IR 05000298/1981023
| ML20040B604 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 12/28/1981 |
| From: | Dubois D, Westerman T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20040B597 | List: |
| References | |
| 50-298-81-23, NUDOCS 8201260235 | |
| Download: ML20040B604 (7) | |
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APPEf4 DIX U.S. NUCLEAR REGULATORY COMMISSION REGI0f4 IV Report fio.
50-298/81-23 Docket No. 50-298 License No. DPR-46 Licensee:
Net,raska Public Power District P. O. Box 499 Columbus, Neoraska 68601 Facility Name: Cooper Nuclear Station Inspection At: Cooper Nuclear Station, Nemaha County, Nebraska Inspection Conducted: November 1-30, 1981 bY d-3
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D. L. DuBois, Resident Reactor Inspector Date Projects Section 1 Approved by:/
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T. F. Westerman, Chip, Projects Section 1 Date Inspection Summary Inspection on November 1-30, 1981 (Report No. 50-298/81-23)
Areas Inspected:
Routine, announced inspection of operational safety verifi-
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cition; montTili equipment surveillance; plant design changes; reactor scram review; follow up to licensee events, previously identified items, TMI action plan requirements and independent inspection.
This inspection involved 112 inspector hours on site by two NRC inspectors.
Results: Within the areas inspected no violations or deviations were identified.
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DETAILS g
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Persons Contacted
- l lessor, Plant Superintendent P. Thomason, Acting Operations Supervisor J. Sayer, C & HP Supervisor B. Gilbert, Training Coordinator D. Majeres, Maintenance Planner L. Bednar, Electrical Engineer R. Peterson, Reactor Engineer
- Indicates presence at exit meetings.
2.
Operational Safety Verification
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The NRC inspectors observed control room operations, instrumentation, f
controls, reviewed applicable logs, and conducted discussions with l
control room operators.
The inspectors verified operability of:
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'A' Residual Heat Removal System
'B' Residual Heat Removal System
'A' Core Spray System
'B' Core Spray System
- 1 Diesel Generator
- 2 Diesel Generator Standby Liquid Control System The inspectors reviewed safety clearance records, including verification that affected components were removed from and returneo to service in a correct and approved manner, redundant equipment was verified operable, i
and limiting conditions for operation were adequately identified and
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maintained. The inspe<.. tors also verified that maintenance requests had been initiated for eqt inm: ! discovered to require repair or routine l
preventative upkeep, appropriate priority was assigned, and maintenance commenced in a timely manner commensurate with assigned priorities.
Tours of accessible areas of the facility were conducted to observe normal security practices, plant and equipment conditions including cleanliness, radiological controls, fire suppression systems, emergency equipment, potential fire hazards, fluid leaks, excessive vibration and i
instrumentation adequacy.
These reviews and observations were conducted to verify that facility operations were in conformance with the requirements established in the Technical Specifications,10 CFR and Administrative Procedures.
No violations or deviations were identified in these areas.
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3.
Surveillance Observations The inspector observed portions of Technical Specification required surveillance tests to verify that testing was performed in accordance with adequate procedures, test instrumentation was in calibration, Limiting Conditions for Operations were met, removal and subsequent i
restoration of affected com;,onents was accomplished, test results confonned with Technical Specification and procedure requirements, tests were reviewed by personnel other than the person directing the test, and deficiencies identified during testing were properly reviewed and resolved by appropriate management personnel.
These reviews and observations were conducted to verify that facility operations were in conformance with the requirements established in the Technical Specifications, 10 CFR and Administrative Procedures.
No violations or deviations were identified in these areas.
4.
Design Changes and Modifications The inspectors reviewed selected design changes to verify that they had been reviewed and approved in accordance with 10 CFR 50.59, the licensee's Technical Specifications, and the licensee's QA Program requirements.
The inspectors also verified that the selected der gn changes were controlled i
by approved procedures, subjected to post modification testing, and given final review and approval by licensee management.
The inspectors further verified that procedures and drawings affected by the design changes had been revised accordingly.
Design changes which do not require NRC approval are referred to as Minor Design Changes (MDCs) by the licensee.
The MDC packages which were selected for review were:
MDC 80-064 Replace PC 56 and 57 with MOVs MDC 81-003 Modification of RCIC Turbine Trip and Reset Logic MDC 81-078 Removal and Plugging of MSIV Stem Leakoff Lines 5.
Plant Scram -- Safety System Challenge The inspector reviewed records and interviewed plant personnel concerning an unscheduled reactor scram which occurred November 6,1981, at 6:09 A.M.,
(Scram Report #81-4). The reactor had been critical and pressurized to 150 psig for performance of safety-relief valve testing. During testing of safety-relief valve S/R 710, the valve failed to reclose following an opening exercise. Reactor water level control was in manual providing makeup via the condensate system.
A flange connection in the makeup flow path loosened during this period and dumped makeup to the 'Q' turbine building sump.thut initiating a condition which caused a reduction of makeup flow to the reactor vessel. The reactor operator began reactor shutdown. Other operators attempted to isolate the leakage and establish another makeup flow path but failed to succeed in time to prevent a low reactor vessel level scram.
Appropriate corrective action was taken.
The reactor remained shutdown pending repair of the affected flange and safety-relief valve.
No unreviewed safety questions were discovered.
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l Safety-relief valve S/R 71D is a two-stage valve rcnufactured by Target Rock Corporation.
It is not an assigned Automatic Depressurization
System (ADS) valve.
Repair involved replacement of its associated remote operated solenoid actuator.
The pressure re?ief feature of this safety-relief valve was not affected by this failure.
Particulars concerning the solenoid actuator failure will be provided in a subsequent report.
6.
Licensee Event Follow Up
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The following LERs are closed on the basis of the inspector's in-office review, review of liccasee documentation, and discussions with licensee personnel:
LER 81-04 LER 81-05 LER 81-08 LER 81-10 l
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NUREG - 0737 Clarification of TMI Action Plan Requirements (a) Item II.B.4, paragraph 2.b (Closed) Training for Mitigating Core Damage The NRC inspector verified licensee actions with regard to this item based upon the requirements set forth in NUREG-0737, H. R. Denton's letter to all power reactor licensees, dated March 28, 1980, and NUREG-0660.
The inspector reviewed the training program for adequacy and applicability.
The program emphasis dealt with symptoms recognition and methods concerning the control and mitigation of accidents involving a degraded core. Applicable systems, components, and instrumentation review was also provided which
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General Electric representatives f
i provided instruction for the first two training sessions.
Subsequent i
instruction was provided by licensed senior reactor operators who attended an earlier training session. Training was completed November, 1981.
Training records were reviewed by the inspector to ensure plant managers, licensed operators, STAS, chemistry and health physics, and instrument
and control personnel attended and received training commensurate with
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their responsibilities during accident and post accident conditions.
The licensee's actions in response to this item appear adequate.
(b) Item II.E.4.1, paragraph 2 (Closed) Dedicated Hydrogen Penetrations The NRC inspector verified licensee committments with regard to this item I
based upon their letters from Pilant to Eisenhut dated June 30, 1981, and Pilant to Eisenhut dated December 30, 1980.
J The licensee completed Minor Design Change (MDC)80-064 during April,1981.
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l The change involved replacing manual hand-operated valves, located in the I
two inch bypass lines around the main containment and torus vent exhaust valves, with motor operated valves (MOVs). The newly installed valves,
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designated PC-MOV-305 and 306, control circuits were modified such that the MOVs would close on a containment isolation signal and remain closed
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- as long as the isolation signal was present.
Key switches were installed which can permit override of the isolation signal, thus permitting MOVs 305 and 306 to be opened should containment venting be required for combustible gas control.
The inspector reviewed the following documents to verify revisions necessitated by MDC 80-064 were incorporated:
Prints and drawings listed in Document Change Notices81-153 through 81-171 Procedures 2.2.60, 5.3.7, 6.2.1.5.2, and 6.4.8.3 Technical Specifications Amendment No. 75, pages 169 and 170 MDC 80-064 was reviewed and documented in paragraph 4 of this report.
Applicable licensee personnel have received training relevant to this modification.
The licensee's actions in response to this item appear adequate.
(c) Item II.K. 3, paragraph 13.B (Closed) High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) System Initiation Levels The inspector verified licensee actions with regard to this item based
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upon the following:
(1) General Electric letter from Buchholz to Eisenhut dated October 1, 1980, submitted an evaluation performed by GE on behalf of the BWR Owners Group.
The letter recommended that separation of and changes to the HPCI and RCIC initiating levels not be performed.
(2) NPPD letter from Pilant to Eisenhut dated December 30, 1980, replied to Eisenhut's letter to Pilant dated October 31, 1980.
The reply committed to modification of RCIC automatic reset. The licensee completed Minor Design Change (PDC)81-003 during June, 1981.
The MDC consisted of circuit changes to RCIC steam supply valve M0-131 and the RCIC turbine trip valve reset logic. As a result of the MDC, M0-131 will automatically open and close at low reactor water level and high reactor water level setpoints respectively.
Upon closure, M0-131 resets the trip valve thus enabling the RCIC to automatically restart upon receipt of a low water level signal.
MDC 81-003 was reviewed and documented in paragraph 4 of this report.
Applicable licensee personnel have received training relevant to this modification. The inspector reviewed the following documents to verify revisions necessitated by MDC 81-003 were incorporated:
(1) Prints and drawings listed in Document Change Notices81-195, 81-226 through 234, and 81-241 through 244 (2) Procedures 2.2.67 and 6.3.6.1 The licensee's actions in response to this item appear adequate.
(d) Item III.D.1.1 (Closed)
Integrity of Systems Outside Containment Likely to Contain Radioactive Material for Pressurized Water Reactor and Boiling
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Water Reactor
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This action item was previously closed in NRC Inspection Report 80-16, dated November 14, 1980.
Modification to the outboard Main Steam Isolation Valves (MSIVs) had not been implemented at that time and was designated open item 8016-01.
The following discussion pertains to the l
open item.
The licensee completed Minor Design Change (MDC)81-078 during the Fall 1981, turbine outage.
The change involved removal of the gland seal leakoff lines from the outboard MSIVs with subsequent capping of the
leakoff penetrations located on the valves.
The inspector reviewed and verified that procedure 2.2.56 and as-built drawings were revised as i
applicable to the design change.
MDC 81-078 was reviewed and documented in paragraph 4 of this report.
Applicable licensee personnel have received training elevant to this modification.
The completion of MDC 81-078 closes open item 8016-1.
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MDCs78-022 and 80-037 were completed at earlier dates ana provided similar
leak reduction modifications to numerous valves located in other systems.
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The licensee's actions in response to this item appear adequate.
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8.
Independent Inspection Effort
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(a) The NRC inspector has determined that a potential discrepancy exists between the as-built condition of the control rod withdrawal block instru-mentation system and Technical Specifications applicable to that system.*
There are a total of six Average Power Range Monitor (APRM) instrumentation channels at Cooper Station.
Two APRM Instrumentation block switches, l
located on the reactor control console panel 9-5, are available to block selected APRM inputs to the Reactor Protection System (RPS) and also the
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Rod Withdrawal Block System.
One switch has positions labeled A, C, E, and the other has positions labeled B, D, F which correspor.d to the actual APRM instrumentation channel that will be blocked for RPS purposes.
How-ever, additional contacts on the same switches block APRM channels A, E, i
D, and B, C, F, respectively, for the rod withdrawal blnck function.
As a
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result, it is physically possible, using various position combinations of
the aforementioned block switches, to defeat more than the Technical Specification required minimum number of operable APRM channels / trip systems in the rod withdrawal block system, e.g., placing the first switch to the 'A' position and second switch to the 'D'
position will block two of three APRM inputs to one of two rod withdrawal block sub-system channels, thus not meeting the minimum channel operability require-ment of two, as required by Technical Specifications.*
i Cooper Nuclear Station Safety Analysis Report (SAR), Volume III, Chapter VII,
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Section 5.7.5, entitled " Power Generation Evaluation," states "... because any one of the APRMs can initiate a rod block, this function has a high
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^ Reference -- Cooper Nuclear Station Technical Specifications, lable 3.2.C,
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page 61, entitled, " Control Rod Withdrawal Block Instrumentation," and notes for Table 3.2.C, page 62.
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level of radundancy and satisfies the power generation design basis."
The preceding statement indicates that the Technical Specification may be extremely conservative concerning the minimum channel operability requirement.
This item will remain as an unresolved item (8123-01).
Cooper Nuclear Station SAR, Volume III, Chapter VII, Section 7.4.3.3, entitled " Rod Block Bypasses," states, "The permissible IRM and APRB bypasses (blocks) are arranged in the same way as in the reactor pro-tection system." As indicated above, it appears that the RPS and rod withdrawal block APRM channels block switch assignments do not conform with the SAR requirement.
This item will remain as an unresolved item (8123-02).
(b) The inspector reviewed welder qualifications for the TAD welders.
Records were available indicating qualification in accordance with ASME Seciton IX for all designated qualified welders.
Qualifications were performed by Omaha Nondestructive and Metallurgical Testing, Inc.
Cocper weld procedures PG, PlG, and PIB were used for welder qualification.
Radiographic test results were utilized in accordance with ASME Section IX.
Welders qualification test records were reviewed for technical content and found to be acceptable.
The weld records for the drywell relief valve supports were reviewed.
A review of the records indicated welder qualification to each weld procedure utilized by every welder.
No deviations or violations were identified.
9.
Exit Meetings Exit meetings were conducted at the conclusion of each portion of the inspection.
The Plant Superintendent was informed of the above findings.
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