IR 05000213/1979011
| ML19260C256 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 09/13/1979 |
| From: | Bettenhausen L, Caphton D, Graham P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML19260C250 | List: |
| References | |
| 50-213-79-11, NUDOCS 7912260122 | |
| Download: ML19260C256 (13) | |
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U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT Region I Report No. 50-213/79-11
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Docket No. 50-213 s
License No. DPR-61 Priority
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Category
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Licensee:
Connecticut Yankee Atomic Power enmnany P. O. Box 270 Hartford, Connecticut 06101
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Facility Name:
Haddam Neck Plant Inspection at:
Haddam, Connecticut Inspection conducted: July 17-20, 1979 Inspectors:
- N L.'7f. Bettenhausen, Reactor Inspector h h, b m
I A
P. D. Graham, Reactor Inspectore date signed gj/ /ihn
' 9ll date si ned C. Co g L1 Reactfr spector Approved by:
. / c-D.'L.Caphton, Chief,Nuclearhu._pport Section No. 1, RO&NS Branch Inspection Summary:
Inspection on July 17-20,1978 (Report No. 50-213/79-11)
Areas Inspected:
Routine, unannounced inspection by regional based inspectors of post refueling plant operations; post refueling startup testing including control rods, core power, 'incore-excore detector calibration, isothermal temperaturc coefficient, shutdown margin, and axial flux difference; general employee training; anlicensed personnel training; operator requalification training and licensee action on previous inspection findings.
The inspection involved 64 inspector hours on site by 3 NRC regional based inspectors.
Results: Of the six areas inspected, two items of noncompliance were found in one area (Deficiencies - failure to review temporary procedure change and failure to record required information in plant procedures).
1616 198 Region I Form 12 919,2601 g (Rev. April 77)
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DETAILS 1.
Persons Contacted Principal Licensee Employees
- R. Blewett, Quality Assurance Supervisor
- N. Burnett, Operating Supervisor
- J. Chiarella, I&C Foreman
- J. Ferguson, Engineering Supervisor
- S. Fleming, Training Supervisor R. Gracie, Operations Assistant
- R. Graves, Station Superintendent
- C. Kokoszka, Assistant Reactor Engineer J. Levine, Maintenance Supervisor D. Packer, Training Coordinator
- R. Traggio, Unit Superintendent The inspector also talked with and interviewed other licensee employees during the course of this inspection.
They included reactor operators, technical support personnel, engineering personnel, and general office personnel.
- denotes those present at the exit interview ~on July 20, 1979.
2.
Licensee Action on Previous Inspection Findings (Closed) Unresolved Item (213/77-13-03): The licensee has incorporated into the precautions of all procedures utilizing the strip chart recorder, a step to insure that the recorder is in calibration prior to use.
I&C performs the calibration periodically and presents the calibration data to the reactor engineer.
This item is closed.
(Closed) Unresolved Item (213/78-05-01):
The licensee has revised procedure SUR 5.3-19 for Cycle IX data. This item is closed.
(0 pen) Unresolved Item (213/78-21-03): The program for formal and on-the-job training for maintenance personnel was it.ded as stated in earlier inspection reports; however, inspector inquiry has determined that the program is not yet implemented in detail for all maintenance personnel.
Licensee management agreed to complete implementation by October 1,1979.
This item will be reviewed in the course of a future inspection.
3.
Post Refueling Plant Operations The inspector reviewed normal operation procedures and surveillance procedures to determine the following:
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Systems disturbed during the refu~eling outage were returned to
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service in accordance with approved procedures; and, Preparations for control rod withdrawal and performance of control
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rod withdrawal were conducted in accordance with approved procedures.
The following completed procedures were reviewed:
SUR 5.1-1, Reactor Coolant System - Hydrostatic Test, Revision 4,
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dated November 23, 1977, completed March 9,1979.
NOP 2.13-3, Reactor Containment Access Control, Revision 1, dated
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March 19, 1979.
Completed January 4,1979 and March 22, 1979.
NOP 2.3-3, Operation at Hot Standby - Reactor Shutdown, Revision 1,
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dated October 18, 1976, completed January 27, 1979 N0P 2.1-1, Cold Shutdown to Hot Standby, Revision 5, dated November
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23, 1977, completed March 10, 1979.
NOP 2.4-1, Filling and Venting of Entire System, Revision 3,
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dated November 23, 1977, completed February 27, 1979
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f!0P 2.12-1, Core Cooling System Lineup For Shutdown and at Power
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Operation, Revision 7, dated October 6,1978, completed March 9, 1979.
.NOP 2.1-2, Reactor Startup, Revision 4, dated August 25, 1978,
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completed March 4,1979.
NOP 2.1-7, New Core Initial Critical Approved, Revision 3, dated
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February 27, 1979, completed March 3,1979.
Findings Review of N0P 2.4-1, Filling and Venting of Entire System, indicated that a temporary procedure change was implemented on February 27, 1979, by two members of the plant management, at least one of which was licensed.
Inspection and discussion with a cognizant licensee representative indicated that the required Plant Operations Review Committee (PORC) reviews and Plant Superintendent approval had not been completed as of July 20, 1979. At the close of this inspection, the licensee had begun taking steps to review and approve the temporary procedure change. This failure to properly review and approve a temporary procedural change as required by Technical Specification 6.8.3 constitutes a deficiency level item of noncompliance (79-11-01).
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Review of NOP 2.4-1 also indicated the following examples of failure to follow and complete plant procedures:
a.
Step 7.1.2.7 was not initialed by an operator to indicate that the step was completed, but instead, had "T.P.C" written in the margin.
Inspection determined that no temporary procedure change existed.
b.
Several valves had their required valve position on the procedural checklist changed without a temporary procedure change.
Review of N0P 2.1-2, Reactor Startup, completed March 4,1979, indicated that step 7.6.1 was not completed properly, in that, information required by the procedure to be documented was not recorded.
The above three examples of the licensee's failure to follow and properly complete plant procedures as required by Technical Specification 6.8.1 are considered collectively to constitute a deficiency level item of noncompliance (79-11-02).
During the review of plant procedures, the inspector noted procedures that appeared to be incomplete.
Discussions with licensee representatives determined that the numerous incomplete steps resulted from stopping and starting procedures prior to their completion because of equipment problems, for example, stopping a plant startup for some time period to make repairs and then recommencing the procedure from this new starting point. The old procedure is signed off with the reason for stopping annotated in the remarks, and then a new procedure is started at some intermediate step with all previous steps left blank. Although such cases are common, the inspector stated that all steps in a procedure should be completed to insure that the required plant status exists.
An alternative would be to annotate the procedure in the remarks section with the reason why the steps are not required and a reference made to the halted procedure. A licensee representative agreed with the inspector's comments and indicated that a review and resolution of this matter would be made.
4.
Post Refueling Startup Testing a.
A review of the startup testing program was performed to verify conduct and completion of Technical Specification required tests.
This review encompassed the following:
In-house Startup Report;
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Control rod drop times (hot and cold), and rod position
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indication checks; Core power distribution determination;
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Incore/Excore calibration and axial offset detemination;
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Determination of control rod worths;
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Isothermal temperature coefficient;
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Reactor Shutdown margin; and
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Core thermal power determinations.
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b.
The inspector verified by review of facility records in the following areas that startup testing was conducted in accordance with technically adequate procedures and that the facility is being operated within license limits.
(1) Post Refueling Startup Report Review The inspector and licensee representatives discussed the licensee's plans to submit a Startup Report to the NRC. The licensee stated that a Startup Report would not be submitted because no changes had taken place during the refueling which would require a report submittal as specified in the Technical Specifications.
The inspector verified the licensee's determination by review of the following documents that involved the Cycle 9 refueling:
BAW-1518, Nuclear Design Report, Cycle 9; and
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Fuel Cycle Plan Report Cycle 9.
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The inspector noted that the licensee's Reactor Engineer generated an in-house Startup Physics Test Report which discussed the tests performed and presented the results obtained. The inspector noted and brought to the attention of a licensee representative a calculational error in the report involving the correction of the isothermal temperature coefficient to the Hot Full Power. The inspector noted that this error did not result in any safety question.
The inspector had no further questions in this area.
(2) Control Rod Drop Times Control rod drop times, testing, and position indication verification are conducted to verify conformance with Technical Specification 4.2-1 and 4.2-2.
The licensee performed this verification in accordance with the following procedures:
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SUR 5.3-2, Hot Rod Drop Time Measurement, Revision 3,
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dated February 27, 1979, completed March 2, 1979.
SUR 5.2-21, Control Rod Exercises, Revision 1, dated
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July 22, 1976, completed March 2, 1979.
SUR 5.2-12, Rod Position Digital Voltmeter Calibration
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Revision 1, dated June 28, 1978, completed March 1, 1979.
SUR 5.3-1, Cold Rod Drop Time Measurements, Revision 5,
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dated February 27, 1979.
Completed March 1, 1979.
No items of noncompliance were identified.
(3) Core Power Distribution The procedure and method used by the licensee to verify that the plant is operating within the power distribution limits defined in Technical Specifications were reviewed and discussed with cognizant licensee personnel. The following procedures were reviewed:
SUR 5.3.17, Operation of the Flux Map System, Revision
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2, dated October 27, 1978, completed March 15-22, 1979, June 13-14,1979 and' July 18, 1979; SUR 5.3-22, Flux Map System Voltage Plateau Procedure,
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Revision 1, dated July 6, 1978, completed March 5, 1979; and SUR 5.3-24, At Power Flux Maps, Revision 2, dated
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February 27, 1979, completed March 15-22, 1979, June 13-14, 1979 and July 18, 1979.
The licensee performs the necessary core flux maps at operating conditions and transmits these maps along with supportive plant parameters to the Corporate Headquarters. The data is digitized and fed into a computer which performs the core power distribution determination using the licensee's version of the Westinghouse Incore program. The results are then reviewed and analyzed by the Corporate Reactor Engineering staff, and the plant Reactor Engineer is notified of the acceptability of the results. The inspector reviewed data and computer results from March - June, 1979 for compliance with Technical Specifications.
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During review of procedure SUR 5.3-17, the inspector noted that the procedure references and uses a data sheet, RE-DS-19, Revision 1.
Examination of completed procedures indicated that the operators are using data sheet RE-DS-19, Revision 3.
The procedure also requires the flux traces to be stamped with a rubber stamp used to record identifying data. The inspector noted that the operators are no longer using a rubber stamp because all of the necessary data is now being recorded on the new data sheet. A licensee representative acknowledge the inspector's comments and stated that the procedure would be revised and committed to revision by October 20, 1979. This item is unresolved pending procedure revision (79-11-03).
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Incore/Excore Correlation and Axial Offset Verification The incore/excore correlation is determined in accordance with procedure SUR 5.3-23, Axial Offset Verification, Revision 2, to comply with Technical Specification 3.18.
The inspector reviewed SUR 5.3-23, the completed data and flux maps, and the derived incore vs. excore plots for the period March 15-22, 1979.
The resultant incore vs. excore plots from SUR 5.3-23 are forwarded to the instrumentation and control personnel for use in calibrating the axial offset monitoring channel.
This calibration is performed in accordance with procedure SUR 5.2-48, Axial Offset Calibration, Revision 2.
The inspector reviewed the completed procedure for June 20, 1979.
Review of procedure SUR 5.3-23 indicated that the procedure referenced and used data sheet RE-DS-19, Revision 1.
Examination of the completed data sheets indicated that a new data sheet RE-DS-19, Revision 3, was being used by the operators. When informed of the discrepancy, the licensee representative committed to revising the procedure by October 20, 1979.
The item is unresolved pending procedure revision (213/79-11-04).
(5) Control Rod Worth and Boron Worth The control rod reactivity worth and ejected rod worth are determined in accordance with procedure SUR 5.3-6, Control Rod Group Reactivity Worth Combined with Boron Worth and Ejected Rod Worth, Revision 6, to meet the requirements of Technical Specification 3.10.
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Procedure SUR 5.3-6, completed March 4,1979, was reviewed to determine technical adequacy and that the measured data compared with the predicted values for control rod worth.
The following results were noted.
Control Rod Bank Measured Worth Predicted Worth (% delta k/k)~
(% delta k/k)
B 0.83 0.76 + 15%*
A 1.76 1.69 T 15%
D 2.44 2.21 T 15%
C 1.90 1.79 T 15%
- Acceptance region is + 15% of the predicted worth.
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Ejected Control Rod Measured Worth Predicted Worth (% delta k/k)
(% delta k/k)
Bank B at 242 steps Banks A, D, C at 320 steps B-ll 0.021 0.027 + 0.1 B-31 0.015 0.020 T_ 0.1 Bank A at 155 steps, Bank B at 0 steps Bank C, D at 320 steps B-ll 0.158 0.154 + 0.1 B-31 0.140 0.174 T 0.1 The boron worth determinations were conducted in accordance with SUR 5.3-6, and SUR 5.3-3, All Rods Out Just Critical Boron Concentration, Revision 4.
Procedure SUR 5.3-3 was reviewed for technical adequacy and boron worth results.
The following results were noted.
Measured Predicted Boron Worth (p m/%
135 147 + 10%
delta k/k
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While reviewing procedure SUR 5.3-6, the inspector noted that data sheets associated with the reactimeter strip charts did not have all of the required plant parameters at the time of measurement recorded.
This failure to follow and properly complete plant procedures is an additional example of the item of noncompliance specified in item 3 (79-11-02).
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(6)
Isothermal Temperature Coefficient Measurement The licensee measures the isothermal temperature coefficient in accordance with procedure SUR 5.3-5, Moderator Temperature Reactivity Coefficient Measurement, Revision 4 to comply with the requirement of Technical Specification 3.16.
This measurement was performed on March 3,1979.
Procedure SUR 5.3-5 was reviewed for technical adequacy and the calculations were verified to be within Technical Specification limits.
The following results were noted.
Isothermal Temperature Predicted Measured Coeficient delta k/k/F)
(x 10-'+)
( x 10-4)
All rod out~
-0.47 1 4-0.5 Banks A & B at 0 steps, C & D out-1.11 i.4-0.8 Average Value-0.65*
- compares with predicted coefficient of -0.74 x 10-4 delta
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k/k/F No items of noncompliance were identified.
(7)
Reactor Shutdown Margin The licensee assures maintenance of an adequate shutdown margin during power operation by complying with the control rod insertion limits of Technical Specification 3.18.
During plant startups and shutdowns, the shutdown margin is verified by performance of procedure SUR 5.3-19, Boration Requirements For Reactor Shutdowns, Revision 5, or procedure SUR 5.3-20, Reactivity Balance Procedure, Revision 1.
No items of noncompliance were identified.
(8) Core Thermal Power Evaluation The procedures and methods used by the licensee to determine core thermal power were reviewed and discussr.d with members of the licensee's Operations and Reactor Engineering staffs.
Procedure SUR 5.1-71, Reactor Core Power Surveillance, Revision 5, is performed daily to insure compliance with Technical Specification Table 4.1-1.
The plant on-line computer scans the appropriate data points and performs the 1616 206
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core thermal power determination. A check of the computer calculated core thermal power is performed at least once a month by a hand calorimetric calculation in accordance with procedure SUR 5.3-18, Hand Calorimetric and Analysis, Revision 3.
Hand calorimetrics were reviewed for the period May 25, 1979 to July 16, 1979.
No items of noncompliance were identified.
5.
Plant Operation On Three Loops During the inspection, the plant was being operated on three loops in order to perform maintenance on Number 3 Reactor Coolant Pump.
Independent inspection verification was made of plant operating parameters to verify compliance with Technical Specification limits, and a tour of the work area inside of containment was conducted on July 19, 1979, which included observation of crew preparation and entry into containment.
Except as noted below, the inspector: had no further questions in this area.
During the different modes of reactor operation, the nuclear overpower trip setting is selected from one of three positions on the " reactor trip selector switch'.'.
During three loop plant operation, Technical Specifications require the nuclear overpower trip setting to be 1 4% of rated power.
To establish this trip setting, the operator places the " reactor trip selector switch" in the "MID" position.
The inspector verified by observa-tion that the selector switch was in the "MID" position, but determined, through discussions with licensee representatives, that the trip setting was not verified by the licensee after establishing three loop operating conditions.
In fact, the last time the "MID" trip setting was verified was during the Cycle 9 refueling. A licensee representative acknowledged the inspector's coments and concerns, and stated that a procedural requirement for verification of the trip setting after establishing three loop conditions would be developed by October 20, 1979.
This item is unresolved pending the licensee's resolution and additional inspector review (79-11-05).
6.
General Training a.
Program Definition The inspector reviewed the Ticensee's training programs with respect to the program definition requirements of:
10 CFR 50, Appendix B, Criterion II; 10 CFR 19.12; 10 CFR 73.50; and ANSI N18.1. The training manual sets forth formal training programs for:
new employees; temporary maintenance or service personnel; operations personnel; technicians; and, craft personnel. These programs establish training which covers: administrative controls and procedures; fire training; and, quality assurance indoctrination.
Fomal training is also provided for female employees on the contents of Appendix A to Regulatory Guide 8.13.
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No items of noncompliance were identified.
b.
Program Participation The inspector reviewed the licensee's records to assure that the required training had been given.
In addition, the inspector conducted interviews with certain of those individuals whose records were reviewed. The interviews were to verify that:
the scope of the training was similar to that contained in the licensee's records; the training as conducted was meaningful to those attending; and, that the areas presented were covered accurately and sufficiently from the participants' point of view.
Records were reviewed for personnel listed below:
two employees with less than one year in their current
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position; two employees with more than one year of service;
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two temporary employees; and,
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three female employees
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Interviews were conducted with one employee from each group identified above.
No items of noncompliance were identified.
7.
Requalification Training a.
Program Items The inspector verified that, the program currently established Haddam Neck Plant Operator Requalification Program, submitted December 21, 1977, includes the following items:
an established, planned, continuing lecture schedule;
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documentation of personnel attendance;
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required reactivity control manipulations; (simulator training
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program)
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discussions / reviews of changes in facility design, procedures, and facility license; and, review of abnormal / emergency procedures.
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The inspector reviewed the lecture schedule for 1978-1979 and examined the lesson plans and instructional material for OP-REQ III LO-7, LO-8 and LO-0.
Simulator training was last accomplished in 1977; all operators are scheduled for simulator training in this cycle by December,1979.
b.
Record Review The inspector selected and reviewed the records of seven licensed operators to verify that each contained the following documentation for the requalification cycle 1978-1979:
completed course and yearly examinations with answers;
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manipulation of controls for reactivity changes required by
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the program; simulation or discussion of emergency / abnormal procedures
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and responses.
c.
Personal Interviews The inspector selected and interviewed three licensed personnel who participated in the requalification program.
The interviews were directed at obtaining subjective appraisal of the content and effectiveness of the requalification training as presented.
The inspector identified no discrepancies nor inconsistencies between the interview results and the licensee's records.
No item; of noncompliance were identified.
8.
Non-Licensed Personrel Training a.
Program Review The inspector reviewed the licensee's training program manual for plant staff training.
This review verified that no major changes in training programs have been made in the past year and that the programs contained on the job training, review of appropriate LER's, provision for attendance at selected vendor schools, and formal training commensurate with job classification. One item discussed in paragraph 2 remains unresolved; the item number is (413/78-21-03).
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b.
Interviews / Record Reviews The inspector selected the records of one instrument specialist, one QA specialist, one experienced mechanic, and one administrative staff member to verify that the required training had been conducted.
The inspector then interviewed certain of those personnel whose records had been reviewed to verify that: the scope of the training was as documented in the records; the training presented was meaningful to those attending; and, that the training was adequate from the participants' point of view.
9.
Unresolved Items Items about which more information is required to deterinine acceptablility'
are considered unresolved.
Paragraphs 4.b.3, 4.b.4 and 5 of this report contain unresolved items.
10.
Exit Interview At the end of this inspection, the inspector held a meeting (see Paragraph 1 for attendees) to discuss the inspection scope and findings.
The unresolved items and apparent items of noncompliance were identified.
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