(50-315/82-04-05; 50-316/82-04-05)
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11.
NUREG-0737 TASK ACTION PLAN Item I.C.1.2:
Inadequate Core Cooling Procedures The Westinghouse Owner's Group has corresponded with NRR on this item and delayed implementation schedule is anticipated. The licensee has not yet approved any further changes to their procedures to meet the January 1, 1982, recommendations. This item will be reviewed further when additional guidance is available.
Item I.C.1.3:
Transient and Accident Procedures The status of this item is the same as that described for Item I.C.1.2 above.
Item II.B.1:
Reactor Coolant System Vents Physical work on Design Change RFC-DCC-12-2462 has been verified complete on both units. Procedures have been drafted following the Westinghouse guidelines, but they and the system design have not yet been approved for use by NRR. This item remains open.
Item II.B.3.2:
Post Accident Sampling The licensee's letter to NRR dated December 23, 1981, (AEP:NRC:0652)
requested an extension on this item until 30 days after the 1982 refuelings. Due to the unscheduled Unit 1 outage in February, the Design Change (RFC-DCC-12-2465) is nearly complete on this Unit.
This item remains open.
Item II.F.1.1:
Noble Gas Monitor The Licensee's letter to NRR dated December 23, 1981, requests an extension for completion until 30 days after the end of the 1982 refuelings. The licensee's letter to NRR dated July 10, 1981, describes the system and how it deviates from the recommendations.
This monitor is to be part of Design Change RFC-DCC-12-2448 and will be reviewed when Phase 1 of the change is complete, and NRR has accepted the system design.
Item II.F.1.2:
Iodine / Particulate Sampling The licensee has requested an extension for this item in their January 8, 1981, letter (AEP:NRC:0398) to NRR. This item will be reviewed after the 1982 refuelings.
Item II.F.1.3:
Containment High-Range Monitor This item is also part of Design Change RFC-DCC-12-2448 which is expected to be complete after the 1982 refuelings. NRR is to provide an evaluation of the design before any further inspection is conducted.
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Item II.E.4.2.7:
Radiation Closure Signal to Purge Valves The licensee has not completed the safety grade instrument modifica-tion (RFC-DCC-12-2578) committed to in their March 25, 1980, letter (AEP:NRC:00295B) to NRR. This item is undergoing further review by NRR. Other areas relating to Containment Purge were addressed in an evaluation by NRR dated October 31, 1981, and were found acceptable (II.E.4.2.1.6).
The licensee indicated in subsequent correspondence (January 15, 1982, (AEP:NRC:0370A) that they are evaluating the possible replacement of the existing purge valves with those not utilizing resilient seats. Status of the valve replacement program and revised Technical Specifications will be available by September 1, 1982.
Item II.F.1.5:
Containment Water Level Monitor Design Change designated RFC-DCC-12-2451 was installed during the 1981 refueling outages and tested in July and April 1981, for Units 1 and 2 respectively. The system was described in a letter to NRR on December 23, 1981, (AEP:NRC.0628). The inspector reviewed the instal-lation records, approvals, and testing procedure.
It was noted that the ITT Barton Transmitters were purchased qualified to IEEE-323-1971 and that ITT Barton has a program which will qualify these transmitters to the IEEE-323-1974. The inspector found the system to be installed as described.
Item II.F.1.6:
Contairment Hydrogen Work is progressing on this Design Change (RFC-DCC-12-2463). An extension was requested of NRR in a December 23, 1981, letter until the end of the 1982 refuelings.
Item II.F.2.3:
Reactor Vessel Level Instrument This Design Change (RFC-DCC-12-2444) has been physically installed, but several problems in achieving operability are being resolved with Westinghouse. An extension of the completion date for this item has also been requested of NRR in their December 23, 1981, letter.
Item II.K.3.1:
Automatic Pressurizer Power Operated Relief Valve (PORV) Isolation The licensee reported in their December 29, 1981, letter (AEP:NRC.0649) that the Westinghouse Owner's Group has evaluated this recommendation.
Based on the owners group report, WCAP-9804, the license stated "that the installation of an automatic PORV isolation system is not necessary." This issue is to be reviewed further by NRR.
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Item II.F.1.4:
Post Accident Monitoring of Containment Pressure.
The licensee committed to increasing the span of the upper contain-ment pressure transmitters in their October 24, 1979, letter (AEP:NRC:0253). Design Change RFC-DCC-12-2450 respanned the upper containment pressure instruments (PPA-310 and PPA-312) to a range of -5 to +36 psig, and the lower containment pressure instruments (PPP-300, PPP-301, PPP-302 and PPP-303) to a range of -5 to +12 psig.
The change was reviewed and found to have been implemented utilizing the appropriate controls, procedures, reviews, approvals and testing.
This modification appears to meet the intent of the recommendations in NUREG-0737 and is considered closed.
Permanent scales for the control room recorders are not yet installed, but it was verified that they are on order.
Item II.K.3.5:
Automatic Reactor Coolant Pump Trip The licensee's December 29, 1931, letter (AEP:NRC:0649) to NRR discusses the Westinghouse analysis (WCAP-9584) LOFT predictions, and simulator data which support their position"...that automatic reactor coolant pump trip is not necessary since sufficient time is available for manual tripping of the pumps." This item remains open pending NRR review.
Item II.K 3.10: Anticipatory Trip Modifications As discussed in the licensee's June 20, 1980, letter and NRR's letter of September 8, 1981, no change in the setpoint for the reactor trip or turbine trip is anticipated. This item is closed.
Item II.K.3.12:
Confirmation of Anticipatory Trip on Turbine Trip The licensee's June 20, 1980, letter confirms the existence of this trip signal. A turbine trip at power levels greater than approximately 10% cause a reactor trip. This item is closed.
Item III.D.3.4:
Control Room Habitability The licensee's response to this item dated February 9, 1981, (AEP:NRC:0398C) described the existing system and proposed five modifications. NRR reviewed the submittal and concluded in their February 11, 1982, Safety Evaluation that the system design is acceptable. The modifications are scheduled to be completed by January 1, 1983, at which time they will be inspected.
12.
Exit Interview
The inspectors met with licensee representatives (denoted in Paragraph 1)
throughout the inspection period and at the conclusion of the inspection and summarized the scope and findings of the inspection activities. The licensee acknowledged the items of noncompliance (Paragraphs 5 and 6) and committed to revise the licensee event Report No. 50-316/058/03L-0 (Paragraph 7).
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