ML20078S889
| ML20078S889 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 12/21/1994 |
| From: | Rainey G PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9412290274 | |
| Download: ML20078S889 (10) | |
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Ccrald R. Rainey Vice Presidrnt Peach Bottom Atomic Powsr Station 4
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O W RGY eeco eme<ov co-a ev RD 1, Box 208 Delta, PA 17314 9739 717 456 7014 December 21, 1994 Docket Nos. 50-277
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Ucense Nos. DPR-44 1
U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
Subject:
Peach Bottom Atomic Power Station Unit 2 Reply to Notices of Violation (NRC Inspection Report No.
50-277/94-26)
Gentlemen:
In response to your letter dated November 21,1994, which transm;+ted the Notices of Violation concerning the referenced inspection report, we submit the attached reply. The subject report concerned a special safety inspection conducted October 19 through October 29,1994, that investigated an event associated with the Unit 2 reactor pressure vessel leakage pressure test and included continuous NRC coverage of the Unit 2 start-up activities following refueling.
If you have any questions or desire additional information, do not hesitate to contact us.
Gerald R. Rai ey Vice President Peach Bottom Atomic Power Station Attachment CCN#94-14182 D s 941229o274 941221 DR ADoCK 050o 7
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cc:
R. A. Burricelli, Public Service Electric & Gas R. R. Janati, Commonwealth of Pennsylvania T. T. Martin, US NRC, Administrator, Region I W. L Schmidt, US NRC, Senior Resident inspector H. C. Schwemm, VP - Atlantic Electric R. I. McLean, State of Maryland A. F. Kirby lil, DelMarVa Power i
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e RESPONSE TO NOTICES OF VIOLATION i
Restatement of Violations 1.
10 CFR 50, Appendix B, Criteria XVI, Corrective Actions, reguires, in part, that measures be established to assure prompt identification of conditions adverse to quality.
Contrary to the above, during a one and one-half hour period between October 16 and 17,1994, PECO's licensed operators did not promptly identify that the reactor vessel head flange temperature had exceeded 212* F while in a natural circulation mode of core cooling. Operators monitored and recorded the temperatures every 15 minutes for the one and one-half hour period without identifying the potential for a reactor mode change from cold to hot shutdown.
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Technical Specification 6.8.1 requires that written procedures and policies be established and implemented that meet the requirements of Section 5.1 of ANSI N18.7-1972. ANSI N18.7-1972, Section 5.1.2 reguires: that procedures be followed; that requirements for use of procedures be prescribed in writing; and that rules be established for the use of temporary changes.
1 Administrative Procedure A-C-79, " Procedure Adherence and Use,"
Revision 1, Section 7.4, and Operations Manual Section 9.1 (Procedures) requires that operators follow test procedure restoration if conflicts occur l
the procedure shall be stopped, supervision consulted, and a temporary l
change performed, if necessary.
L Contrary to the above, on October 16,1994, the operations sts'f did not follow the test procedure restoration from the reactor pressure test (ST-J-080-675-2). Further, conflicts between the pressure test and the normal procedure for shutdown cooling (operating procedure S0-10.1.B-2) did not cause operators ??
l stop the test, involve supervision, nor did they perform a i
j temporary change.
l These are Severity Level IV violations (Supplement 1).
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'l Reason for the Violation 1.
Surveillance Test ST-J-080-675-2," Reactor Pressure Vessel (ASME Class I) Leakage Pressure Test" provided guidance to Operations personnel on the preparation, performance and recovery from the reactor pressure vessel (RPV) test during the tenth Unit 2 Refueling Outage (2R10).
Reactor Operators (RO's) recorded reactor vessel temperatures with surveillance test (ST)-O-080-500-2," Recording and Monitoring Reactor Vessel Temperature and Pressure" to ensure compliance with Technical i
Specification (TS) pressure-temperature (P-T) curve limits and the'100 degree F per hour heat-up and cooldown rate. They also recorded.
f temperatures using ST-O-080-520-2, " Reactor Vessel Head Flange ;
Temperature Surveillance" to ensure that reactor head flange temperature was maintained above 70 degrees F with the RPV head tensioned, and with coolant temperature less than 212 degrees F. ST-O-080-500-2 required that RPV metal temperatures be recorded every 15 minutes and l
both ST logs were required to be reviewed by a licensed senior reactor operator (SRO). The TS criteria provided in both ST's were met during the pressure test and bulk coolant temperature remained below 212 degrees F during the entire event. Following reactor depressurization with the reactor in the natural circulation mode, however, reactor head flange temperature began to increase.
The RO appropriately logged RPV metal temperatures every 15 minutes, but failed to identify the increasing temperature trend. The RO was more focused on evaluating the Reactor Water Cleanup (RWCU) pump discharge temperature (RWCU inlet temperature) than evaluating the reactor metal temperature data. RWCU inlet temperature is the primary indication of coolant temperature while in natural circulation. The RO clearly understood that bulk coolant temperature could not exceed 212 degrees F and therefore concentrated on this temperature indication.
Approximately one month prior to this evolution, the RO had monitored RPV metal temperatures during the Unit 2 shutdown w%n high metal temperatures (500 degrees F) were normal. In additiori, the RO had just recorded temperatures during the performance of the pressure test where metal temperatures in the range of 200-210 degrees F were required to meet the TS P-T curve. The RO indicated that as a result of previously recording high metal temperatures, his level of sensitivity to the increasing temperature trend was reduced.
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During the event, the RO used the RWCU pump discharge temperature to monitor reactor coolant temperature. Although the RWCU pump.
discharge provided an indication of reactor coolant temperature, metal i
temperatures should have been more closely monitored to give the most accurate indication of overall reactor temperature conditions. It was later determined that the RWCU pump discharge temperature indication was j
j out of calibration (low) by approximately 25 degrees F during this event.
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j Specific RPV metal temperature limits were provided for the performance i
of the pressure test, but there were no specific RPV metal temperature limits provided for the post pressure test condition. The surveillance l
test (s) and GP-12," Core Cooling Procedure", provided no procedural j
guidance concerning the use of RPV metal temperatures to determine i
when shutdown cooling (SDC) should be placed in service.
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i Operators were not fully aware of the potential for rapid temperature.
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stratification under natural circulation conditions. Although Operations I
i personnel received instruction on these issues during generic j
fundamentals training, followup and continuing training activities did not j
emphasize how quickly temperature stratification can occur.
Night orders and an Operations activities list provided a bullet type listing j
of jobs to be accomplished following the RPV pressure test, but activities j
were not prioritized. Additionally, the 2R10 refuel outage restoration j
activities schedule provided tasks to be performed after the pressure test j
was complete, but the restoration of SDC was not given special priority.
j As a result, these schedules did not prompt Operators to place SDC inservice before reactor metal temperatures began to increase.
1 The shift turnover was not thorough enough to allow the increasing trend l
In reactor metal temperatures to be identified. The oncoming' crew did j
not verify the status of critical parameters provided by the previous shift (i.e. temperature and pressure) or review the temperature logs. Neither j
the oncoming assigned RO or Shift Supervisor compared the RWCU pump discharge temperature with the reactor metal temperatures in order to obtain a better indication of reactor bulk coolant temperature.
j Since the oncoming Shift Supervisor was not aware of the increasing trend in reactor metal temperatures, SDC was not immediately placed in i
service. Subsequently, the Shift Supervisor became focused on the TS operability implications of an instrument backfill procedure which further i
diverted his attention from re-establishing SDC.
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Shift Supervision reviewed and signed the RPV temperature data sheets in ST-O-080-500-2, but failed to identify the increasing trend in metal temperatures. The test procedure required Shift Management to ensure that temperature readings, pressure readings and verifications had been j
completed and that no steps were checked off as unsatisfactory. Since no steps were noted as unsatisfactory and the appropriate data had j
been collected, the data sheets were signed off by Shift Supervision without a more thorough review of the data. Although the minimum requirements for the data sheet review were met, Shift Management should have performed a more rigorous review to ensure an acute awareness of important plant parameters.
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l Management expectations concerning Operator responsibility and oversight were not achieved during this event. It is the expectation that personnel involve upper management before initiating plant evolutions which could challenge operating limits. It is also the expectation of management that Operators heighten their awareness end oversight during these evolutions to ensure that operating limits are not exceeded.
In this post test condition, however, personnel did not aggressively pursue the control or oversight necessary to maintain an adequate margin with respect to the reactor coolant temperature limit.
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Reason for the Violation i
j 2.
Surveillance Test ST-J-080-675-2, Reactor Pressure Vessel (ASME Class j
I) Loakage Pressure Test" provided guidance to Operations personnel
'i during the preparation, performance, and recovery phases of the RPV l
pressure test during 2R10. The ST procedure was written to provide j
specific instruction on the use of the recirculation system prior to and j
during the RPV pressurization. It was the intent of the ST writer that j
j procedure GP-12, " Core Cooling Procedure" would be used to control 1
the use of the recirculation and SDC systems following reactor depressurization. Although this intent was not clearly stated in the ST, it j
was known to the Shift Supervisor who had previously reviewed the ST with its author.
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During previous outages, recirculation pumps were utilized to maintain or increase reactor coolant temperature. Due to the short duration of 2R10, an adequate amount of decay heat existed to maintain reactor
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temperature. The extended use of a recirculation pump could add i
unwanted heat to the reactor and cause reactor water temperature to l
increase. Therefore, following the completion of the hydrostatic test, the j
Shift Supervisor directed that the 'B' recirculation pcmp be removed from j
service to stop the heat input created by the recirculation pump by the i
use of System Operating procedure (SO)-2A.2.A-2, " Recirculation Pump l
Shutdown". The Shift Supervisor decided that this action was acceptable j
because he knew the intent of the ST and because the action was not i
inconsistent with the guidance provded in GP-12. The ST provided no j
guidance or restrictions that prevented the removal of a recirculation
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pump from service following reactor depressurization. In addition, the Shift Supervisor was aware that the recirculation pump would eventually 4
need to be secured to place SDC in service per SO-10.1.B-2, "RHR l
System SDC Mode Manual Start". The fact that the ST wou!J have transitioned the shift through the restoration part of the test while j
maintaining forced circulation was by circumstance, and not design.
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An attachment was added to the pressure test to conduct restoration of j
reactor level inconjunction with flushing the residual heat removal (RHR) j system piping, prior to and in preparation for SDC operation. This was i
done to provide Operations with a single procedure for the test evolution l
to avoid confusion between the pressure test and the normal SDC procedure. Test Attachment 3 provided a new method to reduce RPV j
inventory by draining through RHR piping into the torus, in accordance l
with step 6.10.9 of the pressure test ST, reactor water level could be reduced by one or more of five methods which could have been j
performed simultaneously. The shift selected option 1 (Reduction of CRD i'
flow and maximization of RWCU flow), option 2 (Drain through Main j
Steam line drains to the condenser) and option 5 (addressed in i ) where reactor water level was reduced by draining the i
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SDC suction line through an RHR pump to the torus. Although I
precautions were provided in Attachment 3 not to lower reactor water level below 17 inches or increase torus level above 14.9 feet, there Was no guidance or information provided on the total volume of the flush. The shift referenced SO-10.1.B-2 to obtain the total volume of the upcoming RHR piping flush. After this information was obtained, the shift continued the reactor level reduction in accordance~with Attachment 3 of the ST.
When level adjustments had been completed, the shift prepared to place SDC in service as directed by Attachment 3 and SO-10.1.B-2.
Due to the delay for placing SDC in service and a lack of awareness of the potential for rapid RPV temperature stratification under natural circulation conditions, the reactor Was not placed in SDC until the Shift Supervisor was notified of high RPV metal temperatures by the RO. GP-12 was found inadequate in that appropriate guidance and precautions for natural circulation conditions were not provided for Operations personnel. Specifically, the procedure did not provide adequate instruction to prevent the potential for localized boiling and steam induced heating of the reactor head. In addition, there was no discussion or information provided concerning possible indications of thermal stratification while in natural circulation. There were also no precautions for avoiding natural circulation when reactor water level was reduced in the refuel mode.
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As a result of knowing the intent of ST-J-080-675-2 with respect to the usage of recirculation pumps following RPV depressurization, the Shift Supervisor did not perceive his actions to be in conflict with administrative procedure A-C-79, " Procedure Adherence and Use".
Ukewise since the Operators did not perceive a conflict between the ST and SO for placing SDC inservice, they did not stop the test, involve supervision, or perform a temporary change to the procedure. This event I
indicates that a clearer station philosophy must be established regarding how to transition between a major test procedure and normal plant operating procedures.
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Corrective Actions That Have Been Taken A Performance Enhancement Program (PEP) investigation (PEP-10003027) was initiated to determine the causal factors of the event and to develop appropriate corrective actions to prevent recurrence.
Appropriate disciplinary actions were taken commensuate with the individual's level of responsibility.
The Operations Manager reinforced expectations to the Operating Shifts concerning heightened awareness and questioning attitudes during testing or planned evolutions, the need to involve upper level management when operating limits could be unnecessarily challenged, and the need to ensure an effective and comprehensive information exchange during shift turnover.
. RWCU pump discharge temperature indicator (TI)-2-12-137 was successfully re-calibrated October 18,1994, f-l i
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j Corrective Steos That Will be Taken to Avoid Further Violations GP-12," Core Cooling Procedure", ST-J-080-675-2, " Reactor Pressure Vessel (ASME Class 1) Leakage Pressure Test" and ST-O-080-500 2, " Recording and Monitoring Reactor Vessel Temperature and Pressure" will be revised to provide conservative limits and parameters to better control the evolutions. Additionally, the procedures will be revised to ensure the instructions and guidance provided are appropriate and adequately address anticipated plant conditions. These revisions will be accomplished by May 31,1995.
Training will be initiated in the area of natural circulation and resultant thermal startification effects during the next cycle of Ucensed Operator Training for the operating shifts. This training cycle will be initiated in January,1995. In addition, the simulator will be used to reinforce the questioning attitude of shift personnel during simulated operating canditions.
The PBAPS ST Writers Guide will be revised to clearly state and discuss the philosophy and expectation regarding surveillance test evolutions and their relationship to plant operating and general procedures. This revision will be accomplished by March 31,1995. Once completed, the philosophy and expectation of this revision will be properly communicated and disseminated to plant staff.
Date When Full Comoliance Was Achieved Full compliance was achieved October 17,1994, when the Reactor Operator informed the Shift Supervisor of an increasing trend on the RPV flange temperature. The Shift Supervisor then directed that the 'B' Residual Heat Removal (RHR) pump be placed inservice in SDC, which initiated forced circulation and the reduction of bulk reactor temperature.