ML20215E447

Responds to NRC 861112 Notice of Violation & Proposed Imposition of Civil Penalty.Categorization of Event as Severity Level III Inappropriate.No Risk to Public Health & Safety Nor Cause for Significant Concern Noted

ML20215E447
Person / Time
Site:	Catawba
Issue date:	12/12/1986
From:	Tucker H DUKE POWER CO.
To:	Taylor J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE)
References
CAL, EA-86-147, NUDOCS 8612220305
Download: ML20215E447 (18)

Text

-

0 4-

. DUKE POWER GOMPANY e.o. mox am8s CHAMLoTTE, N.C. 28242 1o HAL H. TI.'CKER m .. ....m .

0 0 Df.,C 15 P 2 ' g a3 (704) ara-4s:n

..m...--,

,y. "  ;,

December 12,,1986

/j -

?

, f ',{ , sv-Mr. James M. Taylor, Director .,, '

Office of Inspection and Enforcement U.S. Nuclear Regulatory Commission ,

Washington, D.C. 20555 , ,

~~ ~

• f , , } <

,r .

+

g

Subject:

Catawba Nuclear Station -r Docket Nos. 50-413 and 504414 Re'aponse to Proposed Civil Penalty EA 86-147 w* . . _

,,,f '

- ,f .

Dear Mr. Tay1ori - 6 r -

f* f In acco'rd'nce a with 10 CFR'2.205, Duke'?ower Compa.ny (Duke) hereby files its answer to the *.4otice "of Violation and Proposed I.nposition of Civil Penalty issued by the NRC (Region II) en November 12, 1986. Duke's pesition is that the event which

~

gave rise to'the Notice of Violation (NOV) bis been incorrectly categorized by the NRC as a?"c' a uss'for significant cones ~rn" leading to a Severity Level III violation. 10 CFR'Part 2. Appendix C.III. Specific details in support of Duke's position-tre discussed at length in Attscheent 1. A re'sponse to the NOV is provided in Attachment 2. Briffly, Dub believes that categorization of the event cs deverit'y Level III is. ihappropriate because no risk to the public health and cafety was present, thus there,was no'dcause for sitaificant concern".

The' ev.ent at issue occurred during a pre-planned startup test (the Unit 2 Loss of Control Room Test)~when control of the Attaa Generator Power Operated Relief

~

Valv'es' (S/G PORVs) wah t(ansferred feba'ibe Control Room to the Unit 2 Auxiliary "Teedwinter Pump Turbine Control Panel (AFWPTCP). It was discovered, during the

/ course of the test, that certain v'alv4 controllers on the AFWPTCP were mislabeled.

. Ther event was discussed ind.RC' Inspection'Raport 50-413/86-25 and 50-414/86-27 and

[ [#, , . >' in Licensee Event Report 0,ER) 414/86-28. Various aspects of the Loss of Control

" Rooit Test and the modification to the S/G PORVs which resulted in the mislabeled controllers are discussed in Attachments 1 and 2. As implemented, a portion of this modification changed the fun'etion of the S/G'PORV controllel:a from pressure dsman'd to position demand". , Duke admits that the effect of this chxnge on the AFWPTCP labeling was not recognized during the modification process. On transfer of control of the S/G PORVs from the Control Room to the AFWPTCP, the S/G PORVs opened to appros %ately 75% of full open,recher than remaining closed as intended.

This caused the'aeactor Coolant System pressure and the pressurizer level to

_ ^ decrease. The operator at the AFWPTCP'promptly and correctly curvised that the *

' S/G PORVs were the cause of the pressure decrease. The operact i re:::7t i to close the S/G PORVs by raisihg the steam' pressure setpoint or. WOM controller.

+ ', f- O och0k13 eon +

~

M </-

- .. e - -

.n

Mr. James M. Taylor, Director December 12, 1986 Page 2 However, as a result of the changed function of the controller, which was not correctly labeled, this opened the S/G PORVs further. Approximately six minutes after the initial transfer of control for the S/G PORVs, Control Room personnel, who had been continuously monitoring the event and were fully aware of the fact that the S/G PORVs were open, and were at all times cognizant of all safety margins and operating parameters, directed that control of the unit be transferred back to the Control Room. This action unblocked safety injection and closed the S/G PORVs. Unit 2 then returned to stable conditions within approximately 5.5 minutes. There was no threat to the public health and safety, nor were any of the involved plant systems endangered.

It is important to note that the event occurred during a pre-planned startup test which by its very nature is conducted to detect errors and to discover any deficiencies in design, construction or operating procedures. A natural result of such testing is that errors or deficiencies are detected and corrected prior to routine operation. That being the case, it does not appear that such errors or deficiencies should be viewed as "significant concerns". To penalize Duke -- or Z" any other utility -- for detecting errors during a startup testing program is inconsistent with the very nature of testing. The NRC should consider whether, as a matter of policy, enforcement action based on errors detected during planned testing to discover such errors is warranted.

Moreover, because a pre-planned startup test was underway, Control Room personnel were alert to the fact that anomalies might exist in design, construction, or operating procedures; were aware of plant conditions at all times during the test; were in a position to take action; and did in fact take appropriate corrective actions in a tLaely manner. The NRC Staff highlights the fact that six minutes elapsed from the time the S/G PORVs opened until control was transferred from the AFWPTCP back to the Control Room. This was not due to any failure by Control Room operators to correctly analyze the situation, or to any dereliction of duty on their part. Rather, Control Room operators consciously allowed such a period of time to elapse, consistent with testing mode operations and NRC guidance, to give the auxiliary panel operators opportunity to ascertain the problem and take corrective actions. However, when it became apparent that the auxiliary panel operators could not correct the problem -- and well prior to any situation that could jeopardize the public health and safety or plant systems -- Control Room personnel took appropriate corrective action.

Finally, it should be recognized by the NRC Staff that no Technical Specification Limiting Conditions for Operatien were violated.

It is Duke's view that because the event arose from a test which was conducted to, and in fact did, determine conformity with design, construction, or operating procedures; that because there was continued awareness of the relevant operating parameters by Control Room personnel; that because Control Room personnel retained the continued ability to control (and did in fact control) the plant and protect the health and safety of the public and relevant plant systems, the violation is not one which rises to the level of "cause for significant concern" and thus should not be categorized as a Severity Level III. In Duke's view the violation-at most should be characterized as one which is a " failure to meet reguletory i -

a Mr. James M. Taylor, Director December 12, 1986 Page 3 requirements that have more than minor safety . . . significance" and thus should be categorized as Severity Level IV. ( See 10 CFR Part 2 Appendix C, Supplement I, D.3). Even if a Severity Level III Violation were appropriate for this case, which it is not, Duke believes that further review of the mitigation factors should result in 100% mitigation of the civil penalty.

In support of its position with respect to mitigation, Duke would point out that, as previously noted, the Staff's reference to a six minute delay as a factor weighing against mitigation is inappropriate because Control Room operators, plant personnel and other observers were continuously aware of all plant parameters and what actions were needed to be taken to stabilize these parameters. NRC personnel monitoring the test were also contemporaneously aware of the event. Based on the above, Duke believes 50% mitigation for prompt identification and reporting should be granted.

As acknowledged by the Staff in the November 12, 1986 NOV, Duke has also taken. _,

prompt and extensive corrective actions regarding this event. These actions were~

discussed in response to the NRC's July 3, 1986 Confirmation of Action Letter '

(c4T) issued by Region II. The corrective actions were extensive in that they addr.ased specific concerns highlighted by the June 27, 1986 event in addition to plant-wide human factors and procedure development deficiencies. As documented in Inspection Report 50-413/86-27 and 50-414/86-30, all corrective actions relating to Item I of the CAL vere completed prior to the successful performance of the Loss of Control Roon Test on July 11, 1986. Duke believes these actions warrant the full 50% mitigation allowed by the Enforcement Policy.

The Staff alleges deficiencies in Catawba's past performance as justification for not mitigating the proposed civil penalty, citing a Catawba Unit 2 cold hydrostatic test that resulted in the overpressurization of the residual heat removal system piping and the volume control tank. First, these events were of minimal safety significance as reflected by a Severity Level IV Notice of Violation issued on May 30, 1985. In any event, as explained in Attachment 1, the root cause of the cold hydrostatic test violations was unrelated to the root cause for the Loss of Control Room Test. Therefore, Duke believe3 that reliance on this event as a basis for,not allowing mitigation of the civil penalty is inappropriate as it does not reflect prior poor performance in the area of concern. In fact, Catawba has a good prior enforcement history as evidenced by the most recent SALP report dated December'19, 1985. In accordance with the Enforcement Policy, this history merits 100% mitigation of the civil penalty.

Duke would like to emphasize that, even though no Technical Specification Limiting Conditions for Operation were violated and there was no significant impact on the ability to maintain the health and safety of the public as a result of this incident, the Company both recognizes and appreciates the potential seriousness of incidents of this nature. Duke believes that its record in identifying and correcting potential problem areas in its operations is a good one, and will continue in the future to make every effort to assure that it continues in that fashion. The Notice of Violation states that the civil penalt" was proposed "to emphasize the importance of complete and thorough reviews of . sign changes, the' a

Mr. James M. Taylor, Director December 12, 1986 Page 4 necessity of adequate procedures, and procedural adherence." Duke believes that the corrective actions taken in regard to this incident demonstrate the company's concern with safe operation of the Catawba plant.

In conclusion, Duke reiterates the following arguments: (1) the safety significance of the event warrants a Severity Level IV instead of a Severity Level III violation, and (2) even if Severity Level III were appropriate there are sufficient mitigation factors to completely eliminate any associated civil penalty and thereby requests said mitigation accordingly. It should also be noted that the Notice of Violation was issued for Catawba Units 1 and 2 although Unit I was not at all involved in the subject events.

I declare under penalty of perjury that the statements set forth herein are true and correct to the best of my knowledge.

Very truly yours, ,

1. Ml QFf Hal B. Tucker ROS/(LTP/14) /j gm/slb xc: Dr. J. Nelson Grace, Regional Administrator U.S. Nuclear Regulatory Commission - Region II 101 Marietta Street, NW Suite 2900 Atlanta, GA _30323 NRC Resident Inspector Catawba Nuclear Station

-4

Page 1 ATTACHMENT 1 DUKE POWER COMPANY CATAWBA NUCLEAR STATION COMMENTS ON NOTICE OF VIOLATION AND PROPOSED IMPOSITION OF CIVIL PENALTY (EA 86-147)

The NRC's Enforcement Policy (10 CFR Part 2, Appendix C) is designed to promote and protect the radiological health and safety of the public by:

1) Ensuring compliance with NRC regulations and license conditions;

2) Obtaining prompt correction of violations;

3) Deterring future violations; and

4) Encouraging improvement of licensee performance including prompt identification and reporting of potential safety problems. ,

Relying on the Enforcement Policy, the NRC has proposed that the subject incident-be classified as a " Severity Level III problem," has proposed a civil penalty of

$50,000 and has preliminarily determined that mitigation is not " deemed appropriate in this case." Enforcement Letter, p. 2. The basis for the Staff's action presumably is the need "to emphasize the importance of complete and thorough reviews of design changes, the necessity of adequate procedures and procedural adherence." Id.

The Enforcement Policy states that Severity Level III violations are "cause for significant concern." 10 CFR Part 2, Appendix C.III. As set forth below Duke takes issue with the Staff's proposed action in that the incident did not give rise to a "significant concern." Rather, Duke believes that this incide: should be viewed as " failure to meet regulatory requirements that have more than minor safety . . . significance," which would place it as a Severity Level IV violation.

10 CFR Part 2, Appendix C. Supplement I.D.3. In the event the Staff continues to believe this incident is a Severity Level III, it is clear that mitigation is appropriate and the amount should be reduced to zero.

Three reasons support Duke's assertion that this incident was not a "significant concern" and thus not a Severity Level III incident. First, Duke was conducting a pre-planned startup test. It is the nature of testing to detect errors; to identify any deficiencies in the design, construction or operating procedures prior to the unit going into routine service. A natural result of testing is that identified errors are corrected prior to plant operation. Therefore, matters detected and corrected in a normal pre-planned startup test mode should not be viewed as "significant concerns." To penalize Duke for detecting errors is inconsistent with the concept of testing. Duke appreciates what it views to be the underlying staff concern in this regard; i.e., that even though the test detected the errors and the errors were corrected, the functions involved were important plant functions and Duke should be emphatically told that errors regarding such functions are of significant concern to the Staf f. In response, Duke stresses the nature of testing, which simply put, is utilized to detect and-correct deficiencies on both significant and less significant systems.

Irregularities are routinely found in plants as complex as nuclear power plants i

I

Attachment i Page 2 and while Duke is desirous of eliminating such irregularities prior to testing, such is unrealistic to expect. Sometimes the irregularities are significant; sometimes they are insignificant. However, as long as a process is in place to detect these irregularities, which it was in this case, the Staff can take comfort that safety concerns are properly being addressed at the testing stage. Thus, while errors were detected in this case, these errors should not be viewed as raising significant concerns. To penalize utilities for detecting errors undercuts testing programs.

Second, because the pre-planned startup test was underway, Control Room personnel were alert to the fact that anomalies might exist in design, construction or operating procedures; were aware of plant conditions at all times throughout the test; were in a position to take action; and did in fact take appropriate corrective actions in a timely manner. As a preliminary basis for categorizing the event as a Severity Level III and imposing civil penalties, the Staff high-lights the significance of 6 minutes elapsing from the time the steam generator PORVs opened until operator action was taken to transfer control to the Control, Room. Control Room operators and supervision were fully aware of plant conditions'I throughout the 6 minute period. Consistent with NRC guidance, Control Room personnel were determined to give the operators at the auxiliary panels an op-portunity to ascertain the problem and to take corrective action. Such a practice is consistent with testing mode operations. However, when it became apparent that the operators at the auxiliary panels could not correct the situation, and well prior to any situation that would jeopardize the public health and safety, the Control Room personnel took corrective actions. The Staff states the "it is fortuitous that there were no adverse thermal-hydraulic or nuclear effects on the plant because of this design control failure." This statement is not consistent with the NRC's Augmented Inspection Team Report issued July 25, 1986 where it states on page 9 that the transient did not result in a violation of the Technical Specification cooldown rate or shutdown margin. The Shift Technical Advisor (STA) and Operator at the Controls (OATC) confirmed that the reactor vessel remained

, full and the reactor coolant pumps continued to operate without indications of problems. The status of the vessel's water level and reactor coolant pump's operation are prime indicators for potential thermal-hydraulic and nuclear ef-fects. Since these parameters were under control at all times, it does not appear that the absence of adverse effects was " fortuitous."

Third, no Technical Specification Limiting Conditions for Operation were violated.

i If the NRC maintains that a Severity Level III violation is appropriate for the subject violations, Duke believes that mitigation is appropriate. The NRC has stated otherwise, noting "the occurrence of previous testing deficiencies . . ."

Enforcement Letter, p. 2. As will be shown in the discussion of the mitigation factors, the preliminary Staff position is in error.

l

Attachment 1 Page 3

1. Prompt-Identification and Reporting The Loss of Control Room Test is a major event in the Power Escalation Test Program. Since the test was witnessed by NRC/ Region II personnel, including the NRC resident inspector, the NRC was immediately made aware of the event.

Notification to the NRC Operations Center was also promptly made. Additional notification was provided in LER 414/86-28 in a letter to the Staff dated July 25, 1986.

Therefore a 50% reduction in the proposed civil penalty is warranted based on this factor.

2. Corrective Action to Prevent Recurrence As a result of the June 27, 1986 event, extensive discussions were held between NRC/ Region II and Duke personnel. These discussions were followed by a July 3,1986. Confirmation of Action Letter (CAL) from NRC/ Region II. The .T*

corrective actions were extensive and included both Catawba Units 1 and 2. -

As documented in NRC Inspection Report 50-413/86-27 and 50-414/86-30, all corrective actions in Item 1 of the CAL were completed prior to the successful performance of the Loss of Control Room Test on July 11, 1986.

Other corrective actions are acknowledged in NRC Inspection Report 413/86-36 and 414/86-39. The November 12, 1986 Notice of Violation acknowledges the prompt and extensive corrective actions taken. Therefore a 50% reduction of the proposed civil penalty is warranted based on this factor.

3. Past Performance The November 12, 1986 Notice of Violation cited the Catawba Unit 2 cold hydrostatic tests, that resulted in overpressurization of the residual heat removal system piping and the volume control tank, as the NRC's basis for not mitigating the civil penalty. The hydrostatic test violations were issued as Severity Level IV in Inspection Report 50-413/85-16 and 50-414/85-12 dated May 30, 1985.

A descriptioE of the overpressurization events is provided in Section 3.9.2 of Supplement 5 to the Catawba Safety Evaluation Report dated February 1986.

It is evident from this description that the root cause of both cold hydrostatic overpressurization events was personnel error. As LER Report 414/86-28 concluded, the root cause of the Loss of Control Room Test transient was design deficiency and defective procedure. Additional investigations revealed that inadequate communications between plant and design personnel was a contributing cause.

l Therefore, the two occurrences did not have similar root causes and the j overpressurization events did not provide any lessons-learned that could have  !

prevented the Loss of Control Room Test transient. In addition, since the cold hydrostatic test event was categorized as a Severity Level IV violation, its significance in past performance considerations is further minimized. It J is our conclusion that the previous overpressurization events are unrelated i I

Attachment 1 Page 4 to this event and therefore not a justifiable basis for not mitigating the proposed civil penalty; therefore, 100% mitigation of the civil penalty based on this factor is warranted.

4. Prior Notice of Similar Event There have been no prior notices of any events that would have provided sufficient knowledge to prevent this violation.

5. Multiple Occurrences The Loss of Control Room Test is a one-time test . The PORV modification that caused the transient during the test was an atypical change. There have not been additional occurrences relating to this violation.

M *w O

4

- - - _ - . - - ,. -. -er-, -----, - y n - . g ---

.c- -

---

r_-- - . .

r Attachment 1 Page 5 Additional Comments It is requested that the following additional comments on the November 12, 1986 Notice of Violation be considered.

1. In the second paragraph of the transmittal letter, it is stated that Violation A involves a significant failure of the design control program. It should be understood that the Design Change Authorization (DCA) program under which the modification to the steam generator PORV's was being implemented was a pre-operational program. This program was replaced by the Nuclear Station Modification (NSM) program prior to each unit going into operation.

The NSM program provides for greater involvement of station personnel in the modification process. This would tend to preclude the type of problem that occurred with the S/G PORV modification. Therefore the emphasis on design changes that is intended by imposition of a civil penalty is lost on a DCA program that is no longer in use.

2. The third paragraph of the transmittal letter discusses identified deficiencies in the procedures used for the Loss of Control Room Test. As discussed below, these supposed deficiencies were either insignificant to the transient, are criticisms t'at go beyond the design basis of the plant, or are beyond the scope of th est.

a) Improper settings of charging flow and seal injection flow control valves - These identified deficiencies are not considered to be significant contributors to the transient and are discussed more fully in Attachment 2, Response to Example B.1.

b) Test procedures not containing specific termination criteria - As discussed in Attachment 2, Response to Example B.2, there are no regulatory requirements for termination criteria in test procedures.

The AIT report at page 22 acknowledges that Step 6.10 of the test procedure stated "if any situation arises that cannot be adequately controlled from the auxiliary shutdown panels, transfer control back to the Control Room". In Duke's opinion this caution is adequate and appropriate. An attempt to cover every possible contingency would add unnecessary volume without any assurance of covering all occurrences.

Instead, reliance is appropriately placed on the training and judgement of the licensed operators who are conducting the test.

c) Safety injection cannot be initiated at the remote shutdown panels - The design of Catawba's auxiliary shutdown panels was reviewed and approved by the NRC in Section 7.4.2.2 of the SER and SSER 1. As noted in the SER, no accident is assumed to have occurred. Therefore inclusion of procedures for actuation of safety injection systems in the test procedure was beyond the design basis of the plant. However such instructions were added to the test procedure prior to repeating the test on July 11, 1986.

Attachment 1 Page 6 d) Pressurizer level and pressure not available to the operator - The limitation of narrow range pressurizer pressure (1700 to 2500 psig) had been previously recognized as HED No. C-1-0102 and is documented in Duke's Response to Supplement I to NUREG-0737. This HED was scheduled for correction during the first refueling outage in accordance with License Condition 7. The test procedure was adequate for its intended purpose since the Reactor Coolant System temperature was to be reduced by 50*F in accordance with FSAR Table 14.2.12-2 (Page 32). At that temperature, the pressurizer level and pressure would have been on-scale.

e) Labeling of valve controllers - These identified deficiencies are not considered to be significant contributors to the transient and are discussed in Attachment 2, Item B.3.

m -

~

(LTP14)

ATTACHMENT 2 RESPONSE TO THE NOTICE OF VIOLATION NRC INSPECTION REPORT 50-413/86-25 AND 50-414/86-27 CATAWBA NUCLEAR STATION Violation A as Stated:

A. 10 CFR Part 50, Appendix B, Criterion III Design Control, requires that measures be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions.

Contrary to the above, as of June 27, 1986, the licensee's program for design control did not assure that Design Change Authorization (DCA) CN-2-M-1527, which changed the design basis for the mode of control for the steam generator power operated relief valves (S/G PORV's), was correctly translated into specifications, drawings, procedures, and instructions. Specifically: 7"

1. DCA CN-2-M-1527 was not properly reviewed by plant personnel as required by Station Directive 3.0.3, Management of Shutdown Requests, for the effects on Operating Procedure OP/2/A/6100/04, Shutdown Outside the Control Room from Hot Standby to Cold Shutdown. As a result, Enclosure 4.5 of Procedure OP/2/A/6100/04 was not modified and incorrectly specified the setpoint of the S/G PORV's. Instead of remaining closed the S/G PORV's opened to approximately 75 percent of full open.

2. DCA CN-2-M-1527 was not properly reviewed by design personnel as required by Design Engineering Procedure EDP 3.17, Control Room Change Handling, in that there was no evaluation done for human factors considerations. The DCA was implemented and changed the control mode of the S/G PORV's without any visible change to the S/G PORV controllers or labeling at the Auxiliary Feedwater Pump Turbine Control Panel.

Response: --

1. Admission or Denial of Violation Duke Power Company admits the yiolation as stated in A., but denies th 4t the deficiencies related to the modification of the steam generator PORV's constituted programmatic deficiencies. Rather, Violation A reflects an isolated failure in the design review of an atypical modification and the subsequent failure to recognize the impact on station procedures.

Furthermore, examples A.1 and A.2 do not accurately reflect events related to the subject modification.

Attachment 2 Page 2

2. Reasons for Violation if Admitted DCA CN-2-M-1527 was initiated in January 1985 for the purpose of upgrading the S/G PORV's to safety-grade in accordance with previous commitments to the NRC. As a part of this DCA, the function of the S/G PORY control on the Auxiliary Feedwater Pump Turbine Control Panel (AFWPTCP) was changed such that the PORV's opened or closed in direct response to the position dialed-in on the controller. Previously..the controllers had provided a pressure setpoint selection such that the PORV's opened when the setpoint was exceeded.

Example A.2 states that there was no evaluation done of DCA CN-2-M-1527 for human factors consideration. This DCA was, in fact, reviewed in accordance with EDP-3.17 as documented by a Control Room change turnover certification dated July 8, 1985. The acknowledged oversight was in failing to revise other design documents to reflect needed scale changes and panel labeling.

We believe this error was due in great part to the atypical nature of the - -.

modification-nohardwarewaschangedontheAFWPTCP,ratherthefunctionof the PORV controllers was changed. It was not a failure to review the DCA from a human factors standpoint.

Example A.1 states that DCA-2-M-1527 was not properly reviewed for its effect on Operating Procedure OP/2/A/6100/04. Duke believes that station personnel did exercise due diligence in reviewing the subject DCA for required changes to station procedures in accordance with Station Directive 3.0.3. Due to the root cause failure discussed above, the needed change to PORV controllers was not clearly identified in the subject DCA. Therefore example A.1 should not be considered an independent example of Violation A, but rather, a result of the deficiency in the design documents.

3. Corrective Actions Taken and Results Achieved A. A review of all Unit 2 design changes and shutdown requests implemented af ter Hot Functional Testing and prior to Fuel Load was conducted prior to restart of Unit 2.

B. A review of' Main Control Board, Auxiliary Shutdown Panels (ASP's) and AFWPTCP's fo,r both units was performed to identify differences between units, and to verify proper labeling nomenclature and units of measure.

This was accomplished prior to restart of Unit 2.

C. The ASP operating procedure and the Loss of Control Room abnormal procedure have been revised to reflect the changes to the panels. A new procedure. TT/2/A/9100/03 Auxiliary Shutdown Panel and Supplemental Test, was written to verify proper function of the various valves while at the ASP. This test was performed satisfactorily prior to the Loss of Control Room retest on July 11, 1986.

Attachment 2 Page 3 D. Design procedure EDP-3.17 (Control Room Change Handling) was revised to clarify the need for review of modifications to the ASP and the AFWPTCP and to clarify responsibility for initiating the Control Room Change Form. I&C Workplace procedure PR-3 has been revised to assure emphasis on labeling and scaling of manual loaders, controllers, etc.

E. As required by the Confirmation of Action Letter of June 27, 1986, Duke reviewed all Human Engineering Deficiencies (HEDs) identified in NSM-CN-20227 and their schedules for implementation. As a result, complete re-engraving of all nameplates on the ASPS and AFWPTCP was accomplished prior to August 22, 1986. As discussed in a July 30, 1986 letter, the remaining portions of NSM-CN-20227 could not be implemented at that time since the remaining section requires de-energizing part or all of the systems on the ASPA or ASPB or the AFWPTCP. The design will be complete by 5/4/87. Implementation could follow during a Unit 2 outage (af ter 5/4/87) of sufficient duration prior to the first refueling outage. In any event, all HED's are required to be correct'edII prior to restart from the first refueling outage in accordance with -

License Condition 7 of Facility Operating License NPF-52.

F. Training has been provided to appropriate personnel and included a description of this incident, procedural changes, labeling and surface changes made to the ASP's and Unit 1/ Unit 2 control differences.

4. Corrective Actions to be taken to Avoid Further Violations No additional actions are required.

5. Date of Full Compliance Duke Power Company is in full compliance with regulatory requirements.

a Attachment 2 Page 4 Violation B as Stated:

B. 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings.

Contrary to the above, as of June 27, 1986, certain procedures affecting quality were inadequate or not properly implemented and contributed to the depressurization event of June 27, 1986. Specifically:

Example 1

1. Operating Procedure OP/1/A/6100/04, Shutdown outside the Control Room ,7*

from Hot Standby to Cold Shutdown, was inadequate in that Enclosure 4.5 to the procedure specified initial settings for valves 2NV-309 and 2NV-294 which resulted in these valves going to incorrect positions when control was transferred to the remote shutdown panels (Auxiliary Shutdown Panels A and B and the Auxiliary Feedwater Pump Turbine Control Panel).

Re sponse :

1. Admission or Denial of Violation Duke Power Company denies the violation as given by example 1. The settings for valves 2NV-309 and 2NV-294 were established based on the predicted flow requirements and the controller test data.

The Catawba Nuclear Station FSAR Summary of Test Program and Objectives (Section 14.T.1) states, in part:

"The General Objectives of the Initial Test Program at Catawba Nuclear is to provide assurance that: Operating and emergency procedures are appropriate to the extent practicable." (emphasis added)

Test Procedure TP/2/A/2650/03 Loss of Control Room Functional Test, was being conducted to meet this objective for the use of OP/1/A/6100/04. The initial positions of valves 2NV-309 and 2NV-294 had been specified on the basis of the best available information. Deficiencies identified as a result of this testing are not in themselves violations unless some other regulatory requirement had been violated. -

Attachment 2 Page 5

2. Reasons for Violation if Admitted

,Not Applicable

3. Corrective Actions Taken and Results Achieved As a result of Test Data gathered from TT/2/A/9100/03, Auxiliary Shutdown Panel and Turbine Control Panel Supplemental Test, the initial positions of 2NV-294 and 2NV-309 have been adjusted in OP/1/A/6100/04, Enclosure 4.5.

4. Corrective Actions to be Taken to Avoid Further Violations No additional actions are required.

5. Date of Full Compliance Duke Power Company is in full Compliance with regulatory requirements. . _,

~

m =

O

O Attachment 2 Page 6 Example 2

2. Test Procedure TP/2/A/2650/03, Loss of Control Room Functional Test, was inadequate in that the procedure lacked specific criteria for terminatica of the test if the plant was determined to be in an uncontrolled or uncontrollable condition with control transferred to the remote shutdown panels.

Response

1. Admission or Denial of Violation Duke denies this example is a violation of regulatory requirements.

Duke has not. identified any NRC requirement for test procedure termination 7*

criteria. Neither Regulatory Guide 1.33, Quality Assurance Program -

Requirements (Operation), its reference industry standard ANSI N18.7-1976, Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants, nor Regulatory Guide 1.68. Initial Test Programs for Water-Cooled Nuclear Power Plants, specifically require termination criteria in test procedures. The essentially identical Unit 1 Loss of Control Room Test procedure was reviewed and approved by the NRC (Inspection Reports 50-413/84-40 and 50-413/84-66) with no mention of termination criteria. Step 6.10 of the test procedure stated "if any situation arises that cannot be adequately controlled from the auxiliary shutdown panels, transfer control back to the Control Rcom." In Duke's opinion this caution was adequate and appropriate.

2. Reasons for Violation if Admitted Not Applicable

3. Corrective Actions Taken and Results Achieved Even though Duke agreed to add termination criteria to the Unit 2 test procedure prior to rerunning the test, such criteria are not considered necessary. Instead, reliance is placed on the licensed operators to monitor unit conditions and take any necessary actions.

4. Corrective Actions to be Taken to Avoid Further Violations No additional actions are required.

5. Date of Full Compliance Duke Power Company is in full compliance with regulatory requirements.

Attachment 2 Page 7 Example 3

3. Operations Management Procedure OMP l-6, Control Panel Information Changes, dated May 10, 1982, was not followed by plant operations personnel in the labeling of valves 2NV-294 and 2NV-309 at the remote shutdown panels. OMP 1-6 states that changes to the control panel will conform to human factor guidelines and conventions. However, the labels placed on the controllers for valves 2NV-294 and 2NV-309 were reversed and indicated opposite of the intended and anticipated meaning.

Response

1. Admission or Denial of Violation Duke Power Company admits Example 3 of Violation B with clarification. '

i*

Duke admits that the labels placed on the controllers for valves 2NV-294 and 2NV-309 were reversed and indicated the opposite of the intended and anticipated meanings. However, Duke denies that the mislabeling was a result of a failure to follow Operations Management Procedure OM7 1-6.

The effect of the valve mislabeling was not significant to the cooldown event.

2. Reasons for Violation if Admitted The mislabeling of the subject valves occurred due to the lack of a formal mechanism for labeling the valves. As a result of the Control Room Design Review, the station had been requested to add "O" and "C" tags to the controllers on the Main Control Board and the Auxiliary Shutdown Panels. It was left up to the station to determine the correct placement of the tags.

The labels should have been pidced in accordance with a formal change -

either an NSM or a work request.

Operations Management Procedure OMP l-6 provides a mechanism for review and approval of amplifying information to be provided on Control Board components. This procedure was not intended for implementing labels such as the "0" and "C" tags.

3. Corrective Actions Taken and Results Achieved The labels were removed and reinstalled correctly. OMP 1-6 was used to document the re-installation of these labels.

4. Corrective Actions to be Taken to Avoid Further Violations All functional controllers on each unit's main control panel, Auxiliary -

Shutdown Panels and Turbine Driven Auxiliary Feedwater Pump Control Panel were reviewed with respect to valve position nomenclature and units of

r s

Attachment 2 Page 8 measure to verify correct labeling. Any required changes were reflected in appropriate procedures and training. These actions were complete before Unit 2 was restarted on July 7,1986.

5. Date of Full Compliance Duke Power Company is in full Compliance with regulatory requirements.

~

?

E e

,96