BSEP-88-0099, Responds to Violations Noted in Insp Repts 50-324/87-31 & 50-325/87-32.Corrective Actions:Governing document,ENP-20, for Engineering Work Request (EWR) Program Extensively Revised & Team to Work on EWRs Created

Responds to Violations Noted in Insp Repts 50-324/87-31 & 50-325/87-32.Corrective Actions:Governing document,ENP-20, for Engineering Work Request (EWR) Program Extensively Revised & Team to Work on EWRs Created

ML20196C613
Person / Time
Site:	Brunswick
Issue date:	02/09/1988
From:	Dietz C CAROLINA POWER & LIGHT CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
BSEP-88-0099, BSEP-88-99, NUDOCS 8802160187
Download: ML20196C613 (14)

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o CP&L Carolina Pomr & Light Company Brunswick Steam Electric Plant P. O. Box 10429 Southport, NC 28461-0429 February 9,1988 FILE: B09-13510C SERIAL: BSEP/88-0099 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 BRUNSWICK STEAM ELECTRIC PLANT UNITS 1 AND 2 DOCKET NOS. 50-325 AND 50-324 LICENSE NOS. DPR-71 AND DPR-62 R_ESPONSE TO INFRACTIONS OF NRC REQUIREMENTS Gentlemen:

The Brunswick Steam Electric Plant (BLZP) has received I&E Inspection Reports 50-325/87-32 and 50-324/87-31 and finds they do not contain information of a proprietary nature. This report identified four items that appeared to bo in noncompliance with NRC requirements. Enclosed is Carolina Power & Light Company's resronse to those violations and the status of selected Engineering Work Requests requested by the audit team.

Per a telephone conversation on December 22, 1987, between Mr. K. E. Enzor of my office and Mr. S. J. Vias of NRC Region II, a 30-day extension was granted for submittal of the response to the violations.

Very truly yours, Y

C. R. b.etz, Gener 1 Manager Brunswick Steam Electric Plant RMP/byc Enclosure cc: Dr. J. N. Grace Mr. E. D. Sylvester BSEP NRC Resident Office 8802160187 880209  ; i PDR ADOCK 05000324 i O PDR P

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VIOLATION A 10CFR50, Appendix B, Criterion XVI, and the licensee's accepted Quality Assurance Program, Final Safety Analysis Report (FSAR), Section 17 a.16, collectively require that measures be established to assure that conditions adverse to quality shall be promptly identified and corrected.

Contrary to the above, although measures have been established for identifying conditions adverse to quality, measures have not been established to assure these conditions are promptly corrected in that there exist approximately 1900 Engineering Work Requests (EWRs) that are currently outstanding. A detailed review of 30 EWRs identified multiple exampics of EWRs that had been written in 1979-1984 with the corrective action remaining unresolved. Typical examples include EWRs83-083, 84-362,82-294, 84-1024,83-417, 84-872A, 83-1100, 84-809A,83-463, 83-168,80-272, 84-170,84-719, 02151,84-685, and 84-472. This list is not intended to be all inclusive.

This is a Severity Level IV violation (Supplement I).

RESPONSE TO VIOLATION A I. Admission or Denial of the Alleged Violation Carolina Power & Light Company (CP&L) acknowledges that the violation occurred as stated.

II. Reason for the Violation CP&L has identified two root causes for this violation:

A. Application of site engineering resources to completing higher priority requirements, and B. Failure of the EWR management system to raise for management review / reassessment those Engineering Work Requests (EWRs) overdue past their planned completion date.

The net result of these two factors has been the development of a large backlog of engineering work, a great deal of which is,of only minor significance. At the time of the quality verification audit addressed in IER 50-325/87-32 and 50-324/87-31, CP&L management personnel could not address with certainty the contents of that backlogged work nor the safety significance thereof. Although a high confidence that no items of safety significance existed, appropriate management level reviews had not been completed to ensure that proper consideration was given to timely processing of important issues within the backlog. In some cases, budgetary limitations were cited as being a constraint by personnel interviewed by the inspection team. In fact, the process lacked i sufficient management visibility to enable proper budgetary l decisions to be made.

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III. Corrective Steps Which Have Been Taken A. A review of the Brunswick plant corrective action programs was requested and completed by a corporate level team. This review was completed in December 1987 and resulted in several meaningful recommendations regarding the EWR program.

B. The governing document (ENP-20) for the EWR program was extensively revised to:

1. Include pertinent recommendations from the aforementioned corporate level review.

2. Provide a documented initial review of EWRs for cafety significance. This review, completed by the originator's supervisor and backed up by an Engineering Project Engineer, provides an initial assurance that safety and operability issues are handled in a timely manner. If these reviews indicate the potential for an EWR to affect safety or operability, the issue is referred to appropriate personnel for further real-time review in accordance with existing plant procedures.

3. Provide a formal, documented safety and operability assessment including a written basis evaluating the significance of the item. This evaluation, completed by the assigned evaluating engineer, constitutes the formal documented review required by 10CFR50, Appendix B, Criterion XVII.

4. Provide a systematic method of assigning priority and action target dates for EWRs based on their nature and significance.

5. Provide for a monthly report to management on EVRs overdue past their assigned target due date so that management action can be taken to reassess the need to either raise the priority of the EWR or else defer its accomplishment until some future date based on its low significance relative to other engineering issues being addressed.

C. A separate team (nine personnel and one project team leader) was created from existing personnel resources to work exclusively on EWRs to provide focused attention to problems identified therein. The first priority of this teaa is to review incoming EWRs, document the ,

results of that review, and give the new EWRs appropriate relative  !

attention. The second priority task is the rereview of existing i EWRs to the criterion and procedures of the revised EWR program j instruction. Thus, in time, the existing backlog will either be ,

dispositioned or else enveloped by the vastly improved processing afforded by the revised ENP-20. The size and composition of the EWR l project team will be adjusted in the future as workload demands. I 1

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. i D. CP&L reaffirms its commitment to fund:and/or provide resources to l work significant safety issues raised by_the EWR process. The '

budgetary problems cited by individual engineers, while real from their perspective, did not and do not exist where safety is concerned. CP&L management personnel are sensitive to the need to  ;

ensure that significant safety and operability affecting issues are j dealt with promptly,and responsibly, j e

IV. Date When Full Compliance Will Be Achieved  !

I CP&L considers that full compliance will be achieved when the rereview of the existing backlog is completed and properly documented. CP&L will monitor the progress of the EWR team charged with that task and advise the Commission by May 30, 1988, of the projected completion date for_that

! effort. Earlier commitment is not considered prudent because the full productive potential of the group will not be realized for two to three [

months and accurate projections of task completion would thus be biased.

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VIOLATION B 10CFR50, Appendix B, Criterion XVII, and the licensee's accepted Quality Assurance Program (FSAR, Section 17.2.17) collectively require that sufficient records shall be maintained to furnish evidence of activities affecting quality. The records shall include . . . the results of reviews . . .

Contrary to the above, the results of reviews have not been documented as required for various EWRs. The results of the reviews for EWRs were not documented to demonstrete that the known deficiency did not impact plant safety, thereby justifying continued plant operation. Typical examples include EWRs84-685, 84-496,83-463, 83-168, 01334,84-472, 84-391, 84-1024,79-399, 84-809A,84-362, and 82-294. This list is not intended to be all inclusive.

This is a Severity Level IV violation (Supplement I),

RESPONSE TO VIOLATION B I. Admission or Denial of the Violation CP&L acknowledges that the violation occurred as stated.

II. Reason for the Violation CP&L failed to recognize the Engineering Work Request (EWR) system as a corrective action program which must conform to 10CFR50, Appendix B, particularly Criterion XVII. As a result, the requirement for documented initial ansessments of EWRs was not required by ENP-20, the governing ,

procedure for EWR processing and management.

III. Corrective Actions Taken and Results ENP-20 was extensively revised to document an initial safety / operability assessment both by the EWR originator's supervisor and by an Engineering group Project Engineer. This initial assessment is completed at the front end of the EWR initiation process to ensure safety / operability concerns raised by the EWR (where existing) are promptly identified and escalated for appropriate action through established plant procedures.

ENP-20 now also provides a detailed, formal, documented assessment form for EWRs which provides a documented basis evaluating the safety and operability significance of the EVR. This review, completed by the investigating engineer, follows the initial assessment above within thirty days and provides the documented basis required to establish that the plant can safely continue to operate with the problem cited in the EWR.

Recognizing the need to document the safety significance of the existing backlog pending final action to disposition the EWRs therein, CP&L has established a dedicated nine-man team (plus one project team leader) to work full time on EWRs. One of the taskings of this group will be the rereview of the existing backlog to the new review and documentation criterion and procedures of ENP-20.

IV. Date Full Compliance Will Be Achieved CP&L considers that full compliance will be achieved when the existing backlog is rereviewed to the new screening criteria of ENP-20. CP&L will advise the Commission by May 30, 1988, of the date by which this rereview will be complete.

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VIOLATION C

, 10CFR50.55a(g) requires for a boiling water-cooled nucicar power facility whose construction permit was issued prior to January 1, 1974, that components shall meet the requirements of paragraph (g)(4) to the extent practical.

Paragraph (g)(4) requires that throughout the service life of a boiling water-cooled nuclear power facility, components which are classified as American Society of Mechanical Engineers (ASME) Code Class 1, 2, or 3 shall meet the requirements set forth in Section XI of editions of the ASME Boiler and Pressure Vessel Code (B&PVC) and Addenda. ASME BSPVC,Section XI, Part IAW 7210, allows replacements (including system changes) to be accomplished with the original design requirements. The original design requirements for the Core Spray System committed to by the licensee are United States of America Standard B31.1, Power Piping. Paragraph 102.2.5B of this standard states that:

"Exhaust and pump suction lines for any service and pressure shall have relief valves of suitable size unless the lines and attached equipment are designed for the maximum pressure to which they may accidentally or otherwise be subjected or unless a suitable alarm indicator, such as a whistle or free blowing relief valve, is installed where it will warn the operator."

Contrary to the above, the core spray suction relief valves (1-E21-F032B and 1-E21-F032A) were removed and blank flanged in 1979 and 1985, respectively.

This is a Severity Level IV violation (Supplement I).

RESPONSE TO VIOLATION C I. Admission or Denial of che Alleged Violation CP&L admits to the removal of the auction relief valves and the installation of the blank flanges. However, it is believed no safety concern existed due to system design configuration and adequate precautions being implemented to negate overpressurization of the suction piping.

II. Raason for the Violation Engineering Evaluation Reports (EER) provide a means to deviate from certain requirements for reasonable periods of time. Engineering Evaluation Reports assessed and documented the safety hazards of operating the Core Spray S istem without the suction relief valves installed and authorized the ir temporary removal, f

k The 1979 EER written to justify initial removal of the 1-E21-F032B valve determined the safety hazard was insignificant if either the torus-i suction valve or the condensate storage tank suction valve was opened and required the torus suction valve to be open. The 1985 EER written to ,

justify continued removal of the 1-E21-F032B valve and to also justify

, removal of the 1-E21-F032A further determined the dischargs relief valve would provide adequate protection during a momentary pressure surge 1 during a momentary suction path transfer (and required caution tags for the valves). Even though the EERs differed slightly in content, the conclusions were correct and adequately justified in each case.

In addition, the miniflow valve (E21-F031) is open (except with flow-rates greater than 475 + 14 gpm) and provides a direct path to the torus which would prevent overpressurization'of this piping with both suction valves closed.

The intent of USAS B31.1 and 10CFR50.55a is to provide requirements for- '

safe operation. It is felt that the intent was satisfied in this case.

III. Co:rective Steps Which Have Been Taken Since the 1979 EER was issued, a change to the EER management directive  !

has been made such that formal action items are required whenever  !

temporary repairs are conducted to rectify a condition. Engineering and l QA (for Q-list items) approval is required for extension of these items.  :

Training has been conducted with appropriate Technical Support engineers i responsible for approving temporary repairs to the plant on the need for  ;

i timely restoration to full conformance.  !

A review of other temporary repair engineering evaluations and mechanical jumper logs was undertaken to ensure no instances of code deviations have ,

been approved as temporary repairs. No instances were found. In ,

3 addition, a review of the plant special procedures is ongoing.

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IV. Corrective Steps to be Taken and Date for Full Compliance I i

l The engineering evaluation procedure will be revisad to require plant i

, General Manager concurrence on all extensions of temporary repair t conditions in the plant. The planned completion date.is March 15, 1988.

The Core Spray System will be restored to design configuration upon

, receipt of the replacement valves. The planned completion date is July 15, 1988. l 1

The remaining special procedures will be reviewed for instances of design changes. The planned completion date is February 29, 1988.

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l VIOLATION D 10CFR50, Appendix B, Criterion XVIII, and the licensee's accepted Quality Assurance Program (FSAR, Section 17.2.18) collectively require that a s, stem of planned and periodic audits shall be carried out to verify compliance with all aspects of the Quality Assurance Program. The FSAR, Section 1.8, commits to Regulatory Guide 1.144, Auditing of Quality Assurance Programs for Nuclear Power Plants. This Regulatory Guide endorses ANSI /ASME N45.2.12 - 1977, Requirements for Auditing of Quality Assurar.co Programs for Nuclear Power Plants. Paragraph 4.3.2.3 of this standard states that selected elements of the Quality Assurance Program shall be audited to the depth necessary to determine whether or not they are being implemented effectively.

Contrary to the above, the corrective action system was not audited to the depth necessary to determine it was being implemented effectively in that corrective action audits conducted from 1985 to 1987 did not include the Engineering Work Request system.

This is a Severity Level IV violation (Supplement I).

RESPOSSE TO VIOLATION D I. Admission or Denial of the Alleged Violation CP&L acknowledges that the corrective action system (EWRs) was not auditeu to the depth necessary to determine if it was being implemented effectively.

II. Reason for the Violation The EWR program was not considered to be a corrective action program as defined by 10CFR50, Appendix B, by the CP&L Corporate Quality Assurance Services Section nor the oa-site QA/QC Unit; therefore, this program was not audited to ensure effective implementation. The EWR program was understood to be an engineering work tracking system designed to identify, prioritize, and schedule those tasks required by the Technical Support subunit personnel. Based on this understanding, the program was not audited as an Appendix B program.

III. Corrective Steps to be Taken and Daue for Full Compliance An independent assessment of the corrective action programs at the

] Brunswick Nuclear Project was initiated in September 1987. This assessment has been completed and the results are currently being reviewed by CP&L site, corporate, and Quality Assurance management. From this assessment and any other required reriews, Quality Assurance management will develop and implement audtt and surveillance coverage of appropriate corrective action programs. Problems identified during the ,

audit and surveillance activities will continue to be reported to management in accordance with quality assurance procedures. These audit l and surveillance activities will be implen'4ented by September 1,1988.

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r-EWR 84-809A Comments Several conversations have been held with both General Electric and Rosemount in an attempt to resolve this problem. Both stated that the best course of action was to modify the power supply system. Modification of the circuit board and the use of any existing trip unit designs was ruled out.

A review of the options available led to the decision to add a redundant power supply from the other battery. Plant modifications85-020 and 85-021 were, therefore, initiated.

Action Planned Plant Modification 85-021 has been approved and Plant Modification 85-020 is awaiting approval. A budget package will be submitted in 1988 for work in the 1989 and 1990 outages (Unit 2 and Unit 1 respectively). The procedure changes made to date will help to minimize the chances of this event occurring in the interim.

The inspection report states that the initial modification was unsatisfactory.

Without the knowledge of the deenergization/reenergization trip problem, CP&L had no reason to design for such an event. With the fact that General Electric initially told Rosemount that the system would always have power, Rosemount also had no reason to design for such an event. The modification was designed satisfactorily given the knowledge at the time.

The report also states that Engineering has determined that the modifications proposed will not resolve the deficiency. This statement is misleading. It is known that the modification will not resolve the root cause of the problem in that the circuit boards are not being modified. This, however, would have to be a vendor function and Rosemount does not wish to modify the boards. The cost involved in the redesign and requalification would be extensive and, in their minds, not justified. The modifications being prepared are in line with vendor recommendations and will reduce the probability of a loss of power to the units. Therefore, even though the modifications do not modify the boards, the addition of a redundant power source is in lina with the actions taken by other plants, the recommendations of the vendor and General Electric, and thus reduces the probability of the event.

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EWR 84-685 Comments The SBGT skid is fed from a 480-volt vital ac MCC with a 60-amp feeder breaker at the MCC. There is a 50-amp ITE circuit breaker at the skid. This 50-amp breaker feeds the SBGT fan motor, resistance heater, and control logic. At the time the EWR was initiated, the fan motor had a nameplate current of 19.2 amps.

The original fan motors have since been replaced (part of environmental qualification modifications) with motors that have a nameplate current of 17.7 amps.

The vendor (Farr Co.) was contacted about the methodology of the calculations that determined the 50-amp breskers were undersized. It was stated by Mr.

Tuul that his calculations were based on the following data with a derating factor of 1.15:

Item Method Result Motor (15 horsepower) Table 21.0 amps Heater (18 KW) Calculation 22.6 amps Transformer (500 VA) Table 1.1 amps Based on the above data that Mr. Tuul used in his calculations, the 50-amp breaker would be considered undersized.

If calculated data and nameplate ratings are used instead of tables, the  !

data (at 480 Vac) under existing conditions is:

Item Method Result .

Motor (15 horsepower) Nameplate 17.7 amps l Heater (18 KW) Calculation 22.6 amps Transformer (500 VA) Calculation 1.04 amps The Farr Co. uses 1.15 as the derating factor for breaker sizing. Most l breaker manufacturers use a derating factor of 1.25 and the National Electric Code requires a derating factor of 1.25 for breaker sizing. Using the 1.25 derating factor and an additional derating factor of 0.97 based on the gauge and length of the wire used (reference GET-2779G), the 50-amp breaker is adequate for the load. This data was discussed with the Farr Co.,

l and they concur that the 50-amp breaker is adequate for the load.

i If the 19.2-amp nameplate rating of the original motor is used in conjunction with the 1.15 derating f actor that the Farr Co. uses, the j 50-amp breaker was adequate with the original motor installed.

This EWR is being closed with an expected closure date of February 19, 1988.

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EWR 84-1024 Comments This EVR is currently open and active as a tracking document for the installation of the close latch antishock springs in the plant ITE SHK circuit breakers. The current plan is to close the EVR as a tracking document. The EWR will be closed because the inspection for and installation of the close latch antishock springs are part of the procedural instructions in Maintenance Procedure OPM-BKR001 (formerly MI-10-2H).

Current Plans for Resolution of the EWR The plans for completing the installation of the close latch antishock springs remain unchanged. The springs will be installed during routine breaker maintenance per OPM-BKR001.

To date, approximately 70 percent of the Class 1E SHK breakers and approximately 51 percent of the BOP 5HK breakers have had the springs installed. Current plans for the 1988 Units 1 and 2 outages are to complete 10 more Class 1E SHK breakers. After the Units 1 and 2 1988 outages, approximately 88 percent of the Class 1E SHK breakers will have the close latch antishock spring installed. The Class 1E breakers which will not have the close latch antishock spring installed after the outages are:

1. Control Rod Drive Pump 2A Feeder Breaker

2. Control Rod Drive Pump 2B Feeder Breaker

3. Nuclear Service Water Pump 1B Feeder Breaker

4. Residual Heat Removal Pump 2C Feeder Breaker

5. Tie breaker from Emergency Bus E3 to Emergency Bus El

6. Tie breaker from Emergency Bus E4 to Emergency Bus E3

7. Tie breaker from Emergency Bus E4 to Emergency Bus E2 The BSEP tie breakers are administrative 1y controlled. The normal condition of the tie breakers is racked out with the control power fuses pulled, therefore, after the 1988 outages, there should only be four active Class 1E breakers which have not had the close latch antishock spring installed.

Expected Completion Date The tentative completion date for the installation of all close latch anti-shock springs is after the 1990 Unit I refueling outage currently scheduled for November 1990. This completion date is based upon the fact that breaker maintenance is performed on a three-year frequency pending the availability of plant equipment.

As the inadvertent closing of SHK breakers due to the lack of a close latch ancishock spring has only been experienced twice industry wide since February 25, 1983, BSEP does not consider the problem a significant safety concern and does not believe that accelerated spring installation is necessary.

EWR 82-294 Comments This EVR addresses the fact that the fire protection tank low level alarm is set below the technical specification minimum level of 200,000 gallons.

The stated issue is that the tank water level is not, therefore, electrically supervised as required and that the plant did not perform an evaluation to justify continued system operability, submit a variance or technical specification change to the NRC, nor administratively control the fire tank level above the limit.

CP&L contends that electrical supervision is provided as required. There are three bases for our position. First, monitoring tank level for technical specification compliance is accomplished via visual monitoring by Operations personnel in accordance with and documented in the Auxiliary Operator's Daily Surveillance Report (DSR).

Second, the statement in the Appendix A Safety Evaluation Report (SER) that the level is electrically supervised is correct. The tank is equipped with two alarms, one set slightly below the technical specification limit (the low level alarm) and one set at approximately 35,000 gallons (the low, low level alarm). These alarms provide the operator with notification of diminishing water level in sufficient time to allow fire pump suction transfer to the alternative water supply.

These are the same conditions which existed at the time that the NRC fire protection review team was developing the SER.

Third, the fire protection SER was written by a multidisciplinary fire protection team which included at least two fire protection consultants.

These consultants were assigned to focus on fire protection systems and fire suppression capabilities among other topics. Given that these fire protection consultants found electrical supervision of fire tank level adequate, the meaning of "supervision" to fire protection personnel needs to be discussed.

In fire protection parlance, "supervision" is a term applied to monitoring a system and previding indications of a problem. For example, fire l detection systems are required to be supervised. This means generally that in the event of an open, a short, a significant ground, a switch out 1 I

of position, or a loss of normal power, an alarm is provided to notify the appropriate party so that the requisite action can be taken.

Considering this perspective and considering that a backup water supply was available, the basis for the SER finding that electrical supervision was provided is clear. The existing installation notifies the operator in sufficient time to shift to the backup supply prior to the loss of suction to the fire pumps.

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b Additionally, the issuance of the technical specification occurred  !

virtually simultaneously with the issuance of the SER. The authors of the SER were familiar with-and involved in the review and approval of the l technical specification for issuance. With this knowledge, the review teau found that electrical supervision was satisfactorily provided. This further indicates that the concept of "supervision" as discussed in the preceding paragraphs was that utilized by the authors of the SER in their finding.

Although CP&L considers that electrical supervision is provided, having the low level setpoint above the technical specification limit is also both prudent and desirable. Action to affect the setpoint change will be pursued. This action

is not considered to.be a safety concern and will be processed and scheduled as a routine item.

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Comments This EWR addresses the fact that some sprinkler systems can draw more water than can be supplied for a two-hour period. The two-hour period is a guideline t

, provided in Appendix A to BTP 9.5-1 and which the Appendix A SER considered ,

Brunswick to satisfy. The stated issue is that BSEP did not perform an evaluation to justify continued system operability, submit a variance or j technical specification change ta the NRC, or administrative 1y control the fire 1 tank level above the new 240,000 gallon figure. ,

The 240,000 gallon figure originated in a set of calculations performed to address the concern identified in the subject EWR, These calculations, while i demonstrating that the maximum amount of water that could be drawn by the worst case system actuation was 240,000 gallons, indicated that the minimum required  ;

water voluma to provide the system's design demand

• for two hours was less than l the 200,000 gallon technical specification limit. Thus, there was no basis for submitting a variance or technical specification ch nge. Nor was there a basis  ;

for controlling level above the 240,000 gallon figure. l The data used as the basis for the above statements is preliminary in nature i and will be formally reviewed and approved by CP&L by July 15, 1988, i *The design demand is the water pressure and flow condition necessary to 1 provide the design density over the area of application. The design density is taken from NFPA-13, Standard for the Installation of Sprinkler Systems, ,

Table 2-2.1(B), based on the hazard classification of the occupancy and the selected area of application. This approach is standard fire protection ,

practice and is consistent with the guidance in BTP 9.5.1, revision 2. /

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