ML20153C362

Responds to Notice of Violation & Proposed Imposition of Civil Penalty Issued on 880725 Re Inoperability of Two LPCI Subsystems of Rhr,Primary Containment Integrity Maint & Primary Containment Air Lock

ML20153C362
Person / Time
Site:	Brunswick
Issue date:	08/24/1988
From:	Mcduff M, Mcduffie M CAROLINA POWER & LIGHT CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM), NRC OFFICE OF ENFORCEMENT (OE)
References
EA-88-131, NLS-88-206, NUDOCS 8809010104
Download: ML20153C362 (10)

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i carouna Power a ueht company P. O Box 1451 e RaMgh, N. C. 27002 AUB24 MIS M. A. McDUFFIE sen.or V,ce Pmident Nxkar O*"'** SERIAL: NLS 88-206 Director, Office of Enforcement United States Nuclear Regulatory Commission ATTENTION: Document Control Desk Washington, DC 20555 BRUNSWICK STEAM ELECTRIC PIANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50 325 & 50-324/ LICENSE NOS. DPR-71 & DPR 62 ANSWER TO A NOTICE OF VIOLATION EA 88 131 Centlemen:

The NRC issued a notice of violation and proposed imposition of civil penalty for the Brunswick Steam Electric Plant (BSEP), Units 1 and 2 on July 25, 1988. Pursuant to 10CFR2.201, Carolina Power & Light Company (CP&L) hereby submits a reply to the notice of violation. Each of the violations is addressed in Enclosures 1, 2, and 3, respectively. Each response includes (1) an admission or denial of the violation, (2) the reason for the violation, (3) corrective actions which have been taken, (4) future corrective actions, and (5) the date by which compitance will be achieved.

Carolina Power & Light Company agrees that the violations, when viewed together, identify an issue of critical importance to the safe operation of the Brunswick Plant and meet the criteria for imposition of a civil penalty. The Company does not agree, however, that escalation of the civil penalty for an event lacking serious safety significance is justified, simply because the event has been collectively incorporated with two other events.

The three events were collectively categorized as Severity Level III in accordance with Supplement I of 10CFR2, Appendix C. It is the prerogative of the NRC to combine events, activities, and/or violations together if circumstances or conditions so warrant. Cop-ination of such events allows the overall safety significance of similar issues to be put into proper perspective. However, once these evaatt have been combined to represent a more significant concern, they lose their unique identity. Thus, cor.2iderations for escalation of the penalty must be evaluated aga. inst the violation as a 9 hole (i.e., the combine. tion of the three violations) since that is what provides the justification for the base civil penalty.

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NLS 88-206 / Page 2 As stated in the Notice of Violation, the three violations cited individually do not have serious safety significance and, therefore, if cited individually, would not warrant a civil penalty. Considerations for escalation were based solely on Violation B; not on the violation as a whole. Thus, the Company believes that the escalation of the civil penalty under 10CFR2, Appendix C, item V.B.3 (Past Performance) is inappropriate. Consistent with the responses included in Enclosures 1, 2, and 3, Carolina Power & Light Company hereby submits a check in the amount of $50,000.00 for payment of the proposed civil penalty and takes exception to the $25,000.00 escalation.

Please refer any questions regarding this submittal to Mr. Stephen D.

Floyd at (919) 836-6901.

Yours very truly, k '

M. A. McDuffie BAB/bab(\cor\vresp)

Enclosures cc: Dr J. Nelson Crace Mr. W. M. Ruland Mr. B. C. Buckley M. A. McDuffie, having been first duly sworn, did depose and say that the information contained herein is true and correct to the best of his information, knowledge and belief s and the sources of his information are officers, employees, contractors, and agents of Carolina Power &

Light Company.

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ENCLOSURE 1-RESPONSE TO VIOLATION A

- Descrintion of Violation The NRC's Notice of Violation states the following:

"Technical Specification (TS) 3.0.4 states that entry into an OPERATIONAL CONDITION or othec specified applicability state shall not be made unless the conditions of the Limiting Condition for Operation are met without

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reliance on provisions contained in the ACTION statements unless otherwise excepted.

"TS 3.5.3.2 requires in OPERATIONAL CONDITIONS 1, 2, and 3 that two '

independent low pressure coolant injection (LPCI) subsystems of the residual heat removal (RHR) system be OPERABLE with each subsystem comprised of two pumps and an OPERABLE flow path capable of taking suction from the suppression pool and transferring the water to the reactor pressure vessel, i

( "TS 3.6.1.1 requires in OPERATIONAL CONDITIONS 1, 2, and 3 that primary containment integrity be maintained. .

"TS 3.6.1.3 requires in OPERATIONAL CONDITIONS 1, 2, and 3 that the primary containment air lock be OPERABLE with: (1) both doors closed except when the

, air lock is being used'for normal transit entry and exit through the containment, then at least one air lock door shall be closed; and (2) an i

l overall air lock leakage rate of less than or equal to 0.05L,at Pa '

49 psig.

"Contrary to the above, at 4:35 a.m. on April 26, 1988, Unit 2 entered [

OPERATIONAL CONDITION 2 when the unit's mode switch was placed in the  !

Startup/ Hot Standby position without RHR Division II being aligned for i

automatic LPCI initiation, without primary containment integrity being

( established, and with the primary containment air lock doors open."

Resconse to Violation A I. Admission Or Denial Of The Violation CP&L acknowledges the requirements of Technical Specifications 3.5.3.2, 3.6.1.1, and 3.6.1.3 were not met on April 26, 1988, when t"s unit entered OPERATIONAL CONDITION 2 for surveillance testing. s event was previously reports. in Licensee Event Report 1-88 015, A-1

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II. Reason For Violation While preparing to perform a reactor startup on April 26, 1988, the reactor mode switch was placed in the Startup position at 1435 to perform the rod worth minimizer (RVM) system surveillance Periodic Test (PT)-01.6.2 2 and the rod sequence control system (RSCS) surveillance test PT 01.6.1. Following completion of these tests, the mode switch remained in the Startup position until the actual reactor startup was begun at approximately 1600 hours0.0185 days <br />0.444 hours <br />0.00265 weeks <br />6.088e-4 months <br />. Section 1.0 of the Technical Specifications defines OPERATIONAL CONDITION as "...any one inclusive combination of mode switch position and average reactor coolant temperature as indicated in Table 1.2." Technical Specification 3.0.4 states:

"Entry into an OPERATIONAL CONDITION or other specified applicability shall not be made unless the conditions of the l limiting condition for operation (LCO) are met without relia;we on provisions contained in the action statements unless otherwise excepted."

A mode change is initiated by either placing the mode switch to another position or by changing the reactor coolant temperature. In addition to this definition, the TS provide specific conditions or situations where the OPERATIONAL CONDITION is not defined by these two parameters exclusively. An example of this is found in Footnotes #,

1. 1. , and *** associated with Table 1.2 of the Technical Specifications.

These footnotes include provisions allowing the reactor mode switch to be placed in an otherwise unauthorized position to perform a specified function while not changing operating modes. While one of these footnotes is being epplied, the operating mode remains the same as that established prior to moving the mode switch to the position allowed by the footnote.

The surveillance testing required by Technical Specifications 3/4.1.4.1 and 3/4.1.4.2 which relate to the RVM and the RSCS each have a Footnote

• associated with them which states:

i "Entry into Condition 2 and withdrawal of selected control rods l is permitted for the purpose of determining the operability of the RVM (RSCS) prior to withdrawal of control rods for the purpose of bringing the reactor to critically."

The operations staff believed that repositioning the mode switch for I the performance of the RWM and RSCS surveillance tests was allowed by Footnote *. The difference in wording between the footnote associated l

with TS Table 1.2, "... mode switch acy be placed in the STARTUP/ HOT ,

STANDBY (REFUEL) position. . . " and Technical Specifications 3/4.1.4.1 i and 3/4.1.4.2, "Entry into Condition 2...," was not recognized. This interpretation was found to be consistent within the Operation staff and had been in effect as long as anyone could remember. Moreover, even though the requirement to meet OPERATIONAL CONDITION 2 was not met, plant procedures and supervisory controls were in place to A2

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prevent control rod withdrawal for the purpose of bringing the reactor to criticality. The operations staff believed that they were in OPERATIONAL CONDITION 4 and knew that shutdown cooling was in service per TS 3.5.3.2 and that the drywell was open for maintenance per '

TS 3.6.1.1 and 3.6.1.3.

4 III. Corrective Actions Which Have Been Taken A Standing Instruction was issued on April 26, 1988 which-identified the failure to properly position the mode switch and provided the requirements to ensure proper mode switch-operation and OPERATIONAL CONDITION changes. A review was conducted of other Technical Specification notes to determine if similar problems existed. No problems were identified.  ;

A real-time training package on the reactor mode switch change event was developed and training conducted for operations personnel. In '

addition, procedure changes have been completed which provide controls j of the mode switch /0PERATIONAL CONDITION changes during the reactor startup process. ,

IV. Corrective Actions To Be Taken No further actions are required as a result of this event.

V. Date When Full Compliance Will Be Achieved CP&L is now in full compliance with the applicable requirements.

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ENCLOSURE 2 RESPONSE TO VIOLATION B Description of Violation The NRC's Notice of Violation states as follows:

"Technical Specification 6.8.1.a requires that written procedures shall be implemented for applicable procedures recommended in tppendix A of Regulatory Guide 1.33, November 1972. Appendix A requires operating procedures for the RER systcs. Operating Procedure, OP-17, RHR System Operating Procedure, Revision 76, implements this requirement and requires that the RHR heat exchanger outlet valve (Ell-F003A) be either in the fully open or closed position during the shutdown cooling mode.

"Contrary to the above, OP-17 was not fully implemented on May 11, 1988 in that valve Ell-F003A was used in a throttled position during the shatdown cooling mode on Unit 2."

Resnonse to Violation B I. Admission or Denial Of The Violation CP&L acknowledges that OP-17 was not fully implemented in that the Ell F003A valve was not in the open position as required by that procedure. It is noted that this event was identified by the licensee and that there was no safety significance to the event.

II. Reason For The Violation A root cause which led to the event is that operations personnel failed to recognize that throttling of the Ell F003A valve was not an evolution allowed by Operating Instruction (01)-01, paragraph 4.4, "Simple Evolutions."

An additional root cause which led to the event (inadvertent heatup) is believed to be an apparent design inadequacy within the shutdown cooling system. As noted below, during periods of low decay heat generation, the ability to throttle cooling systems (the Ell-F003A valve) to match heat load does not exist.

Prior to the event, shutdown cooling had been established using the "A" loop of RHR. Due to the low decay heat load and the inability to throttle the Ell-F003A valve by design (the valve logic allows only full-open or full closed), coolant temperature was maintained by opening and closing the Ell-F003A valve in accordance with OP-17. At approximately 1255 hours0.0145 days <br />0.349 hours <br />0.00208 weeks <br />4.775275e-4 months <br /> on May 11, 1988, it was determined that the B1

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. r E11-F003A valve-could not be opened with the control switch. To allow s maintenance personnel to troubleshoot and repair the problem, the valve needed to be de energized; however. this action would remove the ability to control the coolant temperature.

01-01 provides guidance to the operations personnel on evolutions that are considered "simple" and, therefore, do not require specific procedures for implementation. Examples of such evolutions are the-changing of chart paper,-venting a heat exchanger when the operator opens and closes the valves in a relatively shore period, and blowing down an air receiver. The operations staff on duty at.the time of this event believed that the manual throttling of the E11-F003A vaive with the breaker de energized met "simple evolution" criteria. With the breaker de energized, the valve would remain in the throttled position (maintaining coolant temperature ) until the repairs were completed, at which time, the breaker would be re energized and coolent temperature would again be controlled by OP-17, by opening and closin5 the E11-F003A valve.

During the troubleshooting and repair process, the E11-F003A valve was inadvertently closed and not recognized by the operations staff. The method which should have been used to throttle the E11-F003A valve was to initiate a temporary change to OP 17. This process would have required a safety ar.alysis and increased the potential for establishing controls for insuring the position of the E11 F003A [

valve.

III. Corrective Actions Which Have Been Taken Training has been initiated for the operations staff by.the Operations  :

Manager concerning what constitutes a "simple evolution," which is defined as an evolution not requiring a procedure.

1 IV. Corrective Actions To Be Completed -

An evaluation is to be conducted to determine if the E11 F003 valves in both trains of RHR for both units should be modif ted to make them throttle valves. This would require possible valve changeouts (gate valves to globe valves) as well as associated logic changes.

V. Date When Full Compliance Will Be Achieved The evaluation described above will be completed by October 3, 1988.

Any changes or modifications resulting from this evaluation will be scheduled through the normal canagement process. In addition, OP-17 will be revited by October 3, 1988, to provide procedural guidance on throttling the E11-F003 valves as required in the future. Any modification implemented in the future will be reflected in OP-17 through the plant modification process.

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ENCLOSURE 3 RESPONSE TO VIOLATION C Descrintion of Violation The NRC's Notice of Violation states as follows:

"Technical Specification 3.3.1 requires, as a minimum, that the rr. actor protection system (RPS) instrumentation channels shown in Technical Specifications Table 3.3.1-1 be operable. Accordingly, notation "b" of Technical Specifications Table 3.3.1-1 requires that while in OPERATIONAL CONDITION 5 "shorting links" be removed froid the RPS circuitry prior to and during the time any control rod is withdrawn.

"Contrary to the above, from 3:50 a.m. until 7:48 p.m. on March 8, 1988, with the reactor in OPERATIONAL CONDITION 5, Unit ? control rod 10 39 was in the fully withdrawn position and the shorting links were not removed from the RPS circuitry."

Response to Violation C I. Admission Or Denial Of Violetion CP&L acknowledges that control rou 10-39 was fully withdrawn with the shorting links not removed. This event was identified by the licensee and reported in Licensee Event Report 1 88-06. Tnere was no safety significance involved with this event since the plant is analyzed as safe with the highest worth control rod withdrawn. In addition, the refueling interlocks would prevent withdrawal of additional control rods.

II. Reason For The Violation The failure to insert control rod 10-39 prior to inserting tle shorting links was due to personnel error by the licensed operator.

The failure to recognize that control rod 10-39 as being withdrawn was impeded due to an inoperable position indicatien switch and a lack of training on a computer software ch.nge which displays control roc posi t. ions .

At 2025 hours0.0234 days <br />0.563 hours <br />0.00335 weeks <br />7.705125e-4 months <br /> on March 7,1988, the shorting lin's were removed to permit performance of Periodic Test (PT)-14.1,"Contro' Rod Operability Check," and PT-14.lA, "Control Rod Coupling Check and Control Rod Drive Testing." These tests were being performed on several control rods following maintenance during the outage. At 0029 hours3.356481e-4 days <br />0.00806 hours <br />4.794974e-5 weeks <br />1.10345e-5 months <br /> on March 8, 1988, control rod motion was secured with control rod 10 39 C1

being the last rod tested. Step 7.10 of PT-14.1A requires that the control rod being tested be returned to the fullzin position; however, this step did not require a sign off or independent verification. The operator failed to follow this step. At 0350 hours0.00405 days <br />0.0972 hours <br />5.787037e-4 weeks <br />1.33175e-4 months <br />, the shorting links were installed following a verification that control rods were inserted, thus initiating the violation. This condition existed until 2052 hours0.0238 days <br />0.57 hours <br />0.00339 weeks <br />7.80786e-4 months <br />, at which time control rod 10 39 was inserted by its individual scram switch.

As noted, a verification that control rods were inserted was performed prior to inserting the shorting links. Control rod 10 39 was not identified as being full out due to two problems: (1) the full-out position indication switch was inoperable daring this time period; and (2) the computer program for verifying the control rod position was modified during the outage prior to this event.

The position indication switch problem had been identified prior to this event and was scheduled to be repaired prior to unit startup.

Access to the drywell is required as these switches are located within the control rod drive unit. This switch provides a signal which energizes the full out indication (a red light) on the full core display on the control panel.

Prior to the refueling outage during which this event occurred, the computer program used to verify control rod positions was OD-7, option 2. This program would print out "48" for those control rods that were full out. Late in core life, most, if not all control rods are in the full out position, thereby making the OD 7, option 2 printout more difficult to review. To provide better human factored printouts for control rod position indication (to make them less "busy'), another OD 7 option was developed which would not print anything for a control rod at position 48 (full out). By doing this, a full core printout of control rod positions late in core life would only print values for those rods not fully withdrawn.

Confusion arose because the new option was oufined as OD-7, option 2, while the "old" OD-7, option 2 was renamed OD 7, option 3. The operators were not aware of this change at the time the incident occurred, so that when the printout of control rod position indications was reviewed, the operotors would have been looking for a "48." The printout that they saw showed all "0's" except for one small blank (no number), which was not identified. Training was scheduled to address this program change prior to the completion of the cutage.

III. Corrective Actions Which Have Been Taken The following corrective actions have been completed: l

1. Operations personnel involved with this event have been counseled.

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2. A Standing Instruction was initiated following identification of this event to inform operations personnel of the computer program change. This instruction has since been deleted as training has been provided.

3. PT-14.lA has been revised to require sign off and independent verification of control rod position following rod testing.

4. The full-out position switch was repaired prior to the unit startup.

IV. Action Which Will Be Taken No further actions are required as a result of this event.

V. Date When Full Compliance Will Be Achieved CP&L is now in full compliance with the applicable requireuents.

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