ML19350C077
| ML19350C077 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 01/12/1981 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML19350C073 | List: |
| References | |
| 50-317-80-16, 50-318-80-15, NUDOCS 8103300362 | |
| Download: ML19350C077 (4) | |
Text
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O APPENDIX A NOTICE OF VIOLATION Baltimore Gas and Electric Company Docket Nos. 50-317 50-318 Based on the results of an NRC inspection conducted on October 1-31, 1980, it appears tMt several of your activities were not conducted in full compliance with conditions of your NRC Facility License Nos. DPR-53 and DPR-69 as indicated below.
A.
Environmental Technical Specification 2.3.A.5 requires that the equipment installed in the liquid radioactive waste system be monitored and operated to process all liquids prior to discharge when releases will exceed 1.25 curies per unit during any calendar quarter, excluding tritium and dissolved gases.
Contrary to the above, during the last quarter of 1979 the licensee had liquid effluent releases of 4.54 curies, excluding tritium and dissolved gases.
This quantity exceeded the Technical Specification waste processing setpoint of 1.25 curies; however, the miscellaneous waste evaporator which could have further reduced the effluents prior to discharge, was not operated.
This is an infraction applicable to DPR-53 and DPR-69.
B.
Technical Specification 6.8, Procedures, states in part:
"6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:
e.' ' Emergency Plan implementation."
The Site Emergency Plan Implementing Procedures state, in part, in SEPIP A,Section I.A, Initial Actions:
.. 3.
Shift Supervisor evaluate and classify the situation utilizing Section II.
4.
Control Room sound a 5 sec. burst of the emergency alarm, and notify all personnel over the P. A. System giving TYPE and LOCATION of the emergency, i.e.
Emergency; (Specifics of Casualty; (Special Personnel Instructions) Repeat the alarm, and pass the word again..."
Y
l Appendix A 2
Contrary to the above, following classification of a situation as a Plant Emergency by the shift supervisor at 4:20 PM on October 8, 1980, the alarm was not sounded and no PA system notification was made.
This is an infraction applicable to DPR-53.
C.
Facility license paragraph 2.c(2) states in part:
"The Tehnical Specifica-tions contained in Appendices A and B, as revised..., are hereby incor-porated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications."
Technical Specification 6.3, Facility Staff Qualifications, states in part:
"6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1 - 1971 for comparable positions.
ANSI N18.1 - 1971 states in part:
"4.3.2 Supervisors Not Requiring AEC Licenses.
At the time of initial core loading or appointment to the active position, a supervisor in this category shall have a high school diploma or equivalent and a minimum of four years experience in the craft or discipline he supervises."
Contrary to the above, licensee activities were conducted in violation of the facility license when a Production Maintenance Department-Modification Group Foreman was promoted to that position about July 1, 1979 without possessing a high school diploma or equivalent; in that identification of this condition by a licensee quality assurance audit conducted on September 21, 1979 did not result in corrective action or a request for license revision by October 31, 1980.
l This is a deficiency applicable to DPR-53 and DPR-69.
l 0.
10 CFR 50.72, Notification of Significant Events states in part:
1
"(a) Each licensee of a Nuclear Power Reactor licensed under Section 50.21 or 50.22 shall notify the NRC operations Center as soon as possible and in all cases within one hour by telephone of the occurrence of any of the following significant events and shall identify that l
event as being reported pursuant to this Section...
(7) Any event resulting in manual or automatic actuation of Engineered Safety Features, including the Reactor Protection System..."
l Contrary to the above a Unit 2 manual plant trip due to low condenser vacuum at 4.58 PM on October 26, 1980 was not reported until approximately 7 AM on October 27, 1980.
l Appendix A 3
This is a deficiency applicable to DPR-69.
E.
10 CFR 50.59, Changes, Tests, and Experiments states in part:
"(a)(1) The holder of a license authorizing operation of a production or utilization facility may (i) make changes in the facility as dascribed in the safety analysis report... without prior Commission approval, unless the proposed change... involves a change in the technical specifications incorporated in the license or an unreviewed safety question...(2) A proposed change... shall be deemed to involve an unreviewed safety question (i) if the probability of occurrence or the consequences of an accident or malfuction of equipment important to safety previously evaluated in the safety analysis report may be increased... (b) The licensee shall maintain records of changes in the facility... made pursuant to this section, to the extent that such changes constitute changes in the facility as described in the safety analysis report... These records shall include a written safety evaluation which provides the bases for the determination that the changes... does not involve an unreviewed safety question...
The licensee shall furnish to the appropriate NRC Regional Office..
. annually or at such shorter intervals as may be specified in the license, a report containing a brief description of such changes..
., including a summary of the safety evaluation of each.
The records of changes in the facility shall be maintained until the date of termination of the license, and records of changes in procedures...
l shall be maintained for a period of five years.
10 CFR 50.34 (b), Final Safety Analysis Report, states in part:
"Each application for a license to operate a facility shall include a i
final safety analysis report. The final safety analysis report shall include information that describes the facility... and shall include l'
the following:... (6) The following information concerning facility operation:... (iv) Plans for conduct of normal operations, including I
maintenance, surveillance, and periodic testing of structures, systems, and components."
The Calvert Cliffs Final Safety Analysis Report, Section 9.6.2.5, Gas Analyzing System, states in part:
"The gas analyzing system is used to determine the hydrogen concentra-tion of six points inside the containment and of four samples from the reactor coolant waste tanks (receiver and monitor tanks) as well as the oxygen concentration of several samples from the reactor coolant i
.and miscellaneous waste systems.
The gas analyzing system is installed in the sample room located in the Auxiliary Building (elevation at -10 l
feet) and consists of three analyzer cabinets and separate manifolds l
for the isolation valves and sample selection solenoid valves.
Two of l
Appendix A 4
the analyzer cabinets are for hydrogen measurement and include a sample pump, cooler, piping, valves, and instrumentation.
Each hydrogen cabinet panel contains one hydrogen analyzer, one multipoint recorder for recording each measured sample, one programmer for sequential and/or random selection of individual readout, and alarm contacts for activation of a master alarm in the control room.
The third analyzer cabinet is for oxygen measurement identical to the above system described for hydrogen."
Contrary to the above, the Gas Analyzing System Oxygen Analyzer has not been operated for the past several years, thereby removing a warning indica-tion of buildup of flammable or explosive gas concentrations with no 10 CFR 50.59 review and report of this change having been made, and with no deter-mination made that an unreviewed safety question does not exist.
This is an infraction applicable to DPR-53 and DPR-69.
t
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