ML18066A875

LER 97-001-00:on 970106,TAVE Temp Dropped Below Minimum Temp for Criticality.Caused by Control Rod Withdrawal Rate to Increase Power Not Sufficient to Match Increase in Steam. Turbine Bypass Valve Actuator repaired.W/970205 Ltr

ML18066A875
Person / Time
Site:	Palisades
Issue date:	02/05/1997
From:	Flenner P, Thomas J. Palmisano CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-97-001-01, LER-97-1-1, NUDOCS 9702120343
Download: ML18066A875 (6)

Text

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~'.!~ }ij. Power Thomas J. Palmisano Plant General Manager POWERiNii MICHIGAN'S PRDliRESS Palisades Nuclear Plant: 27780 Blue Star Memorial Highway, Covert, Ml 49043 February 5, 1997 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 DOCKET 50-255 - LICENSE DPR PALISADES PLANT LICENSEE EVENT REPORT 97-001 -VIOLATION OF T.S. 3.1.3, TAVE LESS THAN 525°F WHEN REACTOR CRITICAL Licensee Event Report (LER)97-001 is attached. This event is reportable in accordance with 10 CFR Part 50.73)a)(2)(i)(B) as a condition prohibited by Technical Specifications. '

SUMMARY

OF COMMITMENTS

• This letter contains no new commitments or revisions to existing commitments.

Thomas J. Palmisano*

Plant General Manager CC Administrator, Region Ill, USNRC Project Manager, NRR, USNRC NRC Resident Inspector - Palisades 97021203.43 970205 .

PDR ADOCK 05000255 S PDR A CMS' ENERGY COMPANY

NRC FORM 366 U.S.NU AR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104.

(4/95) EXPIRES 4/30/98 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTT1i THIS MANDATORY INFORMATION COUECTION REQUEST: 50.0 HRS. REPORTED LESSONS LEARNED ARE INCORPORATED INTO LICENSEE EVENT REPORT (LER) THE UCCNSING PROCESS AND FED BACK TO INDUSTRY. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TOTliE INFORMATION AND RECORDS MANAGEMENT BRANCH (T-6 F33), U.S.

NJCLEAR REGULATORY COMMISSION, WASHINGTON, DC 2055!>-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3151).01()(, OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC (See reverse for required number of digits/characters for each block) 20503 FACILITY NAME (1) DOCKET NUMBER (2) Page (3)

PALISADES NUCLEAR PLANT 05000255 1of5 TITLE (4) LICENSEE EVENT REPORT 97-001, VIOLATION OF TECHNICAL SPECIFICATIONS 3.1.3, TAve LESS THAN 525°F WHEN REACTOR CRITICAL EVENT DATE (5) LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (8)

MONTH DAY YEAR YEAR I SEQUENTIAL REVISION NUMBER NUMBER MONTH DAY YEAR FACILITY NAME DOCKET NUMBER 05000 01 06 97 97 - 001 - 00 02 05 97 FACILITY NAME DOCKET NUMBER 05000 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check one or more) (11)

MODE (9) N 20.2201 (b) 20.2203(a)(2)(v) 50.73(a)(2)(i) 50.73(a)(2)(iii)

POWER 20.2203(a)(1) 20.2203(a)(3)(i) x 50.73(a)(2)(ii) 50.73(a)(2)(x) 20.2203(a)(2)(i) 20.2203(a)(3)(ii) 50.73(a)(2)(iii) 73.71 20.2203(a)(2)(ii) 20.2203(a)(4) 50.73(a)(2)(iv) OTHER 20.2203(a)(2)(iii) 50.36(c)(1) 50.73(a)(2)(v) Specify in Abstract below or in 20.2203(a)(2)(iv) 50.36(c)(2) 50.73(a)(2)(Vii) NRC Form 366A LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER (Include Area Code)

Philip D. Flenner, Licensing Engineer (616) 764-2000 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE TO TONPRDS NPRDS SUPPLEMENTAL REPORT EXPECTED (14) MONTH DAY YEAR I YES If yes COMPLETE EXPECTED COMPLETION DATE x I NO EXPECTED SUBMISSION DATE (15)

ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16)

On January 6, 1997, at 1410, while synchronizing the Main Generator to the grid, TAVE dropped to approximately 524°F. This temperature was below the minimum temperature for criticality, 525°F, specified in Technical Specifications. TAVE was below 525°F for less than one minute. The event occurred because the control rod withdrawal rate to increase power was not sufficient to match the_ ..

increase in steam demand.

On January 6, 1997 at 1917, during a plant shutdown, TAvE dropped below 525°F for six minutes.

Reactor Engineering has determined that the reactor was subcritical by 0.3%~p; however, reactor a

power was greater than 10-4% pow~r. which is violation of Technical Specifications.

In both of these events the task planning was not comprehensive enough to prepare the operators to recognize and implement the necessary contingency and compensatory actions required for the plant and equipment responses that were actually experienced.

NRC FORM 366a U.S. NUCLEAR REGULATORY COMMISSION 4/95 LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1 l

• DOCKETl2l LER NUMBER 6) PAGE 13)

CONSUMERS POWER COMPANY 05000255 YEAR I SEQUENTIAL NUMBER REVISION NUMBER 2 OF 5 PALISADES NUCLEAR PLANT 97 - 001 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

EVENT DESCRIPTION On January 6, 1997, at 1410, while synchronizing the Main Generator to the grid, TAvE dropped to approximately 524°F. TAvE was below 525°F for less than one minute.

On January 6, 1997 at 1917, during a plant shutdown, TAvE dropped below 525°F for six minutes with reactor power greater than 10-4°/o power.

Technical Specification 3.1.3.a states, "Except during low-power physics test, the reactor shall not be made critical if the primary coolant temperature is below 525°F." Technical Specifications define REACTOR CRITICAL as, "The reactor is considered critical for purposes of administrative control when the neutron flux wide range channel instrumentation indicates greater than 10-4% of RATED POWER."

These events were discovered by the on shift operators at the time of the events.

CAUSE OF THE EVENT Task planning was not comprehensive enough to prepare the operators to recognize and implement the necessary contingency and compensatory actions required for the plant and equipment responses that were actually experienced.

Additionally, Turbine Bypass Valve CV-0511 may have contributed to the severity of the primary system cooldown. An unattached packing follower on the valve actuator may have resulted in erratic valve movement during its operation that would not have been readily noticed by the operators.

ANALYSIS OF THE EVENT Event #1 On January 6, 1997, operators practiced synchronizing the Main Generator to the grid using the simulator. A reactor power increase of 2-3% was assumed to account for the additional steam flow with the turbine on line and the Turbine Bypass Valve fully closed. A reactivity change using control rods was estimated using the Technical Data book. It was determined that rod withdrawal of 3-4 inches should balance the increased steam flow. This assumption was validated when the task was successfully performed on the simulator.

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NRC FORM 366a U.S. NUCLEAR REGULATORY COMMISSION 4/95 LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1 l

• DOCKET(2) LER NUMBER 6) PAGE 13l CONSUMERS POWER COMPANY 05000255 YEAR I SEQUENTIAL NUMBER REVISION NUMBER 3 OF 5 PALISADES NUCLEAR PLANT 97 - 001 00 TEXT (If more space is required, use additional copies of NRG Form 366A) (17)

The same operators proceeded with the Main Generator synchronization. They expected to only have to withdraw rods -4 inches or at most 8 inches to stabilize TAvE* When TAvE did not stabilize above 525°F, significant additional control rod withdrawal (a total of 18 inches) was required that had not been anticipated by the operators. Therefore, the careful and deliberate manner that the operators took in withdrawing the control rods on this occasion failed to increase reactor power sufficiently to prevent TAvE decreasing below 525°F. The 18 inches of withdrawal is consistent with

. what was seen during the startups on December 27, 1996, and on January 14, 1997.

Other differences between the expected plant response and the actual plant response contributed to this event. The Turbine Bypass Valve was discovered to have an unattached packing follower on the valve actuator. Therefore, the actual operating characteristics of the Turbine Bypass Valve during this evolution are unknown.

Additionally, feedwater oscillations occurring at the time may have aggravated the plant cooldown.

Both conditions, therefore, probably caused an actual plant response of greater magnitude than the operators were led to anticipate.

Based on interviews and discussions with the operators involved, there were no knowledge deficiencies or inadequate skills that contributed to this event. ..

The lessons learned from this event included:

• The over-reliance placed on the simulator and the Technical Data book to exactly model actual plant response led to a task plan that was not sufficiently comprehensive to prevent the event.

Operators did not fully anticipate the differences between the expected plant response and actual plant responses to develop the necessary contingencies and compensatory actions that would have prepared the operators to handle any departure from what was practiced.- - * - ---- -- --

• *Equipment and system conditions and perfo~mance should be investigated and factored into task plans. In this event, a feedwater regulation system,oscillation should have been evaluated for its effect on temperature control. Additionally, the low amount of decay heat and a Moderator Temperature Coefficient close to zero had an effect on the task performance.

• Task preparation and planning should consider previous operating experience. Placing the generator on line on December 27, 1996, required a control rod withdrawal of similar magnitude as that which was ultimately needed on January 6, 1997.

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NRC FORM 366a U.S. NUCLEAR REGULATORY COMMISSION 4/95 LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1) DOCKET(2) LER NUMBER 6) PAGE (3)

CONSUMERS POWER COMPANY 05000255 YEAR I SEQUENTIAL NUMBER REVISION NUMBER 40F 5 PALISADES NUCLEAR PLANT 97 - 001 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

Event#2 During "C" shift on January 6, 1997, the decision was made to remove the plant from service to correct the steam leaks on the MSIVs. Task preparation and planning time were minimized due to the urgency that was felt to shut down quickly. This shutdown was different from most shutdowns that the operators experience or practice in that there was very minimal decay heat. In carrying out the actions that are normal with a noticeable level of decay heat, the main feed pump was not tripped in time to prevent TAVE from falling below 525°F. Without deliberate forethought to expeditiously remove the main feed pump from service due to the low decay heat level, TAvE was inadvertently reduced below 525°F.

Technical Specification 3.1.3.a was violated by going below 525°F while above 10-4% rated power (Technical Specification definition of criticality). Reactor Engineering has determined empirically that the reactor was subcritical by at least 0.3%t:.p prior to TAVE falling below 525°F.

Based on interviews and discussions with the operators involved, there were no knowledge deficiencies or inadequate skills that contributed to this event.

The lessons learned from this event included:

• Task preparation and planning time were minimized due to the necessity to shut the plant down due to a main steam isolation valve steam leak. However, there may have been a false sense of priority established to shut down quickly that failed to consider the necessary time to allow operators to thoroughly plan and prepare for the task.

• During an event or plant situation,*Operations shift management personnel should be cautious not to become so focused on details that they fail to provide an overview function. The overview function should prepare their personnel, to the extent possible, on procedure implementation, Technical Specification limits, contingencies, compensatory actions, etc.

• Operators did not adequately account for the minimum core decay heat addition and plan for the timely stopping of main feedwater flow.

• Operations shift management personnel did not adequately recognize and stress to their operators the need to maintain primaiy coolant system temperature above 525°F w~en the reactor was critical. This should have been a lesson learned from the event earlier in the day.

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NRC FORM 366a U.S. NUCLEAR REGULATORY COMMISSION 4/95 LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION I

FACILITY NAME (1) . DOCKET(2) LER NUMBER 6) PAGE (3)

YEAR SEQUENTIAL REVISION CONSUMERS POWER COMPANY NUMBER NUMBER 05000255 50F 5 PALISADES NUCLEAR PLANT 97 - 001 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

SAFETY SIGNIFICANCE Event #1 There is no safety significance to a TAVE of 524 ° F for less than one minute with the reactor critical.

Standard Technical Specifications allow a 30 minute time period to correct a low primary coolant temperature. At one time Palisades proposed a Technical Specification change to specify an action time for this condition. This request was later withdrawn. The NRC concurred with this withdrawal in a letter to Palisades dated May 28, 1986, which stated, "The request dated February 5, 1985 regarding a time allowance to take corrective measures prior to initiating the action statement for low primary coolant temperature is not needed because of the allowance already granted by

."Specification 3.0.3. Therefore, we concur in its withdrawal."

Event#2 There is no safety significance to this event. Reactor Engineering has determined empirically that the reactor was subcritical by at least 0.3%6p prior to TAVE falling below 525°F. Therefore we met the intent of Technical Specification 3.1.3.

CORRECTIVE ACTION

1. Senior Operations Management has reviewed these events with all crews to emphasize and discuss the lessons learned stated above. These discussions took place with all crews prior to plant startup.

2. * .The Turbine Bypass ValveActuatorwas repaired.

3. Feedwater controls were tuned to minimize the feedwater flow oscillations.

PREVIOUS EVENTS Previous occurrences of the reactor being critical with the primary coolant system less than 525°F were reported as follows:

LER 84-014 LER 87-018 LER 87-022 LER 87-028