IR 05000454/1988016
| ML20207L007 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 10/05/1988 |
| From: | Hinds J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20207K971 | List: |
| References | |
| TASK-1.G.1, TASK-TM 50-454-88-16, 50-455-88-15, GL-81-21, GL-88-05, GL-88-5, IEIN-88-001, IEIN-88-008, IEIN-88-009, IEIN-88-1, IEIN-88-8, IEIN-88-9, NUDOCS 8810170132 | |
| Download: ML20207L007 (12) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION III
r Report Nos. 50-454/88016(DRP); 50-455/88015(DRP)
L Docket Nos. S0-454; 50-455 License Nos. NPF-37; NPF-66 Licensee:
Commonwealth Edison Company Post Office Box 767 Chicago, IL 60690
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l Facility Name: Byron Station, Units 1 and 2 Inspection At: Byron Station, Byron, Illinois
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Inspection Conducted:
August 16 - September 30, 1988
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Inspectors:
P. G. Brochman
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N. V. Gilles
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P. L. Eng
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Approved By; J.
. Hi ricIs, [r.
C ief to.os. 86 M ctor Projects ection 1A Date Inspection Summary Inspection from August 16 - September 30, 1988 (Report Nos. 50-454/88016(DRP);
50-455/88015(DRP))
i Areas Inspected:
1.
Routine, unannounced safety inspection by the resident l
inspectors of licensee action on previous inspection findings; licensee event i
reports generic letter follow-up; bulletin follow-up; information notice follow-up; Three Mile Island (TMI) action item follow-up; operations summary; l
training; refueling and spent fuel pit activities; complex surveillance; maintenance / surveillance; independent inspection; operational safety; event
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follow-up; and management meetings.
2.
SIMS Issue status for Units 1 and 2: Closed GL-88-05; Open BL-88-08; Open 1.G.1.3; Open MPA-B-66.
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Results: Of the 13 areas inspected, no violations or deviations were identified in 11 areas. One violation was identified in each of the remaining
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One involved the failure to identify locations on safety-re',ated pumps areas.
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to be used for obtaining vibration measurements paragraph 13.
The other involved the failure to collect and analyze a compensatory sample as required by Technical Specification paragraph 3; how2ver, in accordance with 10 CFR 2, Appendix C, Section V.G.1, a Notice of Violation was not issued. The first
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violation, in itself, was of minor safety significance; however, the inability of the licensee to ensure that corrective actions continue to be implemented for a previously identified concern is of safety significance.
Two new unresolved items concerning implementation of the inservice testing (IST)
i program were identified.
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DETAILS 1.
Persons Contacted Commonwealth Edison Company
- T. Maiman, Vice President, PWR Operations
- J. B1tel, Nuclear Safety Manager
- R. Pleniewicz, Station Manager
- H. Bliss, Nuclear Licensing Manager
- B. Shelton, PWR Engineering Manager
- T. Joyce, Production Superintendent
- R. Ward, Services Superintendent
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- V. McIntire. Division Superintendent
- D. Winchester, Quality Assurance Superintendent
- G. Sorensen, PACS Construction Superintendent
- D. Elias, PWR Engineering Superintendent
- T. Tulon, Assistant Superintendent, Operating
- G. Schwartz, Assistant Superintendent, Maintenance
- L. Sues, Assistant Superintendent, Technical Services
- D. St. Clair, Assistant Superintendent, Work Planning
- F. Lentine, PWR Licensing Supervisor
- R. Stobert, Director of Quality Assurance, Operations T. Higgins, Operating Engineer, Unit 0 J. Schrock, Operating Engineer, Unit 1 D. Brindle, Operating Engineer, Unit 2 T. Didier, Operating Engineer, Rad-Waste
- M. Snow, Regulatory Assurance Supervisor
- R. Flahive, Technical Staff Supervisor S. Barret, Rar11ation/ Chemistry Supervisor
- P. O'Neil, Quality Control Supervisor
- W. Pirnat, Regulatory Assurance Staff
- E. Zittle, Regulatory Assurance Staff
- A. Chernick, Training Supervisor
- R. Lucas, Security Administrator
- P. Johnson, Master Instrument Mechanic
- R. Branson, Master Electrician
- E. Cremens, Master Mechanic
- R. Rhonds, Stores Supervisor
- P, Callighan, Assistant to Rockford District Manager
- A. h ow, Assistant Training Supervisor
- W. Dijstelbergen, PWR Engineering Field Supervisor
- S. Campbell, Office Supervisor
- M. Whitemore, Acting Radiation / Chemistry Supervisor
- E. Carroll, Regulatory Assurance, Braidwood
- D. Berg, Nuclear Safety Group
- D. Robinson, Nuclear Safety Group
- P. Barnes, Regulatory Assurance Supervisor, Braidwood
- R. Chrzanowski, Nuclear Licensing Administrator
- R. Stols, Nuclear Licensing Administrator
- B. Long, Maintenance Staff Supervisor
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- D. Johnson, Instrument Maintenance General Foreman
- L. Soth, PWR V.P.'s Staff
- D. Bump, Quality Assurance Inspector The inspectors also contacted and interviewed other licensee and contractor personnel during the course of this inspection.
- Denotes those present during the management meeting on August 29, 1988.
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- Denotes those present during the exit interview on September 30, 1988.
2.
Action on Previous Inspection Findings (92701)
(Closed)OpenItem(454/87033-01(DRP);455/87031-01(DRP)): The inspector identified a concern with alternating the use of the two starting batteries for the diesel-driven auxiliary feedwater pumps (1AF01PB, 2AF01PB). The licensee tests each diesel monthly by starting it; however, no guidance was provided to prevent only one battery from being used all the time, thus allowing an undetected failure of the other ba tte ry. The inspector r^ viewed the licensee's revisions to procedures
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1BVS 7.1.2.1.a-2 and 2BVS 7.1.2.1.a-2 and verified that direction is provided to ensure that both batteries are utilized and tested period-ically. Based on this review, this item is considered closed.
3.
Licensee Event Report (LER) Follow-up (90712 & 92700)
(Closed) LERs (454/87018-1L, 454/88005-LL, 454/88006-LL, 455/87015-1L, 455/88010-LL):
Through direct observation, discussions with licensee personnel, and review of records, the following LERs were reviewed to determine that the reportability requirements were fulfilled, that immediate corrective action was accomplished, and that corrective action to prevent recurrence had been accomplished in accordance with Technical Specifications.
LER No.
Title Unit 1 454/87018-1 Reactor trip caused by main feedwater pump trip due to a broken wire in the thrust bearing wear
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454/88005 Main generator / reactor trip caused by spurious actuation of nicrowave instability trip relays.
454/88006 Fuel handling building booster fan actuation due to voltage transient on offsite transmission line.
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Unit 2
455/87015-1 Feedwater isolation due to personnel error.
455/88010 Technical Specification action limit for a
compensatory samole interval not met due to personnel j
error.
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With regard to LER 455/88010, this LER describes an event on July 27,
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1988, with the reactor at 94% power, in which a Technical Specification required compensatory sample for an inoperable radiation monitnr was not
obtained within the required time limit.
The root cause of the event was
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a personnel error by a Radiation Chen.istry Technician (RCT) wno forgot to collect the sample.
The reactor containment fan cooler 2A and 2C
essential service water outlet process radiation monitor, 2PR02J, had been inoperable since 3:45 p.m. on July 21, 1988.
Technical Specification i
3.3.3.9 allows effluent releases to continue for up to 30 days provided j
that grab samples are collected and analyzed at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Samples had been collected and analyzed until 12:00 m. on July 27, when
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the sample was missed. The missed sar.ple was detected by a required j
supervisory review.
The failure to collect and analyze a grab sample
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from 4:00 a.m. to 8:00 p.m. (16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />) on July 27, 1988 is a violation of Technical Specification 3.3.3.9, Action Requirement b l
(455/88015-01(DRP)); however, this violation meets the tests of 10 CFR 2,
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Appendix C,Section V.G.1; consequently, no Notice of Violation will be
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issued, and this matter is considered closed, i
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With regard to LERs 454/87018-1, 454/88005, 454/88006, and 455/87015-1,
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no violations or deviations were identified.
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4.
Generic Letter Follow-up (92701)
j (Closed) Generic Letter (454/88005-HH; 455/88005-HH): Boric acid t
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corrosion of carbon steel reactor pressure boundary components in PWR i
plants.
This item is also identified as Safety Issues Management System
i (SIMS) issue GL-88-05. The licensee's response was submitted on May 31, l
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1988, describing its commitment to a boric acid leakage monitoring and j
preventive program at Byron.
Tne NRC staff has reviewed the licensee's l
response to this generic letter and concluded that it is acceptable, in
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accordance with a letter from L. N. Olshan to H. E. Bliss, dated l
September 2, 1988.
Based on this review, this generic letter and SIMS
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issue are considered closed.
5.
Compliance Bulletin Follow-up (92701)
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(0 pen) Bulletin (454/88008-8B; 454/88008-18; 454/88008-2B; 455/88008-88;
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455/88008-1B; 455/88008-2B):
Thermal stresses in piping cc,nnected to the
This item is also identified as SIMS issue BL-88-08.
The licensee's response to this bulletin and its two i
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supplements was submitted to the NRC on September 20, 1988, and is being
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reviewed by the NRC staff.
Pending completion of this review, this j
q bulletin and the SIMS issue will remain open.
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Informat, ion Notice _(IN) Follow-up (92701)
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(Closed) IN 88-01:
"Safety Injection Pipe Failure." This IN has been superseded by NRC Bulletin No. 88-08, "Thermal Stresses in Piping i
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Connected to Reactor Coolant Systems," which covers the same subject, i
Consequently, the licensee's actions will be tracked using Bulletin 88-08 (see paragraph 5), and IN 88-01 is considered closed.
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b.
(Closed) IN 88-09:
"Reduced Reliability of Steam-Driven Auxiliary Feedwater Pumps Caused by Instability of Woodward PG-PL Type
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Governors." This IN is not applicable to Byron, which does not
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utilize steam-driven auxiliary feedwater pumps, l
7.
TMI Action Item Follow-up (25565)
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(0 pen) TMI Action Item I.G 1: Training during low power testing. This item is also identified as SIMS issue I.G.1.3.
In a letter from R. L.
i Tedesco to J. S. Abel, dated June 10, 1981, the NRC staff concluded that natural circulation testing and training should be performed for all
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future PWR operating licenses, with descriptions of natural circulation
tests included in the Final Safety Analysis Report.
The issue of whether
natural circulation testing needs to be performed at Byron is still under
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review by NRR and is being followed under the guidance of Temporary l
i Instruction (TI) 2515/86, "Inspection of Licensee's Action Taken To l
Implement Generic Letter 81-21, Natural Circulation Cooldown," and SIMS
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issue MPA-B-66. This TI and SIMS issue MPA-B-66 are still open as l
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discussed in Inspection Report Nos. 454/88011; 455/88011.
Conse fsntly, t
TMI Action Item I.G.1 and SIMS issue I.G.1.3 remain open for Un:cs 1 and 2.
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Summary of Operations Unit 1 operated at power levels up to 98% until 1:30 a.m. on September 3, y
1988, when the unit was shut down for a scheduled 10-week refueling i
outage.
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Unit 2 operated at power levels up to 95% for the entire report period.
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An Unusual Event was declared on September 24, 1988, when both trains of
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the containment phase "A" manual actuation circuit were declared
inoperable (see paragraph 15.b).
9.
_ Training (41400)
The effectiveness of training programs for licensed and nonlicensed personnel was reviewed by the inspectors durit.g witnessing of the licensee's performance of routine surveillance, maintenance, and
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operational activities and during review of the licensee's response to i
i events which occurred during August and September 1988.
Personnel l
i appeared to be knowledgeable of the tasks being performed, and nothing
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was observed which indicated ineffective training.
l No violations or deviations were identified.
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10.
Refueling and Spent Fuel pit Activities (60710 & 86700)
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Refueling activities during the Unit 1, Cycle 3 fuel reload were f
observed / reviewed to ascertain that they,tre conducted in accoidance
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Specifications.
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l The fo11 ewing areas were considered / monitored during this review:
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periodic testing and operability of refueling equipment and systems were j
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performed per procedural requirements; fuel handling operations, including fuel assembly reinstallation and transfer of burnable poison assemblies and rod control cluster assemblies, were performed in accordance with approved procedures and Technical Specifications; plant conditions were maintained as required to support refueling; good housekeeping, loose object control, and appropriate cleanliness zones were established and enforced, and the required rai ological controls and practices were established and observed; and the fuel handling activities were performed by qualified personnel.
q The inspectors verified by observation and direct interviews that fuel handling and control room personnel were properly briefed on scheduled refueling activities and that these activities were correctly performed; that source range nuclear instruments, including audible monitors, were operable during reloading operations; that fuel material control requirements were performed; that refueling cavity and spent fuel pool water levels were maintained in accordance with Technical Specifications; that boron concentrations were established and verified as required; and that operation of the spent fuel pool bridge crane, fuel handling building (FHB) crane, spent fuel pool cooling system, and FHB ventilation and radiation monitoring systems was as required by Technical Specifications.
The inspectors identified a cor.cern relating to Technical Specification (TS) Surveillar:ce 4.9.10.
This specificat;on requires that water level be verified to be at least 23 feet above the reactor vessel flange, within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> prior to the start of and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereaf ter during movement of fuel assemblies or control rods.
The licensee did not have a separate surveillance procedure to verify this TS requirement as it does for all other TSs. Also, no entries were made in the reactor operator's log specifically regarding the refueling cavity water level within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of commencing fuel assembly or control rod movement. The licensee had been verifying and logging refueling cavity level every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> since the cavity was filled, but not within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of commencing moving fuel assemblies or control rods. As corrective i
action, the licensee has written a specific surveillance to verify water level within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of commencing movement and is referencing the
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surveillance in the appropriate fuel handling procedures. Based on the
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licensee's prompt corrective actions and the fact that water level was always greator than 23 feet above the reactor vessel flange and was being monitored every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, the NRC has decided not to impose any enforcement actions on this matter, and this matter is considered closed.
No violations or deviations were identified.
11.
Complex Surveillance (61701)
The inspector reviewed activities for the 18-month surveillance test of
the manual actuation of the Unit 1 safety injection and phase A circuits.
The following items were considered during this review: testing was accomplished in accordance with approved procedures; test instrumentation was within its calibration interval; testing was accomplished by
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qualified personnel; test results conformed with Technical Specifications
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and procedural requirements and were reviewed by personnel other than the
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individual directing the test; and any deficiencies identified during the j
t testing were properly documented, reviewed, and resolved by appropriate management personnel.
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The inspector reviewed the procedure and witnessed its performance.
l The inspector verified that the procedure was accomplished as written j
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post-test review, a technical staff engineer identified that the i
continuity of a circuit between the manual phase A switch and the
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containment vent isolation relay had not been verified. This function is only required in Modes 1 through 4.
The licensee will perform a l
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supplemental test prior to entry into Mode 4.
This condition has also
1 existed in Unit 2 since November 1987.
As a consequence, the licensee i
determined that the the circuits were inoperable in Unit 2 and declared
an Unusual Event (see paragraph 15.b).
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This event will be reviewed in a subsequent report, after the LER is l
j issued. No other violations or deviations were identified.
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17. Monthly Maintenance / Surveillance Observations (61720, 61726, 62703, & 73756)
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Station maintenance activities of the safety-related systems and
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components listed below were observed or reviewed to ascertain that they i
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and industry codes or standards, and in conformance with Technical
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Specifications.
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5-year inspection of the 18 diesel generator
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0A SX makeup pump post-maintenance run
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Unit 1 steam generator safety valve setpoint testing l
Unit I steam generator U-bend stress relieving i.
Unit 1 "A" low pressure turbine rotor removal
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Station surveillance activities of the safety-related systems and j
components listed below were observed or reviewed to ascertain that they
were conducted in accordance with approved procadures and in conformance
with Technical Specifications.
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18 diesel-driven auxiliary feedwater pump 18-month surveillance i
18-month calibration of SG 1evel channel 557 l
2A motor-driven auxiliary feedwater pump monthly surveillance j
Unit 1 steam generator eddy current inspection l
Local leak rate test of contamination penetration.
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The following items were considered during these reviews:
the limiting i
conditions for operation were met while affected components or systems i
were removed from and restored to service; approvals were obtained prior to initiating work or testing; quality control records were maintained;
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i parts and materials used were properly certified; radiological and fire i
prevention controls were implemented; maintenance and testing activities j
were accomplished in accordance with approved procedures; maintenance and s
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testing were accomplished by qualified personnel; test instrumentation
i calibrations were performed prior to returning components or systems to j
service; test results conformed with Technical Specifications and t
procedural requirements and were reviewed by personnel other than the individual directing the test; any deficiencies identified during the
testing were properly documented, reviewed, and resolved by appropriate t
i management personnel; work requests were reviewed to determine the status
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of outstanding jobs and to assure that priority was assigned to I
safety-related equipment maintenance which may affect system performance.
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While touring the auxiliary building, the inspectors noted that several
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of the pumps included in the IST program had recently been painted and I
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should be taken had been obscured. The inspectors could find no markings
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on the follecing pumps: component cooling water pumps 2CC01PA and 2CC01PB, l
safety injection pump 2SIO1PA, and auxiliary feedwater pumps 1AF01PA and 2AF01PA. The inspectors also noted that some but not all of the markings
on pumps ISIO1PA and 2AF01PB were obscured.
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10 CFR 50.55a(g) requires that certain nuclear power plant components be i
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periodically tested to verify operational readiness in accordance with
Section XI of the American Society of Mechanical Engineers (ASME) Boiler i
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and Pressure Vessel Code. One of the IST required parameters used to
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i determine pump operability is vibration.
Subsection IWP-4160 of Section j
XI requires that instruments which are position sensitive shall be either
permanently mounted, or provision shall be made to duplicate the position i
j for each test. As discussed in inspection reports 454/84038; 405/84033
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and 454/84064, the licensee previously agreed to specify the locations
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i for those pumps in the IST program where vibration was to be measured in the licensee's procedure BVP 200-1, "IST Requirements for Pumps." The licensee it.tends to inspect all pumps in the IST program and to verify
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i that proper locator markings are present or that markings are installed.
2 The failure to ensure the repeatability c,f the position of vibration
measurements on those pumps included in the IST program is t violation of Section XI of the ASME Code, Subsection IWP-4160 (454/88016-01(DRP);
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455/88015-02(ORP)).
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j Since the inspectora were unable to determine how long these locator
markings were obscured, the licensee is requested to identify when the markings were obscured.
Since the vibration measurements may not have
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been taken in the same location, the licensee is requested to evaluate
the operability of any pumps in the IST program which are determined to l
have been lacking locator markengs.
This evaluation should be included
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in the response to the violation.
Review of this evaluation will be l
followed as unresolved item (454/88016-02(DRP); 455/88015-03(ORP)).
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The inspectors also identified a concern relating to the testing of valve position indication at remote locations.
Subsection IWV-3300 of the ASME Code requires that valves with remote position indicators be observed at least once every two years to verify that accurate operation is indicated. The licensee does not perform this test for valves which indicate at the remote shutdown panel.
No request for relief from this requirement of the ASME Code has been submitted to the Commission for review. The inspectors understand that the licer.see believes that it is rneeting the intent of the Code and that this policy has been premulgated
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at the corporate level.
As this affects all of the licensee's nuclear i
units a letter from E. L. 3reenman to C. Reed, dated September 15, 1988,
was sent to the licensee to request information to resolve this issue, i
This issue will be tracked at By(ron as an unresolved item
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(454/88016-03(DRP); 455/88015-04 DRP)).
No other violations or deviations were identified.
13.
Indepcndent inspection (61726)
NUREG/BR-0051, Vol. 9, No. 3. "Power Reac, " vents," July 1988, describes events at the Callaway Nuclear P wer Station in which the comon recirculation line isolation valve (M-8813) from the safety injection (SI) pumps to the refueling water storage tank (RWST) had been shut for surveillance tests and to allow maintenance testing of valves EJ-8804A and EJ-8804B. The shutting of valve 8813 makes both trains of the emergency core cooling system (ECCS) inoperable. As the Byron emergency core cooling system (ECCS) is basically identical to the
,allaway ECCS systen, the inspector reviewed the surveillance procedures, inservice testing schedule, and electrical schematics for the associated valves at Byron (S!8813, SI8804B, and CV8804A). The IST schedule specifies that valves SI8813 and CV8804A are only to be stroke tested with the unit in Modes 5 or 6.
The ECCS system is required to be operable in Modes 1 through 3.
The surveillance procedures for CV8804A and SI88048, in lieu of closing valve S!8813, require installation of jumpers to defeat the interlock from S!8813 and thereby to allow the 8804 valves to be stroked without repositioning 518813.
Based on this review, the inspector determined that Byron's method of testing should prevent
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the problems which occurred at Callaway.
No violations or deviations were identified.
14. Operational _ Safety Verification (71707, 71709, & 71881)
The inspectors observed control room operation, reviewed applicable logs and conducted discussions with cor. trol room operators during August and September 1988.
During these discussions and observations, the inspectors ascertained that the operators were alert, cognizant of plant conditions, and attentive to changes in those conditions, and that they took prompe action when appropriate.
The inspectors verified the operability of selected errergency systems, reviued tagout records, and verified the proper return to service of affected corrponents.
Tours of the auxiliary, Unit 1 containment, fuel handling, rad-waste, and turbine buildings were conducted to observe plant equipment conditions, including
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potential fire hazards, fluid leaks, and excessive vibrations, and to verify that maintenance requests had been initiated for eqJipment in need of maintenance.
The inspectors verified by observation and direct interviews that the physical security is being implemented in accordance with the station security plan. On September 28, 1988, the licensee idertified that several openings between the protected area and a vital area existed.
These openings were greater than allowed by the licensee's Security Plan, and no compensatory measures were implemented. This event was reviewed by the physical security section and will be addressed further in a subsequent report.
The inspectors observed plant housekeeping / cleanliness conditions and verified implementation of radiation protection controls.
The inspectors also witnessed portions of the radioactive waste system controls associated with rad-waste shipments and barreling.
The facility operations observed were verified to be in accordance with the requirements established under Technical Specifications, 10 CFR, and administrative procedures.
15. Opite follow-up of Events at Operating Reactors (93702)
The inspectors performed onsite follow-up activities for events which occurred during September 1988. These follow-ups included reviews of operatin-.ogs, procedures, Deviation Reports, and Licensee Event Reports (where available), and interviews with licensee personnel.
For each event, the inspector developed a chronology, reviewed the functioning of safety systems required by plant conditions, and reviewed licensee actions to verify consisteacy with procedures, license conditions, and the nature of the event. Additionally, the inspector verified that the licensee's investigation had identified the root causes of equipment malfunctions and/or personnel errors and that the licensee had taken appropriate corrective actions prior to restarting the unit. Details of the events and the licenser's corrective actiors developed through inspector follow-up are provided in paragraphs a and b below:
a.
Unit 1 - Loss _of the 1A Train of Residual Heat Removal (RHR) While T.owering Water Level in The Reactor Vessel On September 19, 1988, with the reactor in Mode 6 with the reactor vessei head removed, operators were in the process of draining the reactor refueling cavity water level to below the reactor vessel flange to allow replacement of a reactor vessel stud hole plug which had come loose.
Level had been lowered two times using the 1B RHR pump to transfer water from the ref ueling cavity / reactor vessel to the RWST, and at 9:57 a.m. operators were lowering level a third time to allow installation of the plug.
Just as the Unit 1 Reactor Operator (RO) was about to shut down the IB RHR pump, he noticed oscillations in letdown ficw pump mctor amps. At 10:59 a.m., the R0 isolated letdown, shut down the IB RHR purp, shut down the 1A RHR pump, and complied wi+h the appropriate abnormal operating f
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procedures and Technical Specifications for loss of an RHR pump.
Corrective actions were taken t0 vent the 1A RHR train, to fill the refueling cavity, and to restorre shutdown cooling.
By 11:14 a.m.,
shutdown cooling had been restored, and by 1:46 p.m., the 1A RHR pump had been vented and rur, and was declared operable.
This event is reviewed in detail in inspection report 454/88019(DRP).
b.
Unit 2 - Unusual Event Declared when Both Trains of the.
Actuation Circuit for the Containment Phase "A" Isolation.
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Were Determined to be Inoperable On September 23, 1988, with the unit at 40% power, the licensee declared an Unusual Event and commenced an orderly shutdown in accordance with Technical Specification 3.0.3.
The cause cf the shutdown was both trains of the Unit 2 marual phare "A" isolation systen being declared inoperable.
The licensee declared both trains inoperable at 11:24 a.m., when an engineer discovered that a portion of the manual phase "A" isolation circuitry which initiates closure of the containment vent isolation valves was not verified as functioning during the last test of the manual phase "A" isolation function in November 1987. Technical Specification 3.3.2 allows only one channel of the manual phase "A" Isolation instrurentation for the containment vent isolation system to be inoperable; conse-quently the licensee entered TS 3.0.3 and initiated a shutdown within one hour to place the unit in hot standby within the next six hours. A special test procedure was written to test the portion of the isolation circuitry in question, the test was successfully completed at 1:01 p.m., tne shutdown was stopped, and the Unusual Event was terminated.
This event will be reviewed in a subsequent report after the LER is issued.
!6. Management Meetings (30702)
On August 29, 1988, Mr. C. J. Paperiello, Deputy Regional Administrator, Region Ill, and members of his staff met publicly with Mr. T. J. Maiman, Vice President, PWR Operations, Comonwealth Edison, and mer.bers of his staf f, denoted in paragraph 1 of this report.
This meeting was held to discuss licensee initiatives in the self-evaluation of its performance and in the identification of areas for improverent.
17. Unfesolved Items Unresolved itens are matters about which more information is required in order to ascertain whether they are acceptable items, violations, or deviations.
Unresolved items disclosed during the inspection are discussed in paragraph 12.
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18.
Violations for which a "Notice of Violation" Will Not Be Issued The NRC uses the Notice of Violation as a standard method for formalizing the existence of a violation of a legally binding requirement.
However, because the NRC wants to encourage and support licensee initiative in the self-identification and correction of problems, the NRC will not generally issue a Notice of Violation for a violation that meets the tests of 10 CFR 2, Appendix C, Section V.G.I.
These tests are: 1) the violation was identified by the licensee; 2) the violation would be categorized as Severity Level IV or V; 3) the violation was reported to the NRC, if required; 4) the violation will be corrected, including measures to prevent recurrence, within a reasonable time period; and 5)
it was not a violation that could reasonably be expected to have been prevented by the licensee's corrective action for a previous violation.
A violation of regulatory requirements identified during the inspection for which a Notice of Violation will not be issued is discussed in paragraph 3.
19.
Exit Interview (30703)
The inspectors met with the licensee representatives denoted in paragraph 1 at the conclusion of the inspection on September 30, 1988.
The inspectors summarized the purpose and scope of the inspection and the findings.
The inspectors also discussed the likely informational content of the inspection report, with regard to documents or processes reviewed by the inspectors during the inspection.
The licensee did not identify any such documents or processes as proprietary.
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