IR 05000289/1978013
| ML19276H233 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 06/26/1978 |
| From: | Haverkamp D, Sternberg D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML19276H231 | List: |
| References | |
| 50-289-78-13, NUDOCS 7910150861 | |
| Download: ML19276H233 (19) | |
Text
U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT Region I
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) ort No.
50-289/78-13 Docket No.
50-289 License No. DPR-50 Priority Category C
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Licensee:
Metrocolitan Edison Comoany
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P. O. Box 542 Readina Pennsvivania 19603 Facil ity.Name:
Three Mile Island Nuclear Station, Unit 1 Inspection at:
Middletown, Pennsylvania
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Inspection conducted:
June 5-9, 1978 Inspectors:
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d /2. L/77 0. R. naverKamp, Reacter Inspector date signed
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care signec
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ca+e signed
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Approved by: N M
b Q(o 0. M. Sternberg, tin Chief, Reactor
/cate' signed Projects Section No. 1, RO&NS Branch Insoection Summary:
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Insoection on June 5-9,1978 (Recort No. 50-289/78-13)
Areas Insoected:
Routine, unannouncea inspection of plant operations including snif t logs and operating records and facility tour; reactor coolant system de-contamination; 10 CFR Part 21 implementation; and licensee followup actions concerning selected previous inspection findings, licensee events., and IE Bulletins and Circulars.
The inspection involved 42 inspector-hours onsite by one NRC inspector.
Results:
No items of noncompliance were identified.
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DETAILS 1.
Persons Contacted Metrocolitan Edison Comoany
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Mr. T. Acker, Shift Foreman
- Mr. R. Barley, Lead Mechanical Engineer Mr. K. Bryan, Shift Foreman Mr. J. Chwastyk, Shift Supervisor Mr. W. Cotter, Supervisor of Quality Control Mr. R. Dubiel, Supervisor of Radiation Protection and Chemistry Mr. C. Hartman, Lead Electrical Engineer Mr. J. Hilbish, Lead Nuclear Engineer
- Mr. G. Kunder, Superintendent - Technical Support Mr. V. Orlandi, Lead Instrumentation and Controls Engineer Mr. M. Ross, Supervisor of Operations Mr. M. Shatto, PORC Secretary Mr. W. Zewe, Shift Supervisor The inspector also talked with and interviewed several other
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licensee employees during the inspection. They included control room operators, technical and engineering staff personnel and general or,f1ce personnel.
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- denotes those present at the exit interview.
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2.
Licensee Action on Previous Inscection Findinas (Closed) Inspector Followup Item (289/76-27-01):
Training Program for Offsite Agencies.
This items was reviewed and satisfactorily resolved for Unit 2, as described in NRC Inspection Report 50-320/78-09.
The concerns noted in the referenced Unit 1 inspectidn report have been incorporated into facility procedures.
(Closed) Unresolved Item (289/77-09-04):
Snubber Maintenance.
Step 6.1.4 of MP1410'Y-34, Revision 1, contains precautions against
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removing snubbers during all modes of plant operatiens except cold shutdown and refueling.
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(Closed) Inspector Followup Item (289/77-19-01):
Valve Stem and Leak Off Incrustation in Borated Water Systems.
Various borated water system leaks were corrected during the 1978 refueling cutage.
The inspector noted the improved conditions and had no further questions concerning this item.
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(Closed) Inspector Followup Item (289/77-19-02):
Diesel Generator Voltage Control Rneostats. Work request #19516 has been completed which installed an additional rheostat inside each diesel generator local control panel.
(Closed) Inspector Followup Item (2S9/77-19-03):
Corrective actions have been completed for LER's 77-3/3L, 77-4/lT and 77-7/lT.
A grease relief kit was installed on the motor operator for BS-V3B; TS Amendment 28 included the correct value of pressurizer code safety valve relief capacity; and personnel hatch strut bearing and latching bar interlock arm modifications were ccmpleted.
(Closed) Noncompliance (289/77-27-01):
Use of Superseded Proce-dures. The licensee's corrective measures have been completed as described in MEC letter to NRC Region I serial GQL 1375, dated October 14, 1977.
The inspector reviewed selected ocerating and surveillance procedures completed during October 1977 through Fay 1978 and verified that current procedures had been used.
(Closed) Unresolv'ed Item (289/77-29-01):
Status of Open Nonconform-ance Reports (NCR's).
Generation Procedure GP-4012 was revised to include a viable mechanism for follow-up actions on overdue NCR's.
The licensee has reduced the number of NCR's overdue for closecut to 34% of the total open NCR's.
The inspector verified that the licensee's measures to close overdue NCR's have been effective.
The inspector noted that the status of open NCR's and continuing
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licensee practices to ensure prompt correction of adverse condi
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tions will be reviewed during routine NRC quality assurance inspec-tions and had no further questions concerning this matter at this time.
(Closed) Noncompliance (289/77-29-04):
Inoperative Battery Room Fan.
The licensee's corrective measures have been ccmdleted as described in MEC letter to NRC Region I serial GQL 1514, dated
- iovember 11, 1977.
Operating logs were revised to recuire shiftly checks of proper battery room ventilation.
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(Closed) Inspector Followup Item (289/77-31-01):
Corrective actions for LER 77-22/lT, as described in licensee letters dated September
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8 and 23, 1977, have been ccmpleted.
Results of periodic decay heat pump shaft ultrasonic testing and vibration measurements have been satisfactory.
The purp shafts of DH-PlA and DH-PIB were re-placed during the Spring 1978 refueling outage.
The replacement shafts have rounded keyways and the appropriate material certifi-cation and heat treatment documentation and have been satisfactorily tested.
These corrective actions were reported in MEC letter to NRC (NRR) serial GQL 0858, dated May 10, 1-3.
In view of these actions, there is no longer a need to perform periodic ultrasonic testing and vibration measurements.
(Closed) Noncompliance (289/77-32-01):
Control Rod Program Verifi-cation.
The licensee's corrective measures have been completed as
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described in MEC letter to NRC Region I serial GOL 1525, dated November 18, 1977.
Surveillance Procedure 1301-9.2, Revision 10, limits movement of control rods to less than 2 inches during pro-gram verification, in confomance with current Technical Specifi-cation requirements. A Technical Specification Change Request to allow movement of 2% of rod length has been incorporated in the licensee's submittal of revised Technical Specifications in the STi
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(Closed) Unresolved Item (239/77-32-02):
Heat Balan'ce Calcula-
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tions.
A memorandum was sent to shift foremen re-emphasizing that input parameter checks shall be completed just prior to perfoming a heat balance calculation. Procedure 1103-16, Revision 8, clari-fied when the input parameter checks shall be run.
(Closed) Inspector Followup Item (289/77-37-04):
Cautionary Equip-ment Control.
SAP 1037, " Control of Caution Tags" has been issued to formally implement the use of Caution Tags and Do Not Operate Tags.
Refer to paragraph 4.a of this report for additional in-formation concerning this item.
(Closed) Inspector Followup Item (289/77-37-08):
Pemanent Repair of DH-V1.
Decay Heat System isolation valve DH-V1 remains in a condition of temporary repair using Fumanite leak-sealing com-pound, as described in LER 76-40/3L.
Routine visual surveillance and DH system operation during the Spring 1978 refueling outage has indicated no need for reinjection of Furmanite due to no additional 2) }l9
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seal ring gasket leakage. An engineering evaluation performed by
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the licensee concluded that CH-VI would not be permanently repairec during the recent outage, and the facility would be operated for another year with the temporary rer.ir.
Routine visual surveil-lance for increased valve leakage will be continued and permanent
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valve repair will be re-evaluated by the licensee prior to the 1979 refueling outage. The inspector had no further quesitons concern-ing this item.
(0 pen) Noncompliance (289/78-01-01):
Hydraulic Snubtar Surveil-lance. The status of this item was described in NRC Region I Inspection Report 50-289/78-07, paragraph 2, however, an incorrect heading (Unresolved Item 289/77-09-04) was contained in the re-ferenced report. The licensee's corrective measures regarding Noncompliance 289/78-01-01 will be reviewed during a subsecuent inspection.
(0 pen) Inspector Followup Item (289/78-02-01):
Corrective actions for LER 77-25/3L nave not yet been completed.
Installation of permanent piping for seal water tank makeuo to the evaporator was
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deferred due to higher priority work during the recent refueling outage.
This item remains open pending completion of the perament piping installation.
(Closed) Noncompliance (289/78-03-02):
Personnel Monitoring / Rad-waste Container Contamination Determination. The licensee's corrective measures have been completed as described in tiEC letter to NRC Region I Serial GQL 0705, dated April 19, 1978.
Procedure HPP 1612 was changed to allow individuals to move tu low background areas to monitor for contamination.
Responsible HP personnel were counselled to ensure that proper procedures are carried out re-garding movement of radwaste containers.
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(Closed) Unresolved Item 289/78-05-01):
Start Work Approval for Maintenance. Supervisor of Maintenance Memorandum dated April 17, 1978, restated the need for maintenance persor.nel to get Shift Foreman approval on Work Request Forms prior to commencing work.
The inspector had no further questions concerning this item.
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3.
Review of Safety Limits, Limiting Safety System Settines and Limiting Concitions for Oceration The inspector ooserved process instrumentation monitoring current operations on June 8 and 9,1978 and reviewed records of previous reactor operations during March 1977 - May 1978.
The following
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logs and records were reviewed on a sampling basis.
Shift Foreman log
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Control Room Log Book
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Completed Operating Procedures 1102-1, " Plant Heatup to 525'."
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Completed OP's 1102-2, " Plant Startup."
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Completed OP's 1102-10, " Plant Shutdown."
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Completed OP's 1102-11, " Plant Cooldown."
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Completed Surveillance Procedures 1301-1, " Shift and Daily Checks."
Completed SP's 1301-2, " Boric Acid Mix Tank or Reclaimed Scric
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Acid Tank Chemistry."
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Completed SP's 1301-3, " Reactor Coolant System Chemistry."
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Completed SP's 1301-4.1, " Weekly Surveillance Checks."
Completed SP's 1301-4.4., "BWST Chemistry."
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Completed SP's 1301-4.6, " Station Storage Batteries - Weekly."
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Completed SP's 1301-5.1, "BAMT Temp Channels, RBAST Temp
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Channel."
Completed SP's 1301-5.6, " Core Flooding Tank Water Samole."
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Completed SP's 1301-5.8, " Station Batteries - Monthly."
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Completed SP's 1301-6.5, " Sodium Hydroxide Tank Concentration."
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Ccmpleted SP's 1301-6.6, " Sodium Thiosulphate Ta k Concentra-
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tion."
Completed SP's 1301-9.2, " Control Rod Program."
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Completed SP's 1303-1.1, " Reactor Coolant System Leak Rate."
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Completed SP's 1303-3.1, " Control Rod Movement."
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Completed SP's 1303-4.16, " Emergency Power System."
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Completed SP's 1303-5.4, " Emergency Feedwater Pumos."
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The observations and records review were conducted to verify that startup, power and/or shutdown reactor operations were in conformance with Technical Specification safety limits, limiting safety system settings, and limiting conditions for operation.
Acceptance criteria for the above items included selected require-
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ments of facility operating procedures and the following Technical Specifications (listed according to their respective systems).
Reactivity Centrol and Power Distribution Technical Specifi-
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cations 2.1.2, 3.1.3.4, 3.1.3.5, 3.2.1, 3.5.2.1, 3.5.2.4, 3.5.2.5, 3.5.2.6, 4.1.2, 4.7.1, and 4.7.2.
Instrumentation Technical Specifications 2.3.1 (Table 2-3.1
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and Fig. 2.3.2), 3.5.1, 3.5.3, and 4.1.1.
Reactor Coolant System Technical Specifications 2.1.1, 2.2.1,
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3.1.1.1.b, 3.1.1.1.c 3.1.1.2, 3.1.2.1, 3.1.2.2, 3.1.2.3, 3.1.3.1, 3.1.3.2, 3.1.5, 3.1.10, and 4.1.2.
Emergency Core Cooling System Technical Specifications 3.3.1.1,
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3.3.1.2, 3.3.1.3, 3.3.1.4, 3.3.1.5, and 4.1.2.
Containment Systems Technical Specifications 3.6.'1, 3.6.2,
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3.6.3, 3.6.4, 3.6.5, and 3.6.6.
Plant and Electrical Power Systems Technical Soc-4fications
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3.12.3, 3.4.1, 3.1.2, 3.4.3, 3.4.5, 3.7.1, 3.7.3, 1.6.1.a, 2.6.2.a, a.5.2.0, a.5.2.c, 2.9.1.1, and 4.9.1.2.
he 3: ems c:servec and reviewed :y ne insoec Or were acce::able.
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4.
Review of Plant Ooerations a.
Shift Lecs and Oceratino Records The inspector reviewed the following logs and records:
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Shift Foreman Log, Control Room Log Book, Control Room Operator's Log Sheets, Primary Auxiliary Operator's Log-Tour Readings, Primary Auxiliary Operator's Log-Licuid Waste Disposal Panels, Secondary Auxil.iary Operator's Log Sheets, and Auxiliary Operator Log Sheets-Out-Building
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Tour; dated February 23 - May 31, 1978.
Shift and Daily Checks; dated February 21 - May 22,1978.
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Jumoer, Lifted Lead, and Mechanical Modifications Log (active and cleared); entries made during February 23 -
May 31, 1978.
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Equipment Lockout Tag Log (active and cleared); entries made during February 23 - June 5, 1978 and May 14 -June
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5,1978, respectively.
G Do llot Operate Tag Log; entries made during February 23 -
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June 1, 1978.
Transient Cycle Log Book; entries made during February
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23 - May 31, 1978.
Unit 1 Superintendent's Operating Mcmos 77-1 and 78-1.
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Unit 1 Operations Department Memos 78-5 through 78-7.
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The logs and records were reviewed to verify the following items.
Log keeping practices and log book reviews are conducted
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in accordance with established administrativ,e controls.
Log entries involving abnormal conditions are sufficiently
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detailed.
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Operating orders do not conflict with Technical Specifica-
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tions (TSs).
Jumper log and tagging log entries do not conflict with
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TSs.
Jumper / lifted lead / mechanical modification and tagging
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operations are conducted in conformance with established
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administrative controls.
Problem identification reports confirm compliance with
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TS reporting and LCO requirements.
Acceptance criteria for the above review included inspector judgement and requirements of applicable Technical Specifica-tions and the following procedures.
Station Administrative Procedure (SAP) 1002, " Rules for
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the Protection of Employees Working on Electrical and Mechanical Apparatus," Revision 12.
SAP 1010, " Technical Specification Surveillance Program,"
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Revision 12.
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SAP 1011, " Controlled Key Locker Control," Revision 16.
SAP 1012, " Shift Relief and Log Entires," Revision 8.
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SAP 1013, " Bypass of Safety Functions and Jumper Control."
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Revision 7.
SAP 1016, " Operations Surveillance Program," Revision 12.
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SAP 1033, " Operating Memos and Standing Orders," Criginal.
SAP 1037, " Control of Caution Tags," Griginal.
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The inspector's findings regarding shift logs and operating records were acceptable, as noted below.
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SAP 1037, issued March 13, 1978, implements use of yellow
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cardboard type CAUTION tags as an informatien tag only.
The tag is to be attached to a ccrocnent, control switch or other device to indicate an off normal condition or to caution personnel to a specific condition which must be
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satisfied prior to using the component er devica.
Additionally, SAP 1037 provides for use of a red bakelite DO-NOT-0PERATE tag in place of a CAUTION tag, particu-larly when used in environments where the CAUTION tags may easily deteriorate under extended use.
7he admini-strative controls provided by SAP 1037 have not yet been fully implemented, including operator training regarding use of CAUTION tags, and replacement of existing white information stickers and D0-NOT-OPERATE tags with CAUTION tags where applicable.
Licensee representatives stated that it was originally intended to gradually implement the controls of SAP 1037, however, the existing informa-tion stickers and D0-NOT-0PERATE tags would be reviewed for continued applicability and replaced with CAUTION tags, as required. The inspector had no further questions
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concerning this matter at this time, and stated that the
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licensee's inlementation of SAP 1037 would continue to be reviewed during periodic NRC inspections of plant operations.
b.
Plant Tour.
On June 9,1978, the inspector conducted a tour of the following accessible plant areas.
Auxiliary Building
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Turbine Building
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Fuel Handling Building
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Control Room
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Switchgear Rocms
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Inverter and Battery Rooms
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Diesel Rocms
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Chemical Additien Room
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' lake-u: Nm Rcoms
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The following observations / discussions / determinations were made.
Control Room and local monitoring instrumentaticn for
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various components and parameters was observec, including
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reactor power level, Reactor Building pressure, SWST level and temperaeure, diesel fuel oil levels, and control rod positions.
Radiation controls established by the licsr.::a, including
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the posting of radiation and high radiation areas, the
condition of step-off pads and the disposal of protective
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clothing, were observed.
Radiation work permits used for
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entry to radiation and controlled areas were reviewed.
Actual radiation level measurements were observed and compared with posted values throughout the plant.
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Plant housekeecing, including general cleanliness con-ditions and storage of materials and components to prevent safety and fire hazards, was observed.
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Systems and equipment in all areas toured were observed
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for the existence of fluid leaks and abnormal piping vibrations.
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Selected piping snubbers / restraints were observed for
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proper fluid level and condition / proper hanger settings.
The indicated positions of electrical power supply breakers
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and selected control board equipment start switches and remote-operated valves and the actual positions of selected manual-operated valves ware observed.
Selected equipment lockout tags and caution tags were
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observed for prcper posting and the tagged equipment was observed for proper positioning.
The Control Board was observed for annunciators that nor-
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mally should not be lighted during the existing plant conditions.
The reascns for the annunciators were dis-cussed with control room c:erators.
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The licensee's policy and practice regarding plant tours
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was reviewed.
Control Room mannirg was observed on several occassions
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during the inspection.
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Acceptance criteria for the acove items included inspector judgement and requirements of 10 CFR 50.54(k), Regulatory Guide 1.114, applicable Technical Specifications and the following procedures.
SAP 1002. " Rules for the Protection of Employees Working
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on Electrical and Mechanical Apparatus," Revision 12.
SAP 1003, " Radiation Protection Manual," Revision 12.
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SAP 1008, " Good Housekeeping," Revision 3.
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SAP 1009, " Station Crganization and Chain of Command,"
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Revision 3.
SAP 1028, " Operator at the Controls," Original.
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SAP 1037, " Control of Caution Tags," Original.
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The inspector's finding regarding the plant tours were accept-able.
5.
In-Office Review of Licensee Event Recorts (LERs)
The LERs listed below were reviewed in the Region I office prcmptly following receipt to verify that details of the event were clearly reported including the accuracy of the description of cause and the.
adequacy of corrective action.
The LERs were also reviewed to deter-mine whether further information was required from the licensee, whether generic implications were involved, whether the event should be classified as an Abnormal Occurrence, and whether the event war-anted onsite followup.
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The following LERs were reviewed.
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LER 78-01/lT, dated January 24, 1978; Potential error between heat balance and nuclear instruments during some transients.
':en :es : nose irs selected for :nsite folicwuo.
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LER 78-02/lT, dated January 26, 1978; Continuous fire waten
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was not posted when the "B" Diesel Generator Radiator Room
.1 sprinkler system was isolated for maintenance.
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LER 78-v3/3L, dated February 3,1978; Emergency Diesel Genera-
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tar 13 failed to start during surveillance testing due to oil pressure limit switch setpoint drift.
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LER 78-04/3L, dated February 3,1978; Packing gland leak from 5--
MU-V47A caused average gaseous radioactivity concentrations in the area to exceed 500 times the values specified in 10 CFR 20.403(b)(2).
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LER 70-05/li. dated February 2,1978; Violation of containment integrity during surveillance testing.
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LER 78-06/3L, dated February 21, 1978; Sample isolation valve CA-V13 failed to close following routine sampling.
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LER 78-07/3L, dated February 24, 1978; Quadrant power tilt limit of +2.66'; was exceeded due to transient power operations and maneuvering to an all rods out configuration.
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LER 78-08/lT, dated February 16, 1978; Steam generator level
founo to be outside the limitations assumed in the Main Steam
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Line Break Analysis.
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LER 78-09/3L, dated April 3,1978; D.C. Breaker CS3 failed to trip during one of several tests due to apparent binding of the breaker operating linkage.
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LER 78-10/lT, dated April 5,1978; During ES surveillance test both diesels failed to start on initial actuation and the make-up pumps tripped when transferred to the diesel generators o'n subsequent actuation.
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LER 78-ll/3L, dated April 17, 1978; Curing visual testing of shock and sway suppressors, snubber MS-201 was found to have
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no fluid in the reservoir due to a pinched rod wiper seal.
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LER 78-12/3L, dated May 5,1978; Snubber DHH-187 would not
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lock up upon compression during functional testing.
' enc es rese 'ias selec ed f:r = site fc1:cwue.
1 Aqa 170 L. J t (._ U
LER 78-13/lT, dated May 9,1978; Error in small break LOCA
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safety analysis.
LER 78-15/lT, dated May 15, 1978; Control rods were positioned
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in the restricted region of TS limits due to use of the wrong rod index curve in the procedure.
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LER 78-16/lT, dated May 17, 1978; Error in feedwater line break accident analysis.
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LER 78-17/1T, dated May 19, 1978; Error in uncertainties applied to measurements of imbalance anc quadrant power tilt as performed by the incore detector system.
The above LERs were closed based on satisfactory review in the Region I office, except those LERs selected for onsite followup.
6.
Onsite Licensee Event Followuo For those LERs selected for onsite followup (denoted in Paragraph 5), the inspector verified that the reporting requirements of Technical Specifications GP 4703 (Original) had been met, that
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appropriate corrective action had oeen taken, that the event was reviewed by the licensee as required by Techncial Specifications, O
and that continued operation of the facility was conducted in con-formance with Technical Specification limits.
The inspector's findings regarding these licensee events were acceptable, unless otherwise noted below.
LER 78-08/1T described operation of the facility in a condition
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not analyzed in the FSAR, e.g. flooding the OTSG feedwater nozzles during operation below 5% full power.
When the Main
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Steam Line Break Analysis was originally perfor ad, the B&W recommended operating procedure required a 30" water level in the OTSGs at low power levels.
As B&W gained operating ex-perience, flooding of the feedwater nozzles at low power levels was reccmmended to prevent thermal shocking the nozzles.
How-ever, operation with feedwater no:zles submerged., when below 55 full power, was not reconsidered for the Steam Line Break Analysis.
Shutdown and cooldown procedures were changed to
"Cen0;as ~ nose.iss seleC'2d fOr Cr. site f0llCWu0.
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prevent flooding the nozzles when <5% full power. Addition-ally, the licensee has determined that should a main steam line break occur during operation below 5% full pcwer with OTSG nozzles flooded, no return to criticality will occur following reactor trip, as described in MEC letter to B&W
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serial GQL 0850, dated May 4, 1978.
The Met-Ed evaluation of OTSG operation at low power levels has not yet received concurrence from B&W.
Final resolution of this matter will be reviewed during a subsequent inscection.
(289/78-13-01).
LER 78-16/lT described the Cycle 4 reanalysis results of a
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feedwater line break accident, which indicated that ceak RCS.
pressure following the accident would exceed the maximum allowable pressure of 2750 psig by about 4.4 psig.
The RCS High Pressure Trip setpoints of Procedure 1303-4.1 have been reduced to 2390 psig to prevent exceeding 2750 psig, howe.ver, an applicable Technical Specification Change Request has not yet been submitted.
This item will be reviewed during a subsequent inspection.
(289/78-13-02).
7.
IE Sulletin and Circular Followuo
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The inspector reviewed the licensee's followup actions regarding the IE Bulletins and Circulars listed below.
IES 77-04,'" Calculational Error Affecting the Design Per-
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formance of a System for Controlling pH of Containment Sump Water Following a LOCA," dated November 4, 1977.
IES 77-05 and 77-05A, " Electrical Connector Assemblies," dated
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November 8 and 15,1977.
IEB 77-06, " Potential Problems with Containment Electrical
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Penetration Assemblies," dated November 22, 1977.
IES 78-01, " Flammable Contact-Arm Retainers in General Elec-
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tric (G.E.) CR120A Relays," dated January 16, 1973.
IES 78-02, " Terminal Bleck Qualifcation," dated January 30,
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1978.
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IES 78-04, " Environmental Qualification of Certain Stem Mounted Limit Switches Inside Reactor Containment," catec February 21, 1973.
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IE3 78-05, " Malfunctioning of Circuit Breaker Auxiliary Con-
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tact flechanism - General Electric Model CRlC5X," dated April 14, 1978.
IEC 77-14, " Separation of Contaminated Water Systems from
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flancontaminated Plant Systems," dated November 28, 1977.
IEC 77-15, " Degradation of Fuel Oil Flow to the Emergency
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Diesel Generator," dated November 30, 1977.
IEC 77-16, " Emergency Diesel Generator Electrical Trip Lock-
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Out Features," dated December 13, 1977.
IEC 78-02, " Proper Lubricating Oil for Terry Turbines," dated
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April 20, 1978.
The review included discussions with licensee personnel, review of selected facility records, and observation of selected facility equipment and components.
With respect to the above Bulletins, the inspector verified that licensee management forwarded copies of the bulletin response to appropriate ensite management representatives, that information and corrective action discussed in the reply was accurate and effected G
as described, and that the reply was submitted within the time period described in the bulletins.
With respect to the above Circulars, the inspector verified that the circular was received by appror ate licensee management, a review for applicablity was performed, and that action taken or planned is appropriate.
Acceptance criteria for the above review included inspector judge-ment and requirments of applicable Technical Specifications and facility procedures.
No items of noncomoliance were identified concerning licensee follow-up to the above Bulletins and Circulars.
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8.
In-Office Review of Periodic Recorts The periodic reports listed below were reviewed in the Region I office to verify that the report included infomation recuired to be recorted and that test results and/or suo:Cr:.'1g information in71 1T1 Ls lJl
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discussed in the report were consistent with design predictions e
and per ormance specifications, as applicable.
The recorts were aisc reviewed to ascertain whether clanned corrective action was adecuate for resolution of identified problems, where applicable, and te determine whether any information contained in the report should be classifiec as an Abnormal Occurrence.
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The following TMI 1 periodic recorts were reviewed.
Summary of Safety Related Maintenance (1977), dated March
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16, 1978.
Summary of Operating Experience (1977), dated March 16, 1978.
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10 CFR~ 50.59b Report of Facility Changes, Tests and Experiments
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conducted without prior Commission Approval (1977), dated April 25, 1978.
January Operating Status Report, dated February 10, 1978.
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Fs5ruary Operating Report, dated March 13, 1978.
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March Operating Report, dated April 14, 1978.
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April Operating Report, dated May 12, 1978.
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The above reports were closed based on satisfactory review of the Region I office.
i 9.
10 CFR Part 21 'molementation The inspector reviewed the following items related to licensee implementation of 10 CFR Part 21 requirements.
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Posted documents for informing employees of Part 21, Section
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206, and procedures adopted pursuant to Part 21, ircluding the location where the regulations and procedures may be ex-amined and to whom reports may be made.
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Selacted purchase orders.
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Noncenformance Reports77-336 and 77-396.
4 n71 132 LJ l-
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- cp These items were rev.iewed to verify the adequacy of the implementa-tion of Part 21 requirements / procedures / controls listed below.
Posting of requisite documents.
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Specification of the application of 10 CFR Part 21 provisions
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in applicable procurement documents issued after January 6, 1978.
Evaluation of deviations or conditions or circumstances, or
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failures to comply, not resulting in a report to the Commission.
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Acceptance criteria for the above review included inspector judge-
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ment and requirments of 10 CFR Part 21 and applicable licensee implementing procedures.
The inspector's findings regarding 10 CFR Part 21 implementation were acceptable.
10.
Reactor Coolant System Decontamination Several facilities have recently carried out or are planning to
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carry out decontamination of reactor coolant pressure boundary components without prior NRC approval.
This is allowed under the G
provisions of 10 CFR 50.59.
Operations of this na.ture generally do not represent unreviewed safety questions provide'd that the following conditions are satisfied.
The components involved have been physically removed from the
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system.
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The licensee's administrative controls and post-decontamination
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surveillance are adequate to assure that residual decontami-nation solution is not introduced into other components within the reactor coolant pressure boundary.
The component is not degraded.
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The inspector informed licensee representatives that NRR is actively cevelooing generic criteria to govern decontamination of reactor coolant pressure boundary components.
Until these generic criteria are develoced, decontacination solutions with unproven corrosion characteristics should,,t be introduced into com enents within the reactor coolan: cressure bcundaries of c:erating reactors without Orior NRC a:Oraval.
'A23 133
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Exit Interview The inspector met with licensee representatives (denoted in Para-graph 1) at the conclusion of the inspection on June 9,1978. The inspector summarized the purpose and scope of the inspection and the findings.
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