IR 05000029/1980007
| ML19318B245 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 05/05/1980 |
| From: | Foley T, Martin T, Raymond W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML19318B243 | List: |
| References | |
| 50-029-80-07, 50-29-80-7, NUDOCS 8006250164 | |
| Download: ML19318B245 (8) | |
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U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT Region I Report No.
50-29/80-07 Docket No. 50-29 License No. DPR-3 Priority Category
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Licensee:
Yankee Atomic Electric Company 25 Research Drive Westborough, Massachusetts 01581 Facility Name:
Yankee Nuclear Power Station (Yankee Rowe)
Inspection at:
Rowe, Massachusetts Inspection conducted: March 17-24, 1980 Inspectors:
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$~ c5 h D W. J.' Raymond, actorfnspector date signed i
date signed Approved by:
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T. T. Martin, Chief, Reactor Projects date signed Section No. 3, RO&NS Branch Inspection Summary:
Inspection on Mar-h 17-24, 1980 (Report No. 50-29/80-07)
Areas Inspected:
Routine, unannounced inspection of licensee emergency procedures for small-break loss of coolant accidents; review of Licensee Event Reports (LERs)
and Monthly Operating Reports. The inspection involved 46 inspector-hours on site by two region-based inspectors.
Results:
No items of noncompliance were identified.
Region I Form 12 (Rev. April 77)
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DETAILS 1.
Persons Contacted
- H. Autio, Plant Superintendent G. Babineau, Engineering Assistant T. Danek, Operations Supervisor
- L. French, Engineering Assistant T. Henderson, Technical Assistant
- F. Hicks, Training Coordinator
- K. Jurentkuff, Day Shift Supervisor L. Laffond, Assistant Training Coordinator P. Laird, Maintenance Supervisor
- N. St. Laurent, Assistant Plant Superintendent
- J. Staub, Technical Assistant to Plant Superintendent J. Trejo, Plant Health Physicist D. Vassar, Assistant Operations Supervisor The inspector also interviewed other licensee employees during the course of the inspectica.
They included Health Physics Technicians and Plant Operators.
- Denotes those present at the exit interview.
2.
Plant Tour The inspector toured accessible areas of the plant, including the Turbine Building, Control Room, Primary Auxiliary Building, Safety Injection Building, Diesel Generator Rooms, Radwaste Building, Switchgear Room, and HP Control Point. The inspector reviewed the status of health physics controls, housekeeping and cleanliness, presence of fire and safety hazards, conditions of piping supports, and existence of fluid leaks.
No unacceptable conditions were identified.
3.
Review of Small-Break LOCA Procedures a.
Procedure Implementation The inspector reviewed the emergency procedures implemented by the licensee, which would be used in the event of a small-break loss of coolant accident (LOCA), to determine if the guidelines developed by the Westinghouse Owner's Group had been incorporated.
These guidelines were submitted to the NRC as part of the generic report WCAP-9600, " Report on Small Break Accidents for Westinghouse NSSS System," and were subsequently approved by NRR for implementation in a letter dated November 5, 1979.
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The f'llowing documents were reviewed:
OP-3051, Loss of Main Coolant Pressure and/or Safety
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Injection Initiation, Revision 1; OP-3106, Loss of Main Coolant, Revision 18;
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OP-3201, Loss of Secondary Coolant, P.a/isions 11;
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OP-3107, Steam Generator Tube Rupture, Revision 10;
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OP-3053, Inadequate Core Cooling, Revision 1;
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OP-3000, Emergency Shutdown From Power, Revision 13;
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YR Memorandum from J. Chapman to N. St. Laurent, dated
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December 3,1979; Yankee Atomic Electric Company letter WYR 79-104 to USNRC
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Office of NRR dated September 12, 1979; Yankee Atomic Electric Company letter WYR 80-11 te USNRC
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Office of NRR dated January 24, 1980; YR Memorandum from W. Reed to W. Jones dated January 22,
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1980.
Except as noted below, no discrepancies were identified. Westinghouse guidelines are referenced in parentheses.
OP-3051 Immediate Actions step 1 directs the operator to
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initiate OP-3000, " Emergency Shutdown From Power." 0P-3000 Immediate Action Step 2 directs the operator to OP-3051.
The procedure OP-3051 Immediate Action step 1 should be more specific and state the applicable steps to be performed in OP-3000.
In each procedure where an operator might restart the Main
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Coolant Pumps, a Caution should be added which ensures that
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the cold leg Main Coolant Cutout Valve (MC^V) for a particular loop is open prior to shutting the cold leg MCCV in an adjacent loop as a prerequisite for starting another Main Coolant Pump.
A Caution should be added to OP-3106 and OP3051 procedures
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which prohibits the use of the loop isolation valves to isolate the break.
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A Caution should be added to each emergency procedure
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that performs a switchover from the Safety Injection mode to the recirculation mode of cooling that requires the operator to perform the actions in an expeditious, precise and orderly manner, and the operation shall not be interrupted until all actions are completed.
(Table E-1.1, Pre-requisite steps C and E)
OP-3106 Step 14 note 1 addresses the fact that Recirculation
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Water will be warmer than SI tank water and system tempera-ture should increase when recirculation is established.
This note ne2ds clarification that system temperatures should not continue to rise, and a verification of adequate core cooling after going into the recirculation mode is required.
(Table E-1.1 step 5)
The above items are collectively unresolved pending incorporation into the appropriate procedures and subsequent review oy the NRC. (50-29/80-07-01)
b.
Procedure Modifications Modifications to Small Break Loss of Coolent Accident Guidelines are described in a. December 27, 1979 letter from the NRC (D.
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F. Ross) to the Westinghouse Owner's Group (C. Reed). The modifications concern revisions to the following criteria given in parts E-0 and E-1 of the guidelines:
(1) A CAUTION statement be added at appropriate locations in the emergency 7rocedures to limit the number of start /stop cycles on the.1igh head safety injection pumps to avoid pump motor overheating or reduced motor life. Thus, if the pumps are to be restarted after termination of SI, an additional 150F of sub-cooling should be added to the required sub-cooling margin prior to the second termination of the SI.
(2)
In those places where the emergency procedures address steam generator water inventory as part of SI termination criteria, the requirement should be that water level in at least one steam generator is statle and increasing as verified by auxiliary feedwater (AFW) flow to that
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unit; and that AFW flow to the generator be sufficient to remove core decay heat until indicated water level is within the range of the narrow range instrument, con-sistent with other constraints which may be in force on steam generator operation, such as water hammer considerations; and,
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(3) The specif'c.value of sub-cooling margin listed as part of SI termination criteria incorporate uncertainties in the temperature / pressure measurement system based upon
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the method of the " square root of the sum of the squares of errors".
The inspector noted that the emergency procedures require a minimum margin of 400F sub-cooling exist as part of the SI termination criteria. -The 400F sub-cooling was not derived
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from an. error analysis of the sub-cooling instrumentation,
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but rather is based upon achieving a stable primary condition, without reactor coolant pumps operating, at 1800 psig with the secondary system pressure at~ a value corresponding to the lowest steam generator safety valve pressure of 935 psig, and is thus conservative.
Documentation and analysis of errors associated with the sub-cooling instrumentation is provided in a January 24, 1980 letter to the NRC. This analysis, based on the method of using the " square root of the sum of the squares" demonstrates a 1130F error associated with the sub-cooling instrumentation and thus, the 400F criteria assures that a sub-cooling condition greater than 200F will be achieved. The inspector reviewed the analysis contained in the January 24, 1980 letter and had no further coments on this item.
In regard to the criteria that adequate yteam generator water inventory be verified orior to terminating SI, the inspector
noted that the procedures already require that control operators attempt to maintain stea i generator water level at 18 feet prior to SI termination. This requirement is not stated uniformly throughout the procedures. The licensee stated that the 18 foot steam generator water level criteria
will be added in the emergency procedures as appropriate
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(for example, in procedure step 10, page 3 of OP-3051; step
2 (D) of Attachment 1 of OP-3106) and inst'uctions will be
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added to verify that auxiliary feedwater flow to the steam generators is greater than 80 opm. Additionally a curve of steam t
generatcr level required to assure sufficient decay heat removal will be added as-an operator guide. The inspector noted,
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based on present operating practices at ifR and the NRC's February _5,1980 Safety Evaluation Report - Steam Generator Water Hammer Issue at YR, the potential for steam generator water hammer using either BFP or EBFP, is not a concern at
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YR due to steam generator modifications made in 1965.
In regard to criteria to : limit SI pump operation, the inspector noted the current emergency procedures do not contain the
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appropriate ' precautions. The licensee stated that CAUTION
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statements consistent with the modified guidelines (as listed above) will be added to the procedures (for example, step 9; page -5 of OP-3051; page 1, of Attachment A, step F of OP-3106).
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This item is unresolved pending incorporation of the above changes in the emergency procedures and subsequent review by the NRC.
(50-29/80-07-02)
c.
Training Requirements The inspector reviewed licensee training records and interviewed several licensed operators to verify that they had received formal training in the revised small-break LOCA procedures.
It is an NRC position that a " hands on" walk through of the
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Small Break LOCA procedure be performed on the control board. The formal training of the Small Braak LOCA procedures
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received by the operators did not include a walk through of
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the procedure at the control board, however the inspector verified that each operator had received a walk through on the control board of the Short Term Lessons Learned Category
"A" Items. The inspector's review of the Small Break LOCA procedures with the operators selected for interview, which included a walk through on the control board revealed no significant deficiencies which an additional walk through would correct. The licensee's representative stated that the operators will receive additional refresher training in this area durina their pre-startup training. The inspector had no further questions in this area.
d.
Operator Interviews The inspector interviewed three senior reactor operators and two reactor operators to determine their familiarity with procedure OP-3051, " Loss of Main Coclant Pressure and/or Safety Injection Signal".
Except as noted below, no discrepancies were identified.
The immediate action steps of the emergency procedures should be performed from memory. The precautions associated with these immediate action steps should also be followed.
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The inspector noted that four of the five operators interviewed failed to check the redundant channels for consistency as required in the precautions of OP-3051. The.came operators did not indicate that the pressurizer water level would be used in conjun: tion with other parameters to evaluate the system response and initiate operator action as required in
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OP-3051 precautions.
The operators in general lacked familiarity with the newly installed saturation monitor and did not in all cases verify
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natural circulation after securing the main Coolant Pumps.
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The licensee's representative stated that the above items would be reviewed during subsequent pre-startup training.
This item is unresolved pending NRC review of pre-startup training tentatively scheduled for the fall of 1980. (50-29/80-07-03)
e.
System Considerations The inspector reviewed with the operators the instrumentation needed to carry out actions in the emergency procedures to insure that th7 procedures are viable in this regard.
This review identified the following concerns:
(1) The saturation monitor installed as a Lessons Learned Category "A" Item has an integral test mode, which if activated requires several specific sequential steps in order to return the monitor to an operational mode.
The licensee's representative stated that a caution sign would be placed near the monitor warning the operators of this characteristic of the saturation monitor.
(2) OP-3106 instructs the operator to declare various categories of emergencies depending upon the radiation level associated with the Accident Area Radiation Monitor (AARM). The levels described in the procedure are simple numeric where as the (AARM) displays exponential notation (i.e.:
500.R vice 005.0X105 mr).
All the operators interviewed experienced difficulty in deciphering the exponential notation display on the AARM. The licensee's representative stated that the procedure would be revised to make the notation consistent with each other or change the display to facilitate reading the AARM.
The above items are collectively unresolved pending subsequent review by the NRC.
(50-29/80-07-04)
4.
In Office Review of Licensee Event Reports (LERs)
The inspector reviewed the following LERs received in the RI office to verify that details of the event were clearly reported including the accuracy of the description of cause and adequacy of corrective action. The inspector also determined whether further information was required from the licensee, whether generic implications were indicated, and whether the event warranted on site followup.
79-30, SI Accumulator Time Delay Relay Failure
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79-31, Steam Generator Blowdown Monitor Out of Service
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79-32, Primary Vent Stack Monitor Pump Failure
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79-33, Loss of Flow Path from BAMT to the Charging Pumps;
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E0-01, Steam Generator Blowdown Monitor Out of Service;
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80-02, Primary Vent Stack Iodine Channel Failure;
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80-03, Sealed Source Missed Surveillance;
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80-04, Secondary Water CL-Concentration Greater than.5
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ppm.
The inspector had no further questions concerning these LERs.
5.
In-Office Review of Monthly Reports The inpector reviewed the Monthly Operating Reports for October, November and December 1979 to verify reporting requirements were met.
No inadequacies were identified.
6.
Unresolved Items Unresolved items are those items for which further information is required to determine whether they are acceptable items or items of noncompliance.
Unresolved items are con"hined in Paragraphs 3 (a), (b), (d) and 3 (e) of this report.
7.
Exit Interview The inspectors met with licensee representatives (denoted in Paragraph 1) at the conclusion of the inspection on March 24, 1980. The inspector summarized the scope and the findings of the inspection as they are detailed in this report.
During this meeting, the unresolved items were identified.
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