IR 05000266/1980019
| ML19340D626 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 11/18/1980 |
| From: | Guldemond W, Hague R, Warnick R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML19340D625 | List: |
| References | |
| 50-266-80-19, 50-301-80-19, NUDOCS 8012310500 | |
| Download: ML19340D626 (6) | |
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U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT
REGION III
i Report No. 50-266/80-19; 50-301/80-19
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Docket No. 50-266; 50-301 License No. DPR-24; DPR-27 Licensee: Wisconsin Electric Power Company 231 West Michigan Milwaukee, WI 53203 Facility Name: Point Beach Nuclear Power Plant, Units 1 & 2
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Inspection At: Point Beach Site Two Creeks, WI Inspection Conducted: November 3 and 4, 1980 RFuJ &
Inspectors:
W. G. Guldemond A/c v / A;,90 RFW frr R. L. Hague jW e /8, 20
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RFMainu'A Approved By:
R. F. Warnick, Chief A/o y /Jf/G'd)
Reactor Projects Section 2
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Inspection Summary Inspection on November 3 and 4, 1980 (Report No. 50-266/80-19; 50-301/80-19)
Areas Inspected: Routine resident inspection of the status of implementation of Three Mile Island Task Action Plan Category "A" requirements per IE Head-
quarters request. The inspection involved 38 inspector-hours onsite by two l
NRC inspectors including 4 inspector-hours during off-shifts.
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Results: No items of noncompliance were identified.
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DETAILS
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Persons Contacted
- G. A. Reed, Manager, Nuclear Power Division
- R. E. Link, Assistant to the Manager T. J. Koehler, Operations Superintend ent J. C. Reisenbuechler, I&C Engineer R. R. Weedon, Health Physics Supervisor J. J. Zach, Superintendent Technical Services F. A. Zeman, Office Supervisor The inspectors also talked with and interviewed members of the Operations, Maintenance, Health Physics and Instrument and Control Sections.
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Implementation of TMI Task Action Plan (TAP) Category "A" Requirements The inspectors ascertained the status of implementation of TMI TAP Category "A" requirements in accordance with an October 24, 1980 memorandum from S. E. Bryan to J. G. Keppler SSINS Number 6500. The results of the inspection are documented below, referenced by NUREG 0578 paragraph number, TAP number and title. These results were forwarded by separate memorandum to IE Headquarters via Region III on November 5, 1980.
No items of noncompliance were identified.
NUREG 0578 TAP Number Number Title a.
2.2.1.b I.A.1.1 Shift Technical Advisor: A Shift Technical Advisor is on duty around the clock. His duties are described in Section 3.13.2 Revision 0 dated March 28, 1980 of the Administrative Control Policies and Procedures Manual (ACPPM).
Provisions have been made to ensure he has instant communications with the Control Room at all times. The Shift Technical Advisor has no line authority to direct the manipulation of controls. His role is advisory to the Shift Supervisor and/or the Duty and Call Superintendent.
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2.2.1.a I.A.1.2 Shift Supervisor Responsibilities:
The reporting relationship, authority and responsibilities of the Shift Supervisor are delineated in the following ACPPM sections:
f Section 3.1, Revision 7 dated 10-12-80, Organization Chart Section 3.2, Revision 3 dated 05-08-78, Manager Nuclear Power Section 3.4, Revision 4 dated 05-08-79 Operations Superintendent l
Section 4.2, Revision 1 dated 05-08-78, Shift Supervisor Section 4.3, Revision 2 dated 05-08-78, Operation Supervisors Section 4.4, Revision 1 dated 12-15-76, Licensed Control Operators I
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These descriptions of the line organization are presently being
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revised. Additional reinforcement for the operational authority of the Shift Supervisor is provided in Section 3.1, Revision 2 dated 08-08-80 of the Duty and Call Superintendent Handbook.
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I.C.1 Short Term Accident & Procedures Review: This item as it relates to Small Break LOCA procedures was closed out in Inspection Report No. 50-266/80-09 and No. 50-301/80-08. This report was reviewed and the inspection results verified. The following procedures address Small Break LOCA's:
E0P-1A Large Loss of Reactor Coolant Revision 17 02-05-80 E0P-2A Secondary Coolant Break Revision 10 02-05-80 E0P-3A Steam Generator Tube Rupture Revision 11 02-05-80 E0P-4A Reactor Coolant Leak Revision 3 02-05-80 Revision 0 dated 12-27-79 of Special Order 80-07 entitled Inade-quate Core Cooling During SBLOCA provides additional guidance in this area.
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2.2.2.a I.C.4 Control Room Access: Access to the control room during normal and accident conditions is re-stricted in accordance with the Duty and Call Superintendent Handbook Section 1.2, Revision 2 dated 08-08-80.
This procedure limits the number of personnel in the control room at any time to manageable levels such that operations will not be impeded.
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2.1.8.a II.B.3 Post-Accident Sampling: Post-accident sampling procedures are promulgated in the Health Physics Administrative Control Policies and Procedures Manual section HP 17.6.2, Revision 4 dated 10-24-80.
Analyses required by this procedure include hydrogen and oxygen content of the reactor coolant by gas partition, radioactive noble gases, boron /pH, chloride concentration, and iodine and gamma scan. This same procedure describes the post accident sampling equipment required.
This equipment was verified to be installed and operational.
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2.1.5.c
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Recombiner Procedures Review and Upgrade: This item is not applicable as hydrogen recombiners are not utilized. Tbf.s was deemed acceptable per an April 9, 1980 letter and staff evaluation from Mr. A. Schwencer, Chief, Operating Reactors Branch No. 1, Division of Operating Reactors to the licensee.
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2.2.2.b III.A.1.2 Upgrade Emergency Support Facilities: Until the new Technical Support Facility which is under construction is completed the licensee is using the conference room in the administrative building as an interim Technical Support Center. This room contains two plant phones, one outside phone line, and Emergency Notification System phone,
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a National Warning System phone, a copy of the Emergency Plan,
and a copy of the Technical Specifications. Radiation detection and air sampling equipment is immediately available in an adjacent room along with six self contained breathing apparatuses.
Copies of plant drawings, prints, and procedures are available in the adjacent administrative offices. Use of the interim center is in accordance with the Point Beach Emergency Plan, Revision 13 dated 09-15-80.
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2.2.2.c III.A.1.2 Onsite Operational Support Center:
The onsite operational support center consists of a room in the basement of the Energy Information Center which is located approximately 200 yards south of the main gate. The center contains two in plant phones and plant general layout drawings.
Its use is specified in the Point Beach Emergency Plan, Revision 13 dated 09-15-80.
This procedure specifies that all licensee personnel not immediately involved in the emergency or directed to report to the Technical Support Center will report to the Onsite Operational Support Center.
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2.1.6.a III.D.I.1 Primary Coolant Outside Containment:
Leakage rate tests have been performed on the SI, CS, WGS, sampling and CVCS letdown, and charging and holdup tank systems. Procedures have been written to perform additional leakage rate tests on the liquid systems annually. The waste gas system will also be checked annually by a contractor. The tests basically consist of system walkdowns to check for leakage. Deficiencies discovered are corrected as soon as practicable. Recent plant walkdowns have not revealed any cases of excessive liquid leakage.
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2.1.8.c III.D.3.3 Inplant Radiation Monitoring: The air monitoring system utilizes portable air samplers. Charcoal or silver zeolite filters are used and analyzed in the counting room.
The analytical ability has been enhanced by the use of a Baird Model 530 Single Channel Iodine Spectrophotometer. Use of this instrument for this purpose is specified in the Health Physics Administrative Control Policies and Procedures Manual section HP 17.6.5, Revision 0 dated 10-24-80.
Appropriate personnel have been trained in the use of this instrument, k.
2.1.8.b II.F.1 Additional Accident Monitoring Instrumentation: The following equipment has been installed and is operational based on observation of control room readouts, previous observation of physical location of detectors, and a review of surveillence testing data:
Monitor Point Range Alarm Auxiliary Building Vent Stack
.01-100R/hr 10er/hr Drum Area Vent Stack 1-10,000mr/hr 1mr/hr Unit 1 Containment Purge Stack
.1-1000R/hr 100mr/hr Unit 2 Containment Purge Stack
.1-1000R/hr 100mr/hr Combined Air Ejector Discharge 1-10,000mr/hr Imr/hr Gas Stripper Building Vent Stack
.01-100R/hr 10mr/hr-4-
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Procedures for the use of these monitors and conversions to release rates are provided in the-Health Physics Administrative Control Policies and Procedures Manual section HP 17.6.4, Revision 2 dated 03-21-80.
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2.1.7.a II.E.1.2 Auxiliary Feed Initiation: The auxil ary feedwater system is a safety grade system which has provisions for automatic initiation and is designed to meet single failure criteria. Documentation of this point has been submitted to the NRC by way of Mr. C. Trammell in a meeting May 11, 1979, a letter dated October 29, 1979 to Mr. D. Eisenhut, and by drawings submitted December 17, 1979 at the request of Mr. A. Schwencer.
Due to a lack of appropriate specifications and standards, on site verification of the safety grade nature of the auxiliary feedwater system cannot be made.
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2.1.7.b II.E.1.2 Auxiliary Feed Flow: Auxiliary feed flow indication is provided by use of a D/P cell at the discharge of each of the four pumps. Power for these flow channels is supplied by two inverter fed instrument panels which in turn are energized from the batteries.
Backup indication is provided by three separate and independent steam generator level instrument channels for each steam generator.
Installation of flow transmitters in the individual steam generator auxiliary feed lines is progressing with expected completion dates'of 01-01-81 for Unit I and during the next refueling for Unit 2.
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2.1.3.b II.F.2 Instrumentation for Inadequate Core Cooling:
Presently three methods of determining subcooling are utilized. The first utilizes the computer with display and trend recorded indications. The inputs are froe the core exit thermocouples and safety grade narrow range pressure (1700-2500)
and non-safety grade wide range pressure (below 1700). The second method is an X-Y plotter with inputs from one hot leg RTD and the wide range pressure sensor. The third method is a manual determination by the operator using control board indications. A fully qualified subcooling meter is planned as part of an accident analysis control board to be installed by 10/1/81.
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2.1.4 II.E.4.2 Containment Isolation: Presently PBNP is operating under administrative controls special order 80-13 for the manual closure of certain identified valves.
Some of these valves-S/G Blowdown and sampling and primary sampling valves-will be rewired for automatic closure during the next
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refueling for each unit. One valve in the auxiliary charging
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line is the subject of correspondence with Westinghouse. This
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licensee has been waiting since July 1980 for a reply. The l
remainder of the identified valves are for seal water return isolation. The licensee committed to install those valves during the next refueling after receipt of the valves on site.
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These valves have yet to be ordered. Diversity is provided by l
the automatic containment isolation signals including any initiation of safety injection and high containment pressure.
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2.2.1.c I.C.2 Shift Relief and Turnover
Procedures:
Shift relief procedures have been modified to include signatures of on-coming and of f-going auxiliary operators, control operators, operating supervisors and shift supervisors on turnover checklists designated OPS 62 and OPS 33. No formal mechanism for turnover exists for instrument technicians or maintenance personnel.
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2.1.3.a II.D.3 Valve Position Indication Relief and Safety Valves: Position indication on PORV's is provided by safety grade limit switches detecting stem motion. The open limit switch provides an alarm in control. The safety valves are moni-tored by four acoustical transducers-one each mounted immediately downstream of the safeties. Direct readout is provided on a panel in the cable spreading room with alarm indication in the control room.
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2.1.1 II.E.3.1 Emergency Power for Pressurizer Heaters: On both units the C&D backup heaters and the control heaters are powered from the 480 volt emergency buses. The control heatert and D backup heaters from BO-4 and the C backup heaters from B0-3.
Those buses are supplied from separate diesel generators upon loss of offsite power.
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2.1.1 II.G.1 Power Supplies for Pressurizer Relief Valves and Level: On both units the PORV's are powered from B-17 and B-9 which receive their power from the station batteries. The block valves are powered from 480 volt emergency buses which will be energized from the diesel generators on loss of offsite power. Two of three pressurizer level channels are powered from vital instrument buses which in turn are powered by inverters fed from the station batteries.
3.
Exit Interview The inspector met with licensee representatives (denoted in Paragraph 1)
at the conclusion of the inspection on November 6, 1980 and summarized the scope and findings of the inspection activities. The licensee acknowledged these findings.
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