IR 05000317/1981018
| ML20033A494 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 11/02/1981 |
| From: | Architzel R, Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20033A485 | List: |
| References | |
| TASK-1.A.1.3, TASK-1.C.6, TASK-2.E.4.2, TASK-2.K.3.01, TASK-2.K.3.05, TASK-TM 50-317-81-18, 50-318-81-17, NUDOCS 8111250503 | |
| Download: ML20033A494 (20) | |
Text
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' DCS Nos. : ' 50320 790328 50317 810723 50317 810826 50318 810703
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50317 811005 50317 810810 '50318 810919 50318 810810 50317 810915 50317 810803 150318 810924 50318 810813 50317 810830 50317-810818 50318 810821 50318 810821 50317 810916-50317 810815 50318 810804 50318 810820 50317 810721-50317 810813 50318 810802 U.S. NUCLEAR-REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT Region I 50-317/81-18 Region No.
50318/81-17 50-317 Docket No.
50-318 DPR-53 Category C
License No.
DPR-69 Priority
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Licensee:
Baltimore Gas and Electric Company P. O. Box 1475 Baltimore, Maryland 21203 Facility Name:
Calvert Cliffs Nuclear Power Plent, Units 1 and 2 Inspection At:
Lusby, Maryland Inspection Conducted:
Aug. 18-Oct. 6, 1981 Inspectors:
f.C. k Col h. 4 s l219 R. E. Architzel, S nior' Resident Inspector date signed Approved By:
P. C. A O h
- l 1t. lti E. C. McCabe, Jr., ' Chief, Reactor Projects date signed Section 2B
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Inspection Summary:
Aug. 18-Oct. 6, 1981 (Report 50-317/81-18; 50-318/81-17)
Areas Inspected:
Routine, onsite, regular and backshift inspection by the resident inspector (49.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, Unit 1; 51 hours5.902778e-4 days <br />0.0142 hours <br />8.43254e-5 weeks <br />1.94055e-5 months <br />, Unit 2)
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of:
the control room and accessible portions of the auxiliary, turbine,
- service, and intake buildings; radiation protection; physical security; fire protection; plant operating records; TMI Action Plan Items; maintenance; surveillance; Plant Operations and Safety Review Committee Activities; Radioactive Waste Releases; open items; and reports to the NRC.
Noncompliances:
Two:
Failure to follow the procedure for a liquid waste release (Detail 7); and failure to include isolation gates on Unit 2 Salt Water P&ID (Detail 3).
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DETAILS
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1.
Persons Contacted The following technical and supervisory level personnel were contacted:
G. E. Brobst, General Supervisor, Chemistry J. T. Carroll, General Supervisor, Operations S. E. Cherry, Principal Chemistry Technician R. E. Denton, General Supervisor, Training / Technical Services C. L. Dunkerly, Shift Supervisor W. S. Gibsen, General Supervisor, Electrical & Controls J. E. Gilbert, Shift-Supervisor R. P. Heibel, Principal Engineer, Technical Support J. R. Hill, Shift Supervisor W. C. Holston, Engineer, Electric Engineering Deparment D. W. Latham, Principal Engineer,.0perational, Licensing &
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Safety Unit J. F. Lohr, Shif t Supervisor R. O. Mathews, Assistant General Supervisor, Nuclear Security G. S. Pavis, Engineer, Operations J. E. Rivera, Shift Supervisor-P. C. Rizzo, Assistant General Foreman, Maintenance-L. B. Russell, Plant Superintendent J. Sites, Supervisor, Instrument Maintenance
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Unit 1 R. W. Talley, Jr., Assistant Genera 1' Foreman, PMD J. A. Tiernan, Manager, Nuclear Power Department C. Yoder, Engineer, Electric Engineering Department D. Zyriek, Shift Supervisor
Other licensee employees were also contacted.
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2.
Licensee Action on Previous Inspection Findiags (Closed) Unresolved item (317/81-15-03; 318/81-14-02); Senior Reactor Operator Limits' of Travel. The licensee revised CCI-1408, Shift Staffing, on 9/3/81 to delete the Shift' Supervisor's Office from the limits of ' travel for the Senior Reactor Operator required to be in the Control Room.
(0 pen) Unresolved Item (317/81-07-03; 318/81-07-01) Status of Full Flood Halon Systems. The Halon Fire Suppression Systems in the-Unit-1 Cable Spreading Room,. required by license to~be installed l
and tested by October 1,1981, has not achieved design concen-trations of Halon in-some areas of the Cable Spreading Room and adjacent-Cable Chase.1C. The licensee's Responsible Design Organization (RDO) reviewed the installed system, latest test results, modifications,-
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~3 and controls to mitigatea fire in the CableLSpreading' Room.
.Their evaluation, documented.in an October 1,1981 memorandum to the Plant Superintendent, concluded that the Halon 1301 Fire Suppression System would aaequately suppress potential fires. The memo stated that further modifications and tests were scheduled to improve performance and reliability for contractual reasons.
Inspector' review of testing'(August 12, 1981 full discharge test) showed that the. worst results were at Test-Point 6 (TP-6), the highest point.in'the vertical cable chase. At TP-6, halon concentration was sustained above-7% for 50 seconds, above 5% for 2.5 minutes, and above 3% for 7 minutes. The RDO determination of acceptability to suppress a fire referenced the expected nature of the potential. fires (surface-versus deep, due to early detection), licensee administra> ve procedures for controlling fire hazards, fire brigade responses, and results of historical fire tests (which showed that 3%
halon would extinguish flaming and put out a cable tray type fire). H e inspector concluded that the RDO had cdequately addressed the ability of the installed system to extinguish fi re s.
The NRR -0perating Reactors Project Manager was contacted, this finding was discussed, and a copy of the licensee's evaluation was forwarded to NRR. This item remains open pending completion of the licensee's implementation of the license requirement (Unit 2 systems scheduled for October 15, 1981).
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(Closed) Noncompliance (317/81-02-03), Restriction of Auxiliary Feedwater Flow (This it a, was reinspected and changed to an Item of Noncompliance in Ir.spection Report 317/81-04). The licensee initially responded to this item in a letter dated August 5, 1981, stating a disagreement with the inspection-findings that a violation of a Limiting Conciticn for Operation existed. They justified this contention on the basis that the (reduced) flow rate which the design analysis required was available.
The letter further stated that this flow rate had been determined to be sufficient to remove decay heat under the most adverse conditions and maintain the HOT STANDBY mode; that if additional flow was required to initiate a cooldown the operator could open the valve subsequent to removing the chain.
The original response continued:
"The Safety Analysis for loss of feedwater flow incidents from the Final Safety Analysis Report only takes credit for auxiliary feedwater system removing residual decay heat.. No credit is taken for the excess flow required and designed into the system to allow for cooldown using the auxiliary feedwater system as described in the Standard Technical Specification Bases 3/4.7.1.2."
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The inspector questioned the licensee's response regarding the design basis of the Auxiliary Feedwater System. The inspector reviewed FSAR section _10.2.3, Auxiliary Feedwater System and noted the following statements:
"The Auxiliary Feedwater System...is designated to provide feedwater for the removal of sensible and decay heat, and to cool the prin.ary system to 300'F in case the main-condensate pumps are inoperative due to loss of normal electric power sources or the main feed pumps are inoperative....
At the low level alarm point, the condensate storage tank provides 300,000 gallons of water for decay heat removal and cooldown of both units. By adjusting the feedwater flow to the permissible cooldown rate of 100'F per hour, decay heat removal and cooldown of both units can be accomplished in six hours....L" The inspector also reviewed FSAR Chapter-14, Safety Analysis, Section 14.10 Loss of Feedwater Flow Incident. The licensee
stated that this section formed the basis for the statement in the initial response regarding capability to remove decay heat.
"... An' auxiliary feedwater system is available to provide
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sufficient feedwater flow to remove residual-heat generation
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from the reactor coolant system following reactor trip from full power...."
The analysis was only performed for the initial 80 seconds
following a loss of feedwater event, so that cooldown was not a parameter of concern in the referenced analysis.
The licensee acknowledged the inspc.ctor's questions and concerns regarding the accuracy of the initial response and stated that an additional response would be sent.
The licensee responded again to this item in a letter dated September 2, 1981. The stated purpose of this letter was to clarify certain inadequacies in the initial reply and respond to additional questions raised by the Resident Inspector. The letter addressed in detail various aspects of the design basis for the Auxiliary Feedwater System flow, communications problems encountered both internal to the licensee and in their dealings -
with the NRC, the lack of completion of a particular section of the FCR Safety Analysis, that a test had not been required to verify system-flow following flow path modification,~that the Technical Specification basis.were unclear, and that Special Instructions accompanying FCR 79-1 35 were incomplet n
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The inspector reviewed the licensee's corrective actions as stated in the letter dated September 2, 1981. These actions included a memorandum to the licensee's Responsible Design Organization (RD0) (dated January 23,1981) regarding appropriate use of Special Instructions implementing FCRs, counseling of.
the individuals involved regarding completion of all parts of the Safety Analysis (inspector verified by questioning RDC personnel), removal of the -flow restriction, testing of the actual flows prior to removal, incorporating a requirement-in CCI 118 (proposed Change E), Reporting Requirements to require review by the RDO of Licensee Event' Reports when the design basis or Safety Analysis is referenced, and request a change to the Technical Specification basis clarifying the Auxiliary Feedwater System design basis.
Several of the delineated actions had previously been inspected by the NRC and the
. remainder were verified du~ ring the current inspection.
The inspector reviewed calculation sheets which had been.used by the licensee's RDO during the performance of the original Safety Analysis. One set of these sheets'was dated Aoril 24, 1980, and the other set was reconstructed by the independent reviewer and contained minor modifications to the final calculations.
The inspector noted that the licensee had perfcrmed these calculations to verify that numbers provided by the vendor for Residual Heat Removal ~1n fact included worst case Decay Heat Removal and a cooldown per the FSAR within six hours (450 gallons per minute stated by the vendor).
The inspector noted that the licensee has performed subsequent calculations for the Auxiliary Feedwater System Flow rate pursuant to TMI Action Plan Item II.E.1.1.
Preliminary calculation shown to the inspector (performed during the summer of 1981)
showed a flow rate of 365 gpm nec2ssary for the design basis.
The inspector independently calculated heat removal capacity at the 9 minute point (The FSAR allows 11 minutes prior to loss of natural circulation capability to initiate AFW). The following par.ameters were used:
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Rated Power 2700 MW Decay Heat Rate ASB9-2 1.92%
(From NRC BTP with 20% uncertainty added)
l Sensible Heat Rate 1.63 MBTV/ degrees F RCS Pumping Input 9.8 MBTU/hr The inspector roughly calculated that a 28 degree F/hr cooldown was attainable at the 9 minute point, a value consistent with
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the licensee's analysis.
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-The inspector concluded that the flows available during the.
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AFW system modification FCR 79-1035 were within the design basis and had been analyzed prior to implementation by the RDO. Various unsatisfactory aspects of this change, as addressed earlier, had been identified and corrected. The inspector stated that the NRC would change item 81-02-03 from a Violation of a Limiting Condition for Operation and close-the item.
3.
Review of Plant Operations A.
Daily Inspection The inspector toured the facility to verify proper manning and access control, and observed adherence to approved procedures and LCOs. -Instrumentation and recorder traces were observed.
Sta+.us of control room annunciators was reviewed. Nuclear instrument panels and other reactor protective systems were examined. Control rod insertion limits were verified.
Containment temperature'and pressure indications were~ checked against Technical Specifications.
Stack monitor recorder traces were reviewed for indications of releases. Panel indications for onsite/offsite emergency power sources were examined for automatic operability.
Control room, shift supervisor, and tagout log books, and operating orders were reviewed for operating trends and activities. During egress from the protected area, the inspector verified operability of radiological monitoring equipment and that radioactivity monitoring was done before release of equipment and materials to unrestricted use.
These checks were performed on the following dates:
August 19, 21, 24, 31; September 2, 8, 9, 11, 14, 15, 17, 21, 22, 24, 28, 29; October 1 and 5.
B.
Weekly System Alignment Inspection Operating confirmation was made of selected piping system trains. Accessible valve positions in the flow path were verified correct.
Proper power supply and breaker alignment was verfied. Visual inspections of major components were performed. Operability of instruments essential to system performance was verified.
The following systems were checked.
No. 11 Containment Spray Train, P&ID M-74 (Rev. 24 dated
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8/10/81), checked 9/2/81.
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Unit 2 Salt Water System, including Service Water Cooling Flow Paths, P&ID's M-450 (Rev. 7 dated 9/24/76) and OM-450 (Rev. 1 dated 12/10/76), checked 9/10/81.
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Unit 1 - Auxiliary Feedwater System, P&ID'M-59 -(Rev.17
- dated 7/16/81) and P&ID OM-39 (Rev. 9 dated 6/26/81),
checked 9/14/81.
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Unit 1 Containment Hydrogen Purge, P&ID OM-65, Sheet 2 (Rev. 1 dated 9/24/76), checked 9/22/81.
Fuel Oil Storage System, P&ID OM-79 (Rev. 3 dated 7/14/81),
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checked 10/1/81.
Findings:
1.
The Unit 2 Salt Water System isolation (sluice) gates installed in parallel (2 for each Salt Water Pump, each to a separate intake set of trash racks and traveling screens) in the suction path to the pumps were not included on the system's Piping and Instrumentation drawings M-450 and OM-450.
These gates can isolate cooling water flow to the salt water pumps (and hence to the ultimate heat sink) and are operated at the intake area, without remote position indication.
Failure to incorporate the described flow path isolations in drawings violates 10 CFR, Appendix B, Criterion III, Design Control, which requires translation of applicable design bases for systems such as the Salt Water Cooling System (10 CFR 50 Appendix A General Design Criteria 44, 45 and 46) into drawings (318/81-17-01).
2.
Checks of the Auxiliary Feedwater System and Fuel Oil Storage System showed that some locked valves were not identified as locked _open or closed on the system P&ID's.
Other valves were locked in positions opposite to that specified on the drawings, and others listed as locked on the drawings were not locked.
In all cases, the inspector verified the systems were aligned as per the System Operating Instructions and locked in the positions required by the OI's, unless deviations were properly authorized.
The licensee stated that Operations personnel do not rely upon the system lineups (and lock status) on the drawings-but use the instructions. The inspector expressed concern about the inaccuracies associated with the locked valve
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status. The licensee stated that P& ids would be revised to either accurately describe locked valve status or delete the status from the drawings. This item is unresolved (317/81-18-01; 318/81-17-02)
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Biweekly Inspection Verification of the following tagouts indicated the action was properly conducted.
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No. 1522405, Unit 2 Salt Water System, checked 9/8/81.
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No. 12540, 12 EDG Air Start Compressor, checked 9/24/81.
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Boric acid tank samples were compared to the Technical. Specifications.
Tank levels were also confirmed.
Tank No.
Conc.%
Date Level (")
Date
7.5 9/5 139 9/9
7.2 9/4 142 9/9
7.4 9/6
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7.1 9/24 126 9/24
7.1 9/24 142 9/24-21 7.1 9/24 119 9/24
7.0 9/24 130 9/24 D.
Other Checks During plant tours, the inspector observed:
shift turnovers; security p'ractices at vital area barriers; and completion and'
use of radiation work. permits, protective clothing and-respirators.
The use and operational status of. personnel monitoring practices, and area radiation and air monitors were reviewed.
Equipment tagouts were sampled for conformance with TS LCOs.
Plant-housekeeping and cleanliness was evaluated. Other TS LCOs, including RCS Chemistry and Activity, Secondary Chemistry and Activity, watertight doors, and remote-instrumentation were checked.
During tours of the Auxiliary Building, the inspector and NRR Project Manager noted a need for improvement in general housekeeping. The' licensee has been actively pursuing modifications (mostly fire protection) for an extended period. The inspector stated;that this condition would be routirely examined during future inspections and did not open a specific item on this
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area. The licensee acknowledged the inspector's comments.
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One area was of more specific concern to the inspector. The Unit 1 Containment Tendon Access Gallery, accessible from the >
45 foot level of the Auxiliary Building, had standing water
'(up to 2-3 inches deep) around a large part of its circumference.
This was a potential personnel safety hazard because the water was ' oily (from Tendon grease drippings) and the concrete floor is smooth. No sump pumps are installed in the~ gallery sumps.
No drainage channels divert water.to the sumps. The inspector-reviewed an October 2,1981 survey of the Unit I gallery. No swipes were taken due to the standing water. When questioned, the licensee: stated that the water had been sampled numerous
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' times in the-past.and the routine source had been shown to be rain-(the-Radiation Control Foreman stated that the Unit:2 gallery had once received contaminated water from a relief'
valve on the 45 foot elevation). The licensee further stated
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that the water was always sampled prior to pumping into.55 gallon drums on the 45 foot elevation (a portable pump is used), that Tendon Access Gallery water would be sampled for radioactivity and that this check would be included in future (quarterly) surveys if ' dry' swipes.were not possible.
In addition, the licensee stated that the general condition (of the standing water)~would be investigated. This item is unresolved (317/81-18-02; 318/81-17-03).
4.
Surveillance Testing-The inspector observed parts of tests to verify:
performance in; accordance with approved procedures; LCOs were satisfied; test results (if completed) were satisfactory; removal and restoration of equipment were properly accomplished; and that deficiencies were
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properly reviewed and resolved. The following tests were reviewed.
STP 0-7-1, ESFAS Logic and Performance Test, Rev. 18L,
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observed 9/14/81.
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TSP 56, Rev. O, 45' Level Unit 2 Switchgear Room Halon System
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Operational Test, observed 9/30/81.
The test achieved only 2%
Halon at a sample point high in the room. This problem.was previously identified.
The Halon Systems had not been declared operational by'the licensee.
Inspector follow item 317/81-07-03; 318/81-07-01 remains open on this item.
No violations were identified.
5.
Plant Maintenance The inspector observed and reviewed maintenance and problem investigation activities to verify:
compliance with regulatory requirements, including the Technical Specifications; compliance with administrative and maintenance procedures; compliance with codes and standards; required QA/QC involvement; proper use of safety tags; proper equipment alignment and use of jumpers; personnel qualifications; radiological controls for worker protection; fire protection; retest requirements; and reportability per Technical Specifications.
The following activities were included.
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MR M-81-2640, Remove and Overhaul No. 22 Salt Water Pump, observed on 9/8/81.
PMS 1-24-M-Q-1, Preventive Maintenance on No. 12 Diesel-
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Generator Air-Start Compressor and MR 0-81-3811, Repair Unloader Valve, observed on 9/24/81.
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. simil a r. concerns. The Plant Superintendent. issued a September
,4, 1981 memorandum to.all'plantzpersonnel' cautioning those
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not' actively-involved in accident investigation to stay clear-of accident areas.
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Unit'l tripped from 90 percent power at 7:25 p.m., 9/15, due-to loss of condenser vacuum. The loss occurred when Circulating Water Pump CWP-15 spuriously tripped while CWP-16 was off for investigation of a condenser tube leak. During restart, another trip occurred at 11:05 p.m..due to' low steam generator
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level. Also, pressurizer level was outside program level (deviation greater than 5 percent is ' normal during reactor startups). The inspector reviewed various recorder charts for-these trips, interviewed Control Room' operators, and reviewed the' sequence of event printouts and alarm typewriter results.
Safety equipment functioned as required.
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A momentary loss of No. 11 4KV feeder breaker occurrid at 6:15-a.m. 9/14. The_ loss was caused by testing the wrong undervoltage (UV) coil during surveillance.
The bus was repowered by a Diesel Generator automatic start.
Pressurizer level dropped
. to 200 inches (a 15 inch drop) while Charging Pump 11 was
'deenergized. The-inspector reviewed control room recorders
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and discussed the sequence of events with the licensee.
Safety Equipment functioned as required.
The Unit 2_ Reactor tripped at-3:09 a.m., 8/21, due to a loss
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of main feedwater and low steam generator water levels (SGWLs).
One feedwater pump was running at 60% power. It tripped when a third condensate pump was started in preparation for starting a second feedwater pump. Auxiliary feedwater was used to refill the steam generators.
To prevent water hammer, use of main feed was not allowed. SGWLs were on scale about one hour later. The inspector reviewed post-trip logs and instrumentation recorder charts for reactor power and SGWL.
Safety systems performed as designed.
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On Aug. 19, 1981 a security officer was taken to Calvert Memorial Hospital after spraining his back while lifting a gas cylinder.
Radioactive materials were not involved. Minimum shift staffing was maintained.
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At 12:44 a.m., 8/30, the unit 1 reactor tripped from full power when a trip current breaker opened. Redundant Reactor Trip Circuit breakers were being. cycled per IE Bulletin 80-19, Failure of Mercury-Wetted Matrix Relays. The cause of the trip was a loose latch arm assembly in the circuit breaker.
During restart, pressurizer level was intermittently outside 5-percent of program allowed per T. S.
The inspector reviewed the event.
Safety equipment functioned as required.
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No viola'tions were identified.
7.
Radioactive Waste Releases a.
Records and sample results of the following-liquid and/or gaseous radioactive waste releases were reviewed to verify conformance with regulatory requirements prior to release.
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Liquid Release R-072-81, release of Reactor Coolant Waste Monitor Tank 12 on 9/14/81,1.49 E-02 curies estimated.
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Liquid Release R-077-81, release of Reactor Coolant Waste -
Monitor Tank 12 on 9/30/81, 9.51 E-3 curies estimated.
The purging sampling, counting, analysis, and conduct of this release was observed.
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Gaseous Release G-052-81, No. 12 Waste Gas Decay Tank, Isolated 8/31/81, released 9/12/81. Group I Release Rate, 29,500 m'/sec; Group II, 0.
b.
The RMS response for G-052-81 had been predicted to be 1.7 E-6 CPM based upon isotopic analysis and sensitivity of Gaseous Waste Discharge Monitor 0-RE-2191. The calculation was performed to RCP l-604 and included a 50% factor, following addition of the predicted responses to background, to indicate a release.
was occurring at a rate greater than predicted.
The monitor-background was noted as being 4000 cpm. An RMS reading of 17,000 cpm will shut the gaseous waste discharge valves.
Because the release was projected to be greater than the Group I Administrative limit of 7,700 m'/sec, approval was required from the General Supervisor-Chemistry. That approval was received, with instructions to proceed if the RMS monitor did not exceed its setpoint. The licensee stated that release of.
gases at the monitor setpoint would result in site boundary concentrations of 10% of the 10CFR20 average annual release limits.
During the actual release, the RMS monitor was noted at a maximum of 5,000 cpm.
The inspector questioned the discrepancy between predicted and actual RMS response, and verified that the calculations had been performed in accordance with RCP l-604, Revision 0,
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i Gaseous Waste Releases. The licensee stated that they were investigating the discrepancy and that they apparently had been supplied incorrect RMS response data by Westinghouse, the manufacturer of the Radio Gas Detector. That data was incorporated in.the latest revision of RCP l-604. The licensee reviewed historical releases and RMS responses and determined that the detector response was closely-approximated by the detector response curves contained in the Westinghouse Radiation Monitoring System Technical Manual (latest figure supplied by a Bechtel
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letter to BG&E dated October 9, 1975). The licensee stated that the referenced curves would be used for future RMS response determinations pending receipt of revised / validated information from the vendor.
'The inspector also questioned the detector's background reading. The response curve indicated that, for a 4,000 cpm reading, radiation levels of 10 R at the detector were required for the background. reading. The licensee stated that the detector's background was not that high, but the reading might be caused by internal (fixed) contamination or detector aging.
The inspector stated that this was unresolved (317/81-18-03; 318/81-17-04) pending receipt of accurate monitor response instrumentation from the vendor, determination of the cause of the high background reading, and completion of corrective action, if required.
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observation of the sampling and release of Reactor Coolant Waste Monitor Tank 12 on-September 30, 1981. The procedure used was RCP l-601, Radioactive Liquid Waste Permits, Revision 0, dated September 1, 1981. The inspector noted that the POSRC had declined to review the subject procedure. The inspector discussed with the licensee the requirements of the Technical Specifications and Regulatory Guide 1.33 for POSRC review of procedures' dealing with-sampling and release to the environment of radioactive liquid vastes. Additional and generic implications were addressed following the completion of the inspection period.
The inspector stated that this item (317/81-18-04;-318/81-17-05) was unresolved and would be further pursued.
The inspector noted that the technician doing the sampling did not perform manual calculations of the release as directed by the procedure, but used computer program EARS-PREP which the licensee developed to obtain isotopic to MPC (10CFR20) ratio summations to ensure releases are within T.S. limits. That program also calculated predicted RMS response. The inspector verified correct input of data into EARS-PREP. During the actual release, the inspector questioned the Control Room Operator about whether a setpoint change had been entered into the Liquid Waste RMS computer alarm point. The operator stated that he had not entered such a change and was unaware of the procedural requirement. The inspector noted that the Operations Notes section of the Releash Permit did not contain directions to effect such a setpoint change. The release permit contained a maximum RMS count of 1574.4, the actual setpoint was, coincidentally, close at 1560 cpm. The inspector hand calculated the RMS response per the procedure and noted that the predicted response was 1481 counts.
(Including a 50% safety margin and adding
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isotopic response to background - 900 counts - such that the identified difference represented a'significant difference in the. hand calculated isotopic response and the computer calculated one, 87 versus 149 cpm, respectively.)_ Because the computer and hand calculations were based on identical numbers, the inspector questioned the software. The-licensee stated that the software was not reviewed and the technician had errored in using computer calculated RMS response. The licensee further stated that the' software was to be reviewed and appropriate ~
procedural steps included if computer calculations were to be performed. The inspector stated that these failures to follow procedures were a-violation. (317/81-18-05; 318/81-17-06).
8.
Observation of Physical Security The resident inspector checked, during regular and off-shift.
hours, on whether selected aspects of security met regulatory requirements, physical security plans, and approved procedures.
a.
Security Staffing l
Observations.and personnel interviews indicated'that a
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full time member of the security organization with authority to direct physical security actions was present, as required.
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Manning of all three shifts on various days was observed to be as required.
b.
Physical Barriers Selected barriers in the protected area (PA) and the vital areas (VA) were observed. Random monitoring of isolation j
zones was performed. Observations of truck and car searches i
were made.
c.
Access Control Observations of the following were made:
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Identification, authorization, and badging;
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Access control searches; l
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Communications; Compensatory measures when required.
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No unacceptable conditions were identified.
9.
Emergency Assembly Drill-Theinspectorobservedandparticipateb(proceededtotheControl Room, as directed by NRC Emergency Plans) in an Emergency Assembly Drill on October 2,1981.
Licensee personnel wereLassembled in
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required fashion (Operations Personnel and the Plant Superintendent in the_ Control Room).
No unacceptable conditions were identified.
10. Plant Operations'and Safety Review Committee-(POSRC)
I Onsite review committee functions were reviewed ~against Technical Specification Section 6 and Calvert Cliffs Instruction CCI-103D, Organization and Operation of the Plant Operations and Safety Review Committee, dated June 1, 1980. That review included all minutes of POSRC meetings between June 24,_1981 (Meeting 81-85) and July 27,r1981 (Meeting 81-99) and Memoranda to' File dated
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June 10,1981, Subject: Alternates to the POSRC appointed by the_ Chairman. The inspector noted that:
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The POSRC met more than the required once a month.
b.
Quorum requirements (the Chairman or his alternate plus
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four members) were met.
Changes in the POSRC membership,.
including alternates, were documented in the meeting
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minutes.
c.
Technical Specification Changes, Licensee Event Reports, Facility Changes, Surveillance Tests, Calvert Cliffs Instructions, and Alarm Manual Changes were reviewed by POSRC.
s The inspector observed a regularly scheuled (Wednesday) POSRC meeting on August 19,1981 (Meeting No'.81-108), when POSRC reviewed
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Licensee Event Reports, Surveillance Test Results, Technical Specification
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-Change Requests, Facility Changes, Maintenance Actions, Alarm
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Manual Changes, Administrative and Op2 rational Procedure Changes, and minutes from previous POSRC meetings (which were also approved).
L Membership and quorum requirements were satisfied. A formal POSRC determination that items reviewed did not constitute an unreviewed safety questions was made.
No unacceptable conditions were identified.
11.
a.
Inspection Review of Licensee Event Reports (LERs)
LERs submitted to NRC:RI were reviewed to verify that the details were clearly reported, including accuracy of the
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description of cause and adequacy of corrective action. The
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from the licensee,;whether-generic. implications wereLindicated,
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LERs were reviewed:
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LER No.
Date of Event Date of Report:
Subject
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Unit 1 81-58/3L'
07/21/81 08/21/81
- 12 EDG'TO #21 4KV BUS INOPERABLE.
81-59/3L 07/23/81 08/18/81 SNUBBER INOPERABLE.'
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81-60/4T~
08/10/81 08/24/81 REGINERATION WASTE
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LEARAGE DISCHARGED TO HAY VIA WNT r
DRAIN VALVE.
SI-61/3L 08/03/81 09/01/81 PLANT COMPUTER INOPERABLE.
- 81-62/3L 08/18/81 09/16/81'
115 VOLT POWER
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SUPPLY TO HYDR 0 GEN ANALYZER INOPERABLE.
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81-63/3L 08/15/81 09/10/81
. COMPONENT COOLING HEAT EXCHANGER SALTWATER OUTLET-
-VALVE INOPERABLE.
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81-64/3L 08/13/81-09/11/81 DG LOST
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SPEED CONTROL AND SEVERAL CONTROL t
ROOM INDICATIONS
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81-65/3L 08/26/81 09/25/81 INCORE MONITORING SYSTEM INOPERABLE.
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i 81-66/3L 08/26/81 09/25/81 CONTROL-ELEMENT ASSEMBLY #1 DROPPED INTO CORE.
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Unit 2 81-36/3L 08/04/81 09/03/81 RPS CHANNEL B HI POWER Tus
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BYDASSED; NI POWER SPIKES CAUSING RPS TRIPS.
81-37/3L
.08/02/81 09/01/81 CONTAINMENT PARTICULATE MONITOR INOPERABLE.
81-38/3L 07/30/81 08/28/81
-#21 EDG INOPERABLE.
81-39/3L 08/10/81 09/09/81-RPS CHANNELID HI-TUs WERE BYPASSED; ERRATIC TC INDICATIONS WERE INVESTIGATED.
81-41/3L 08/13/81 09/11/81 PLANT COMPUTER INOPERABLE.
81-42/3L 08/21/81 09/04/81
' PRESSURIZER LEVEL DEVIATED FROM PROGRAM MORE THAN
+/ 5%.
81-43/3L 08/20/81 09/16/81 RPS TRIP UNITS BYPASSED FOR TROUBLE ON Th-INPUT.
No unacceptable conditions were_ identified.
b.
For LERs selected for onsite review, the inspector verified that appropriate corrective action was taken or responsibility assigned and that continued operation was conducted in accordance with Technical Specifications and did not constitute an unreviewed safety question as defined in 10CFR 50.59.
Report accuracy,
. compliance with current reporting requirements and applicability to other site systems and components were also reviewed.
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LER 317/81-50. This LER addressed a reactor shutdown on June 14,'1981 due to RCS' gross leakage of 16 gallons per minute. The cause was a packing leak on Reactor Coolant Pump 22A differential pressure transmitter root valve.
Leakage was reduced by capping the leakoff line, and the valve was repacked. The LER concluded that the cause was natural end of life packing failure and that no further corrective action was necessary. The inspector questioned the appropriateness of the valve repacking interval. The licensee stated that no formal. program existed to repack susceptible valves, however, the charging / letdown flow control valves were repacked every refueling outage. The licensee further stated that a Facility Change was being investigated to eliminate packing on certain (up.to 140)
critical valves by installing hermetic seals. And,.the
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licensee stated that a formal repacking program would-be established during the next refueling outages.
The valves to be included would be determined after a determination of which valves receive hermetic seals. The inspector stated that establishment of a valve repacking program would be followed by the NRC (317/81-18-06).
12.
Licensee Action on NUREG 0660, NRC Action-Plan Developed as a Result of the TMI-2 Accident The NRC's.0ffice of Inspection and Enforcement has inspection-responsibility for selected action-plan items. These itens have been broken down into numbered descriptions-(enclosure 1 to NUREG
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0737, Clarification of TMI Action Plan Items).
Licensee letters containing commitments to the NRC were used as the basis for acceptability, along with NRC clarification letters and inspector
' judgment.
The following items were reviewed.
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I.A.l.3 (2) - Minimum Crew Size. One aspect of this item
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concerning limits of travel of a Senior Reactor Operator in the Control Room had previously been inspected. The licensee has revised CCI-1400, Shift Staffing, dated
September 11, 1981,. to delete the Shift Supervisor's -
Office from the SRO's surveillance. area. The inspector I
reviewed the revised procedure and verified licensee personnel were cognizant of the additional restriction.
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II.K.3.la - Auto PORV Isolation-Design. The inspector reviewed an August 11, 1981 letter from the. licensee to NRR, stating that automatic PORV isolation is not necessary and that failure of-such c system would present serious challenges to plant safety. The licensee identified steps to assure safe operation of the PORVs, including
accurate indication of PORV status, supply of emergency power to the PORVs and associated block valves, and
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training for plant operators in responses to indications.
This item will remain open pending NRC determination of acceptability ofLthe licensee's position.
l II.K.3.5 - Auto Trip of RCPs. By letter dated August 11,
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1981, the licensee documented the conclusion that~ installation of a system to automatically trip the reactor coolant pumps is not warranted and may degrade plant safety, because failure'of such a system could initiate a potentially severe event in addition to transients which could be aggravateC by loss of flow. -The licensee further~ stated.
.that operators have been provided training.to enable them
to identify situations where reactor. coolant pump tripping is required.
This item will remain open pending determination of acceptability of the licensee's position.
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II.E.4.2.(5) - Containment Pressure-Setpoint. The
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licensee submitted a proposed, lower setpoint.of-2.8 psig in their letter dated December 15, 1981.
Implementation-was scheduled by July 1, 1981. The inspector reviewed
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proposed Facility Change Request.81-10, initiated 9/21/81, the purpose of which was to amend Technical Specifications to reflect lowered Containment Pressure setpoints. -The
. lowered setpoints were in accordance with NUREG 0737 (1.8
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psig maximum LC0 limit + 1 psi margia to prevent spurious trips; previous settings &4.0 psig Reactor Trip; &4.75_
Inspector discussions with both NRC and licensee personnel revealed that II.E.4.2.(5)
required preimplementation review by the NRC. The NRC will review the licensee's proposed action. This item remains open.
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.I.C.6 - Verify Performance of Operating Activities.
The licensee requested a modification to their commitments regarding second, independent verification of position of-equipment important to safety. These commitments and schedules are described in Inspection Report 317/81-13; 318/81-13. The Modification was to establish uniformity between the recommendations of the INPO report of August, 1981 and commitments to the NRC. The licensee stated, to INPO, that " Existing procedures will be expanded by September.1981 to include appropriate instructions for performing an independent verification of components that are positioned'during testing or maintenance, are safety-related and have no control room indication." The inspector stated that the position was consistent with the intent of I.C.6. Item I.C.6 remains open.
13.
Review of Periodic and'Special Reports Upon receipt, periodic and special reports submitted pursuant to Technical Specification 6.9.1 and 6.9.2. were reviewed. That
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review included the following:
Inclusion of information required by the NRC; test results and/or supporting information consistency
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with design predictions and performance specifications; planned corrective action adequacy for resolution of problems; determination
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whether any information should be classified as an abnormal occurrence; and validity of reported information. The following periodic report was reviewed:
L August, 1981 Operations Status Reports for Calvert Cliffs
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No. 1 Unit and Calvert Cliffs No. 2 Unit, dated September 14, i-1981.
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14. Unresolved Items Unresolved items are matters about which more informationis:
required to determine whether they are acceptable. _ Unresolved items are discussed in-Paragraphs-2, 3 and 7 of:this report.
15. Exit Interview
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Meetings were held with senior facility management periodically during the course of this inspection to discuss'the_ inspection scope and findings. A summary of findings was also provided to the licensee at the conclusion of the report period.
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