IR 05000277/1981003
| ML20004C794 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 03/23/1981 |
| From: | Blough A, Cowgill C, Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20004C785 | List: |
| References | |
| 50-277-81-03, 50-277-81-3, 50-278-81-03, 50-278-81-3, NUDOCS 8106050266 | |
| Download: ML20004C794 (19) | |
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U.S. NUCLEAR REGULATORY COMMISSION 0FFICE OF INSPECTION AND ENFORCEMENT g
Region I 50-277/81-03
, Report No. 50-278/81-03 _
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50-277
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Docket No. 50-278
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DPR-44 c
License No. OPR-56 Priority Category c
Licensee:
Philadelphia Electric Company 2301 Market Street'
Philadelphia,Peniis91vania
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Facility Name:
Peach Bottom Atomic Power Station, Units 2 and 3 Inspection at:
Delta, Pennsylvania Inspection conducted: January 1 - 31, 1981
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k-3 23I8I 8.E Inspectors:
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C. J. Cowgill, III, Res'ident R'eactor Inspector
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date signed b
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A. R. Blough, Resident Rea'ctor fnspector date signed date signed bO tD P
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Approved by:
n E. C. McCabe, Jr;, Chief, Reactor Projects date signed Section No. 2, DR&PI
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Inspection Summary:
Inspection on January 1-31,1981 (Contined Inspection Report Nos. 50-277/81-03 and 50-278/81-03)
Areas Inspected:. Routine, onsite regular and backshift inspections by the resident inspectors (65.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Unit 2; 65 hours7.523148e-4 days <br />0.0181 hours <br />1.074735e-4 weeks <br />2.47325e-5 months <br /> Unit 3).
Areas inspected included accessible portions of the Unit 2 and Unit 3 facilities, operational safety, radiation protection, physical security, control room observations, LER review, IE Bulletin followup, Emergency Procedure Review, outstanding item followup, facility tours and review of periodic and special reports.
Resul ts:
Noncompliances:
none in nine areas, three in two areas (improper Limiting Safety System Setpoints during power ascension, Detail 3; improper personnel escort practices, Detail 8; and improper vehicle escort practices, Detail 8).
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l Region I Form 12 (Rev. April 77)
i 8106050 % 4
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DETAILS 1.
Persons Contacted W. H. Alden, Engineer-in-Charge, Nuclear Section W. Corse, Assistant Site Q.A. Engineer J. K. Davenport, Maintenance Engineer G. F. Dawson, I&C Engineer
- R. S. Fleischmann, Assistant Station Superintendent A. Fulvio, Results Engineer N. Gazda, Health Physics, Radiation Protection Manager D. Helker, Test Engineer C. P. Lauletta, Training Coordinator F. W. Polaski, Reactor Engineer S. R. Roberts, Operations Engineer D. C. Smith, Outage Coordinator S. A. Spitko, Site Q. A. Engineer S. Q. Tharpe, Security Supervisor
- W. T. Ullrich, Station Superintendent J. E. Winzenried, Technical Engineer R. H. Wright, Test Engineer Other licensee employees were also contacted during the inspection.
- Present at exit interviews on site and for summation of preliminary inspection findings.
2.
Outstanding Item Update (Closed) Inspector Follow Item (277/80-35-03), review the licensee's training for HP technicians performing manual UT of the Scram Discharge Volume.
The inspector reviewed training materials, including training description letters, certification letters, course handouts, and the procedure in which personnel are trained (ST 9.23 - Appendix A). The l
training description letter states that the training applies only'to
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scram discharge header level measurement and does not qualify the individuals to perform other ultrasonic measurements.
The inspector also discussed the training program with qualified personnel.
No
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l unacceptable conditions were identified.
(Closed) Unresolved Item (277/80-32-03 and 278/80-24-03), clarification was needed of precedure A-7 with respect to gathering operating experience
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assessment (CEA) data from shift technical advisors (STAS) between scheduled monthly meetings. The inspector reviewed administrative procedure A-7, " Shift Operations," Revision 15 dated December 2, 1980.
The procedure now requires STAS to convey operating experience concerns to the Operations Engineer between monthly OEA Committee meetings.
The inspector.had no further questions in this area.
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4 (Closed) Deficiency (278/80-14-01), failure to. log valve. positions daily when a Primary Containment Isolation valve is inoperable. The'
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specific failure to log valve positions'was corrected upon discovery,
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and was verified to be corrected by the inspector, as detailed in
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combined report' 50-277/80-16 and 50-278/80-14.
Cince that occurrence the inspector, through discussions with operators and review of-
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surveillance logs,:has verified that personnel were implement-ing the requirements.
For example.on January 20,-1981.the inspector
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reviewed ST 5.3, " Inoperable Isolation Valve Position Daily Log,"
Revision 3, dated August 14,1979, in use at Unit 2 :for inoperable-valves A0-2512 and A0-2514.
Piping and instrument diagrams were reviewed
to verify that valve positions. logged on ST 5.3. included at least one other valve in each line having an inoperable isolation valve. No:
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unacceptable conditions were identified.
(Closed) Inspector Follow Item (79-29-03 and 79-32-02),. review adequacy of General Employee Training (GET) Program. The licensee's program.
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was reviewed in combined report 277/80-11 and 278/80-11 and.in combined
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report 277/80-34 and 278/80-27. These reviews included examination of licensee procedures, lesson plans, quizzes and records; and discussions A no unacceptable condi-with course attendees.. Those inspections noeE
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tions. Additionally during this inspection, the inspector attended GET i
to. determine that employees were being instructed in administrative
controls and procedures, radiological health and safety, industrial
safety, controlled access and security procedures, emergency plans, l
and. quality assurance as required by the licensee's program. The training session was presented professionally and enthusiastically. The inspector
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verified that female employees in the-training are instructed regarding prenatal radiation exposure and are provided with the applicable Regulatory Guide. The inspector noted that the training included an updated presentation on biological effects of ionizing radiation, Land a new' film on security areas and procedures.
" Nuclear Plant Rules," a licensee summary of important individual responsibilities, was discussed, and copies were made tvailable. Tne inspector provided a licensee representative with critique comments on the training session, includ-ing a list of minor technicai errors on films and slides that were not corrected by the instructor. The licensee's GET examination was com-pleted. The inspector noted that the examination was not administered
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in a controlled mancer, in that positive measures were not taken to preclude student access to notes and visual aids during the exam. When
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apprised of this tae licensee stated that the matter would be discussed with instructors co prevent recurrence. Additionally, each participant
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in the exam was required to be reexamined.
The inspector had no further
questions in this matter.
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(C1csed) Unresolved Item (79-10-11-and 79-11-11), seismic monitoring requirements required additional NRC review. The inspector reviewed Technical Specification Amendments No. 75 (Unit 2) and No. 74 (Unit 3)
and conducted discussions with the NRR Licensing Project Manager.
The Technical Specification was issued to provide requirements equivalent to those of the BWR Standard Technical Specifications.
The inspector noted that the licensee had installed a new seismic monitoring system to meet the requirements. The inspector verified that the system which includes _ control room alarms, was installed and operating. The inspector reviewed ST 4.14, " Instrument Check of the Seismic Monitoring System," Revision 0, dated January 15, 1981, performed January 15, 1981, to verify that surveillance testing was up-to-date.
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All steps of the test had been completed satisfactorily. This surveillance test is required by Technical Specification 4.15 to prove operability per Technical Specification 3.15.
The inspector noted that the test.is considered satisfactory with respect to Technical Specification require-ments if all asterisked steps are completed satisfactorily. However, none of the procedure's 25 steps is asterisked.
The adequacy of this procedure is considered unresolved pending clarification by the licensee (277/81-03-01 and 278/81-03-01).
The inspector also reviewed procedures associated with operation of the seismic monitoring panel and with earthquakes to verify that they had been updatad. An information letter to the operators regarding the modification was also reviewed.
No items of noncompliance were identified.
(Closed) Inspector Follow Item (78-14-01 and 78-18-01), potential misapplication of GE Type 1C2820A relays in safety-related DC Motor Control Centers (MCCs). The inspector reviewed a licensee engineering letter, dated December 9, 1980, documenting an investigation of safety related MCCs and local DC starters.
No use of GE Type 1C2820A relays was identified.
The inspector had no further questions in this matter.
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(Closed)
Infraction (277/73-01-01), MSIV's steam tunnel high temper-ature trip setpoints were manually increased above Technical Specification required values on December 20, 1977.
Based on known reasons for the temperature increase, a loss of ventilation, the licensee had elected to increase the setpoints of less than or equal to 200 F to 250 F
until ventilation could be returned to normal.
The licensee's response stated that site personnel had been instructed at station meetings that Technical Specification limiting conditions for operation must be met.
During discussions with licensed operators the inspector verified aware-ness of the important of adherence to Tachnical Specifications.
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On January 16, 1981 failure of a relay in the Unit 2 Group III Primary Containment Isolation circuitry resulted in loss of most of the main l
l steam tunnel ventilation flow for about 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. The licensee was l
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again concerned that, upon restoration of normal ventilation flow, stagnant air at greater than 200 F would be exhausted past the in-duct temperature sensors, causing a Group I isolation and scram.
The isolation and scram would be unnecessary, since the protective action is designed to provide isolation in event of a steam leak.
The licensee requested and received verbal authorization on January 16, 1981 from NRC:NRR for a temporary change to the Technical Specifica-tions, allowing an increase in the trip level setting for main steam tunnel exhaust duct high temperature from 200 F-to 250 F for a short period of time while ventilation was being returned to normal operating conditions. At about 4:15 p.m., January 16, 1981 the inspector observed restoration of the Unit 2 ventilation system. The inspector verified that only 4 of the 16 temperature sensor's r,etpoints were raised, that temperatures were continuously monitored during the time that setpoints
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were at 250 F, and that the setpoints were. returned to their original values (190-195 F) once the ventilation system was operating normally.
The inspector reviewed a letter from NRR to PECO dated January 19, 1981, confirming the temporary Technical Specification change authorization.
h e inspector had no further questions relative to Technical Specification cdherence.
However, the licensee's response dated February 28, 1980 to the earlier item stated that a design review had been initiated to investigate a means of preventing recurrence of pipe tunnel temperature instruments receiving improper input data. This design review was not available during this inspection and will be examined subsequently.
(81-03-02 and 81-03-02)
3.
Plant Operations Review a.
Logs and Records (1) Documents Reviewed
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A sampling review of lags and records was made to:
identify significant changes and trends; assure that required entries were being made; verify that operating orders and night orders conform to Technical Specification requirements; check correctness of communications concerning equipment and lock-out status; verify jumper log conformance to procedural requirements; and to verify conformance to limiting conditions for operations.
Logs and records reviewed were:
(a) Shift Supervision Log, January 1-31, 1981 (b) Reactor Engineering Log - Unit 2, January, 1981 (c) Reactor Engineering Log - Unit 3, January, 1981 (d) Reactor Operators Log - Unit 2, January 1-31, 1981
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(e) Reactor Operators Log - Unit 3, January 1-31, 1981 (f) LO Log Book - January 1-31, 1981 (g) Night Orders - Current Entries (h) Radiation Work Permits (RWP's) - Various in both Units 2 and 3, January, 1981 (i) Maintenance Request Forms (MRF's) - Units 2 and 3, Sampling Audit, January, 1981 (j)
Ignition Source Control Checklists (Sampling), January, 1981 (k) Operation Work and Information Data - January,1981 Control room logs were reviewed pursuant to requirements of Administrative Procedure A-7, " Shift Operations." Frequent initialing of entrie: by licensed operators,' shift supervision, and licensee on site management constituted evidence of licensee review.
Logs were also reviewed to assure that plant conditions including abnormalities and significant operations were accurately and completely recorded.
Logs were also assessed to detennine that matters requiring reports to the NRC were being processed as suspected reportable occurrences.
No unacceptable conditions were identified.
(2)
Fact 11ty Tours (a) During the course of this inspection, which also-included shift turnover, the inspector conducted daily tours and made observations of:
Control Room - (daily)
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Turbine Building - (all levels)
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Reactor Building - (Accessible areas)
Diesel Generator' Building
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. Yard area and perimeter exterior to the power block,
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including Emergency Cooling Tower and torus dewater-ing tank Security Building, including CAS, Aux SAS, and control
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point monitoring Lighting
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Vehicular Control
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The SAS and power block control points
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Security Fencing
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Portal Monitoring
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Personnel and Badging
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Control of Radiation and High Radiation areas
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including locked door checks TV monitoring capabilities
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Off-Shift Inspections during this inspection period and the areas examined were as follows:
Date Areas Fxamined January 5, 1981 Tour protected area, escort procedures January 7, 1981 Control room tour January 9, 1981 Control room tour January 12, 1981 Control room tour January 13, 1981 Control room tour January 15, 1981 Tour of protected area January 16, 1981 Control room tours; observed restor-ation of MSIV tunnel ventilation January 20, 1981 Control room tour January 21, 1981 Control room tour January 22, 1981 Con' ol room tour January 26, 1981 Tour of protected area and control roon January 28, 1981 Tour of turbine building
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January 29, 1981 Control room tour
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Off-Norr.:a1 Alarms. Selected annunciators were
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discussed w:th control room operators and supervision to assure they were knowledgeable of plan; conditions and that corrective action, if required, was being-taken.
Examples of specific alarms discussed during the report period were: APRM High; Rod Withdrawal Block; Scram Discharge Instrument Volume Not Drained; Reactor Building Equipment Drain Sump Level High; Turbine High Vibration; and Condensate Storage Tank Level, High/ Low.
The operators were knowledgeable of alarm status and plant conditions.
Control Room Manning. On frequent occasions during^
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this inspection, the inspector confirmed that require-ments of 10 CFR 50.54(k), the Technical Specifications,
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and the NRR letter of July 31, 1980 for minimum staffing requirements were satisfied.
The inspector frequently verified that a senior licensed operator was in the control room complex. No unacceptable conditions were identified.
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Fluid Leaks.
No significant fluid leaks were iden-tified which had not also been identified by the licensee nor for which necessary corrective action had not been initiated.
The inspector observed sump status, alarms, pump-out rates, and held discussions with licensee personnel.
No unacceptable conditions were identified.
Piping Vibration.
No significant piping vibration or
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unusual conditions were identified.
Monitoring Instrumentation.
The inspector frequently
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confirmed that selected instruments were operating
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and indicated values were within Technical Specifica-tion requirements.
On a daily basis when the
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inspector was on site, ECCS switch positioning and valve lineups, based on control room. indicators and plant observations,were verified.
Examples of instrumentation observed included flow setpoints, breaker positioning, PCIS status, radiation monitoring instruments, Scram Discharge Volume Continuous Monitoring System level instruments and Suppression Pool level and temperature instruments. No unaccept-able conditions were identified.
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b.
Reactor Water Chemistry Thu following surveillance tests for the periods indicated were reviewed by the inspector to assure that Technical Specification
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Limits were satisfied.
(1) Conductivity and Chloride Ian Content in Primary Coolant During Normal Operation and Time Conductivity and Chloride Are Above Specified Limits Surveillance Tests 7.2.3.A and 7.2.3.C and Nach Bottom Daily BWR Chemistry Ana. lysis - January 1-31, 1981.
Techrical Specification 3.6.B requires prior to startup and when operating at rated pressure, reactor water conductivity at 25 C of less than or equal to 5.0 umho/cm and chloride concentration less than or equal to 0.2 ppm.
Reactor water-quality may exceed these limits for up to two weeks.per year.
Maximum lin:its are established as 10 umho/cm conductivity and 1.0 ppm chlorides.
Inspections at Unit 2 for the report period indicated that during operation the maximum conductivity and chloride con-centrations were 1.25 umho/cm and 0.025 ppm respectively.
In January 1981, there was no time above the 0.2 ppm chlorides
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limit or the 5.0 umho/cm conductivity limit.
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l Inspections at Unit 3 for the period indicated that, during operation, the maximum conductivity and chloride concentrations
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reached were 6.49 umho/cm and.360 ppm respectively.
Through l
January 31 the 1981 total time above the specific "two weeks j
per year" limits for conductivity and chlorides were 12.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> and 12.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> respectively.
No unacceptable conditions
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were identified.
(2) Determination of Dose Equivalent Microcuries/ Gram I-131 in the Primary Coolant
Surveillance Test 7.2.1.A was reviewed. The licensee analyzes
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the following nuclides:
I-131, I-132, I-133, I-134, and I-135 and computes dose ~ equivalent I-131 -- that amount of-
I-131, which alone would produce the same dose as the quantity and isotopic mixture actually present. The Technical Specification Limit is 2.0 microcuries per gram.
Increased sampling frequency is' required if any analysis exceeds 0.02
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microcuries per. gram (2E-3 uCi/g). The representative sample for Unit 2, Enalyzed on January 6, 1981 indicated a dose
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equivalent I-131 concentration of 1.09E-3 uCi/g.,At Unit 3, I
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a sample on January 8,1981 indicated a dose equivalence of 1.96E-3 uCi/g. The inspector also confirmed-that the required surveillance frequency was being satisfied. No unacceptable conditions were identified.
c.
Review of-Reactor Operations Definitions:
FRP = fraction of rated power s power level divided by rated thermal power (3293Mw)
MFLPD = maximum fraction of limiting power. density a the -
maximum ratio, for any core location, of actual linear heat generation rate (LHGR) to design LHGR.
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Technical Specifications require that, when MFLPD is greater than FRP, reactor protection system (RPS) Limiting Safety System. Settings
.be modified by a factor of FRP/MFLPD.
This assures that the linear heat generation cate (LHGR) transient peak is not increased for any combination of MFLPD and FRP.
The inspector reviewed the Unit 2 Assistant Control Operator's log book, the Reactor Engineer's Log, and various process computer print-outs associated with a Unit 2 power reduction for control rod patterr adjustment and subsequent power ascension on January 3, 1981. The inspector noticed that MFLPD was greater than FRP during the power ascension.
The table below lists data derived from process computer
" Periodic NSS Core Performance Lcg":
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.552 0025
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.627 0150
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.616 0214
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.670 0249
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.737 0355
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.742 0429
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.818 0500
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.815 0603
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.880 0620
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.891 0641
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.893 To implement Technical Specification requirements for MFLPD greater than FRP, the licensee uses surveillance test ST 3.3.2'" Calibration of the Average Power Range Monitoring (APRO) System," Revision 4 dated april 30, 1980.
From review of the Reactor Engineer's Log.
and completed surveillance tc:ts the inspector determined that the test was not performed on January 3, 1981 until.0635.
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The inspector reviewed surveillance _ test ST 3.3.1, "APRM Functional and Calibration Test (Scram and Rod Block)," Revision 13, dated May 2, 1978, performed December 31, 1980 and. January 5, 1981 and determined that the actual scram setpoints were within lis to 3% of their allowable maximum values for the case of_ MFLPD less than FRP. Therefore, for the MFLPD values above FRP existing from 0025 to 0635,.the actual scram setpoints were above the maximum allowable. This failure to adhere to Technical Specifica-tion Limiting Safety System Settings is an item of noncompliance (277/81-03-03).
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4.
IE Bulletin Follow-up a.
IE Bulletin 80-24, " Prevention of Damage Due to Water Leakage Inside Containment (October 17, 1980 Indian Point 2 Event)."
The bulletin required licensees to identify open and closed cooling water systems inside containment and to provide summaries of experiences with cooling water system leakage into containment.
For plants with open cooling systems (i.e., systems that use unmeasured intake from and exhaust to an undefined volume, such as a river), additional requirements regarding leak detection
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and water removal applied.
During discussions on this matter, various plant technical personnel stated that there had been very few occurrences of cooling system leakage inside containment. The inspector reviewed the licensee's bulletin response dated January 5, 1981. The response stated that Peach Bottom has no open cooling systems inside containment. The inspector's review of this matter also identified no open systems.
The response discussed a total of four leaks since both units began commercial operation. The licensee's response was reviewed, and a tabulation made for NRC review of all licensee responses.
The inspector discussed with the cognizant engineer the method used to obtain the data --a review of maintenance records and checks of operating logs.
The inspector also verified that PORC had reviewed the belletin. No unacceptable conditions were identified.
b.
IE Bulletin 80-17, " Failure of 76 of 185 Rods to Insert During a Scram at a BWR."
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Supplement 4 to Bulletin 80-17 dated December 18, 1980 required that
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an operability test be performed on the Scram Discharge Volume Continuous Monitoring System (CMS). This test was successfully completed in Unit 2 on January 6, 1981 using Special Procedure 414 " Operability Testing of SDV UT Level Transmitters", Revision 0, dated December 18, 1980. The licensee declared the Unit 2 SDV CMS operable based on the results of this testing.
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The inspector reviewed the results of this test and confirmed that the CMS levels indicated were consistent with independent level measurements. The licensee was not able to test the alarm function using the test. The inspector asked to see the initial calibration data and was shown the instrument calibration sheets for each of the transducers used in the system. The results of the initial calibration confirmed that each alarm actuated at the proper level. The inspector identified no unacceptable conditions.
5.
Review of Licensee Event Reports (LER's)
The inspector reviewed LER's submitted to the NRC:RI office to verify that the details of the event were clearly reported, including the accuracy of the description of cause and adequacy of corrective action. The inspector determined whether further information was required from the licensee, whether generic implications were indi-cated, and whether the event warranted onsite follow-up. The inspector verified that appropriate corrective action was taken and that continued operation of the facility was conducted in accordance with Technical Specifications. Compliance with current reporting require-ments and applicability to other site systems and components were also reviewed.
The following LER's were-reviewed:
LER No.
LER Date Event Date Subject 2-80-26-1P November 6, 1980 November 5, 1980 High Pressure Coolant Injection System inoperable
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3-80-28-1P December 4, 1980 December 3, 1980 High Pressure Coolant Injection System inoperable
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3-80-28-lT December 17, 1980 December 3, 1980 High Pressure Coolant Injection
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System inoperable'
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6.
Review of Emergency Procedures The inspector reviewed licensee emergency and operating procedures to verify adequacy for coping with anticipated transient without scram (ATUS) events.
The following procedures were reviewed:
GP4, " Scram Without Group I Isolation," Revision 15, dated August 5, 1980 GP6, " Scram with Group I Isolation," Revision 18, dated October 18, 1980
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OT-13, " Impending Loss of Ability to Scram," Revision 2, dated November 26, 1980 OT-38, " Loss of the CRD Regulating Functions," Revision 2, dated May 11, 1977 S3.6.B. " Initiation of the Standby Liquid Control System," Revision 6, dated December 26, 1980 GP7, " Turbine Trip Without Bypass," Revision 10, dated October 18, 1980 E-2, " Main Steam Line Break - Outside the Drywell," Revision 5, dated August 3, 1980 E-3, "Small Line Break Outside the Drywell," Revision 3, dated August 2, 1980 E-10, " Loss of All Outside Power," Revision 14, dated December 2, 1980 E-21, " Total Loss of Feedwater with a Stuck Open Relief Valve,"
Revision 2, dated December 2, 1980 OT-1, " Reactor Lov Water Level," Revision 7, dated August 29, 1979 OT-4, " Reactor Vessel High Pressure," Revision 8, dated October 18, 1980 The inspector noted that scram procedures GP-4 ard CP-6 required the cperator to verify the scram, including checking all rods fully.
inserted and checking that neutron power level is decreasing.
Per procedures, verify means if an automatic action has not occurred when it should, then manually carry out the action.
The. inspector verified that the scram procedures require the reactor mode switch in other than "RUN" (i.e., " Shutdown").
The procedures require additional actions if rods are in an abnormal configuration following a scram --
if two or more adjacent rods or if 30 or more control rods fail to insert past position 06, the operator must trip both recirculation pumps and further attempt to insert rods by:
scramming individual rods, repeatedly resetting and inserting the reactor protective system
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scram signal, venting the scram air h ader, and taking action to vent and drain the scram discharge volume.
If at any time rods are still in an abnormal position as noted above and either reactor
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level cannot be maintained or suppression pool temperature reaches
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110 F, standby liquid control is to be initiated.
These provisions meet the NRC criteria as established in IE Bulletin 80-17.
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The inspector reviewed the following sample of procedures to verify that the ooerator is directed to " verify" that a scram occurs when dictared by plant conditions: GP '7, E-2, E-3, E-10, E-21 and OT-1.
- In reviewing procedure OT-4, " Reactor Vessel High Pressure," the
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inspector noted that no verification of automatic actions is required by procedure. A licensee representative explained that the procedure is designed to control pressure below trip setpoints and that, once trip points are reached, other procedures, such as GP-4 and GP-6, take precedence.
The inspector had no further questions regarding OT-4.
The inspector reviewed the authorities and responsibilities of operators with respect-to use of the Standby Liquid Control System.
Procedure 5.3.6.B lists situations in which Shift Supervision is to direct that the liquid poison be irjected.
In ATWS situations, if criteria for initiation are met and shift supervision is not available to grant permission, the licensed operator is to inject the poison. The inspector noted however, that GP-4 and GP-6 require the operator to attempt to contact supervision for 2 minutes before injecting the poison; this is in accordance with a prior revision of 5.3.6.B.
A licensee representative stated 'that GP-4 and GP-6 would be revised to delete the 2 minute stipulation. The inspector will review these changes (277/81-03-04 and 278/81-03-03).
The inspector verified that the key for Standby Liquid Control initiation is immediately available to the operator.
No items of noncompliance were identified.
7.
Radiation Protection During this report period, the inspector examined work in progress in accessible areas of the Unit 2 and Unit 3 facilities. Areas examined included:
a.
Health Physics (HP) controls b.
Badging Us, age of protective clothing c.
d.
Personnel adherence to RWP requirements
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e.
Surveys f.
Handling of potentially contaminated equipment and materials
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Additionally, !nspections were conducted of usage of friskers and portal monitors by personnel. exiting various RWP areas, the power block, and the licensee's final exit point. More than 40 people were observed to meet frisking requirements of Health Physics procedures during the month.
A sampling of high radiation doors was verified to be locked as required.
On January 21, 1981, the inspector received word from NRC:RI of a problem with explosion detectors at some other sites. Certain models containing a nickel-63 source, exempt from licensing, were found to have leakage in excess of SE-3 microcuries.
SE-3 microcuries is the Technical Specification limit for removable contamination of miscellaneous.
radioactive materials. The inspector determined that the licensee's four detectors were one of the models of concern and requested that the
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licensee perform leak checks.
The-inspector reviewed the results of the
' leak checks, all of which indicated less than minimum detectable activity (MDA). MDA for the test conditions was 1.5E-5 microcuries.
No unacceptable conditions were identified.
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THIS PAGE, CONTAINING 10 CFR 2.790 INFORMATION, NOT FOR PUBLIC DISCLOSURE, IS INTENTIONALLY LEFT BLANK.
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9.
In-Office Review of Monthly Operating Reports The following licensee reports have been reviewed in-office onsite.
Peach Bottom Atomic Power Station Monthly Operating Report for:
December, 1980 dated January 10,'1981 This report was reviewed pursuant to Technical Specifications and verified to determine that operating statistics had been accurately reported and that narrative summaries of the month's operating experience were contained therein.
No unacceptable conditions were identified.
10. Unresolved Items Unresolved items are items about which more information is required to ascertain whether they are acceptable items, items of noncompliance, or deviations. An unresolved item is discussed in Detail 2.
11. Management Meetings a.
Preliminary Inspection Findings A summary of preliminary findings was provided to the Station
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Superintendent at the conclusion of the inspection.
During the period of this inspection, licensee management was periodically notified of the preliminary findings by the resident inspectors.
The dates involved, the senior licensee representative contacted, and subjects discussed were as follows:
Senior Licensee Date Subject Representative Present January 5, 1981 Escort requirements, Detail 8 Assistant Station Super-intendent January 9, 1981 Routine discussions Station Superintendent January 16, 1981 Routine discussions Station Superintendent January 23, 1981 Routine discussions Station Superintendent January 27, 1981 Scram setpoints, Detail 3 Reactor Engineer January 30, 1981 Routine discussions Station Superintendent February 13, 1981 Summary of findings Station Superintendent J
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Attendance at Manacement Meetings Conducted by Region-Based Inspectors-The resident inspectors attended entrance and exit interviews by region-based inspectors as follows:
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Inspection, Reporting Date Subject Report No Inspector January 12, 1981 Environmental Protection 277/81-02 & 278/81-02 T. Jackson (Entrance).
January 12, 1981 Chemistry (Entrance)
277/81-01 & 278/81-01 J. Kottan
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January 14, 1981 Environmental Protection 277/81-02 & 278/81-02
.T. Jackson
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January 15, 1981 Chemistry (Exit)
277/81-01 & 278/81-01 J. Kottan
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