ML20236S990
| ML20236S990 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 11/23/1987 |
| From: | Trevors G NEBRASKA PUBLIC POWER DISTRICT |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| 66723, NUDOCS 8711300222 | |
| Download: ML20236S990 (19) | |
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Novembcr 23, 1987 U.' S. Nuclear Regulatory Commission Attention: Document Control Desk i Washington, D.C. 20555 ]
i Gentlemen: !
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Reference:
- 1) Letter G. A. Trevors to NRC dated August 14, 1 1987, " Safety System functional Inspection at Cooper Nuclear Station"
- 2) Letter G. A. Trevors to NRC dated July 24, 1987, " Safety System Functional Inspection at Cooper Nuclear Station"
Subject:
Nebraska Public Power District's Response to Inspection Report No. 50-298/87-10 This letter provides the Nebraska Public Power District (NPPD) response to the subject inspection report. The report presented the findings of the Safety System Functional Inspection performed on the Cooper Nuclear Station emergency electrical system and its auxiliary support systems and identified a number of specific deficiencies as well as broader programmatic issues extending beyond the systems reviewed. Specific responses to the significant findings identified in Section 2 of the report are contained in the attachment. As indicated by these responses, NPPD had previously identified and implemented various programs to correct many of the specific deficiencies identified in the inspection report.
A summary of these general programmatic undertakings is provided below:
I. The report identified a number of deficiencies on certain plant drawings and raised a concern that other plant drawings may also be deficient. NPPD already had an extensive as-built verification program underway prior to the inspection. This program includes a systematic review, update and verification of selected drawings maintained in the Control Room. The District plans to expand the effort to include selected related drawings. This expansion will involve approximately 2,000 additional drawings to be verified.
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Page 2 November 23, 1987 II. The report identified a number of design control problems associated with modifications implemented at Cooper Nuclear Station. In an effort to improve the design control process the District had previously identified and corrected several weak areas in modification and design control. Engineering procedures have been upgraded to meet the applicable :
ANSI standards for the Equipment Specification Change ,
program. This effort has been independently verified l by the District's Quality Assurance Group. In addition, a major upgrade which was in progress at the time of the inspection, is continuing and will clarify the new design change process responsibilities. These revised procedures will incorporate guidance provided l by nuclear industrial groups. This effort is expected to be completed in May 1988. Design control responsibility has been assigned to the Nuclear Engineering and Construction Division at the corporate office.
III. The report identified several problems with design calculations and the design bases for the systems reviewed. The review team expressed concern that these problems may extend to other safety systems as well.
The District had previously identified weaknesses in these areas and had commenced efforts to determine their scope and extent. A configuration management department was created and an action plan to collete the design bases and determine the adequacy of the design calculations was formulated in February 1987. A vertical audit was initiated on the High Pressure Coolant Injection System as part of a pilot program to gain a better understanding of the scope, schedule and resource allotment required to gather the base data for the remaining safety systems. This pilot program, expected to be complete by May 1988, will identify the sources and availability of supporting documents and the depth of detail required to develop design bases darumentation reference. This documentation will taentify the original design requirements, assumptions made, and verify that modifications made to date are within the design parameters. Furthermore, it will establish an accurate and efficient data base to assist design engineers with their identification of the applicable system bases and calculations when planning future modifications or conducting new analyses. A schedule for the remaining essential systems based on the results of the pilot program will be generated by August 1988.
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Page 3 November 23, 1987 NPPD b'elieves that the -attached responses together with the foregoino constitute a positive and aggressive action plan for correcting the problems identified. Each of the Section 2 significant- findings are sequentially repeated in' the attachment followed by the District's response. The District reiterates that' we had previously identified some of the noted weaknesses and had instituted programs to correct them prior to the commencement of.this inspection. Nevertheless the District is y fully cognizant 'of. the concerns expressed by this Safety System Functional Inspection. and is committed to take the appropriate actions to resolve these concerns in a timely manner. In addition, this experience will be utilized in performing our own internal inspections and audits on remaining essential systems.
If you have any questions regarding this response, please contact me.
Sincerely, l 1
. . reVors Division Manager of Nuclear Support GAT /grs:rs18/l(20)
Attachment cc: U. S. Nuclear Regulatory Commission Regional Office, Region IV NRC Resident Inspector Cooper Nuclear Station m_----2-.------- . - - - . - _ . - - -
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Attachment to Nebraska Public Public Power District's Letter to the NRC dated November 23, 1987 This Attachment gives the responses outlined in Section 2 of the Inspection Report.
- 1. Functionality Concern 2.1.1(1)
The actual heat removal capabilities of the Service Water System were not measured. Instead, system flows were measured which did not account for heat exchanger fouling and the resultant loss of heat exchanger capabilities [3.1.1(1)] (50-298/87-10-01).
Response
While quantitative analyses have not been performed on various qsential !
heat exchangers to monitor the actual heat removal capabilities of the Service Water System, the District believes that adequate margins to design heat removal capabilities are being maintained, based on actual plant performance and operating data together with an established Preventive Maintenance program for inspection of the heat exchangers.
Nevertheless, a Special Test Procedure (STP) has been written and approved by the Station Operations Review Committee (SORC) to-quantitatively verify actual heat removal capabilities of the essential heat exchangers. This testing will be performed prior to startup from I
the Spring 1988 Refueling Outage. Based on the methods outlined in the STP, surveillance procedures will then be generated to periodically verify the actual heat removal capabilities of the essential Service Water System heat exchangers. These surveillance procedures will be developed within 90 days of startup from the Spring 1988 ' Refueling Outage.
- 2. Functionality Concern 2.1.1(2)
During system testing, the required flows were not achieved to each essential Service Water System load. Instead, pump flow was measured and compared to the total heat exchanger flow requirements. It appears that adequate testing has never been performed to ensure that adequata flow could be provided to the emergency diesel generators under the design basis scenario of one pump supplying all heat exchanger loads.
[3.2.1(2)] (50-298/87-10-02)
Response
The SSFI inspection team's concern was that the Service Water System had apparently never been fully demonstrated to be capable of providing the flows required by the USAR to each essential 19ad during the design basis accident with one pump supplying all loads. Consequently, the District, with input from General Electric, re-evaluated the existing USAR flow requirements. This re-evaluation determine 6 that the Service Water System, as designed and maintained, will achieve its design basis function following a LOOP /LOCA with one pump operating. The District previously provided a revised " Service Water System Design Basis Docume n t , see Reference 2, which identified the new minimum flow requirements.
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a The District is. currently preparing changes to the USAR to reflect the ]
revised minimum service water flow requirements as described above. I Additionally, the appropriate operating procedure now specifies the required operator manipulations of the Service Water (SW) System after the occurrence of a LOCA with concurrent loss of off-site power. These changes will ensure that Service Water System flows are properly distributed to achieve their required cooling functia.a.
A Special Test Procedure (STP) has also been written and SORC approved to verify that'the Service Water System is capable of providing the revised post-LOCA flow rates. This STP will be performed during any ut planned outage of five days or greater duration, or during the Spring 1988 Refueling Outage, whichever occurs first. Based on the methods outlined in the STP, surveillance procedures will be generated following performance of the STP to periodically verify the Service Water System is capable of providing the required post-LOCA flows. These surveillance procedures will be developed within 90 days of the performance of the STP,
- 3. Functionality Concern 2.1.1(3)
Inadequate operating guidance and training existed for casualty responses such as service water system flow balancing, operation with a loss of non-essential air or manual isolation of non-essential loads. Incorrect operator response to the design basis scenario could result in inadequate flows to essential loads, a pump runout condition for the one remaining pump, and loss of 'all Service Water System Cooling (3.3.1) l (50-298/87-10-01).
Response
As identified in previous correspondence, immediate corrective action taken regarding this item consisted of revising CNS Abnormal Procedure (A.P.) 2.4.8.3.1, " Loss of Service Water Pumps", to specify operator ;
actions associated with the Service Water System after a Loss of Coolant Accident, concurrent with a loss of off-site power. This revised procedure was then routed to all licensed operators for review. Formal training on A.P. 2.4.0.3.1 has been included in the current licensed operator Requalificatian Training Cycle which will be concluded on January 8, 1988.
Subsequently, a more e:ttensive review of A.P. 2.4. 8. 3.1 was conducted which focused on 1) the overall impact of the design basis scenario on the Service Water System, and 2) the required operator response to ensure the functional continuity of the system. This second review resulted in a much more comprehensive revision to this procedure, which is now entitled, " Service Water System Casualties". This new revision, which is currently undergoing internal review, provides additional, detailed directions to the operators to address the pertinent concerns raised in l Inspection Report 87-10-03. Upon approval, this procedure will also be routed to all licensed operators and included in the previously mentioned licensed operator Requalification Training Cycle.
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The details associated with this statement of concern also questioned the adequacy of communication between the Control Room and the Emergency Diesel Generator Rooms. An engineering evaluation has been conducted to l
[ address this concern. As a result, standard Gaitronics headsets will be l provided which offer background sound attenuation similar to that l achieved by normal hearing protection headsets worn by operators in high !
l noise level areas. The installation of these new headsets will greatly improve the communication capability in the Diesel Generator Rooms,' as well as provide for additional operator mobility. The modification required to provide these headsets is scheduled for completion during the Spring 1988 Refueling Outage.
- 4. Functionality Concern 2.1.1(4)
Auxiliary systems to prevent fouling of the intake structure were not designed to ensure adequate post-accident cooling. Fouling of the traveling screens could cause clogging of the pump suction and result in a loss of service water system cooling. [3.1.2(2)) (50-298/87-10-04)
Response
Calculation, NEDC 87-056, submitted during the inspection shows that even a 95% blockage of the traveling screens would not cause a collapse of the screen when encountering the flowrates associated with the operation of two service water pumps in the worst case condition. Also, Dr. John F. Kennedy, Director of the Iowa Institute of Hydraulic Research, and l Dr. Tatuaki Nakato, Scientist with the Iowa Institute of. Hydraulic Research, consider that the debris in the river is not of the type to cause even 90% blockage of screens under worst-case conditions. The worst case is considered to be maximum run off in the spring. A 90%
blockage of the E bay traveling screens is not considered to be a credible event. Dr. Kennedy and Dr. Nakato performed research on the hydraulic flow paths with regard to the Intake Structure and designed the weir wall for CNS. They have been involved with designs and studies associated with the Missouri River for a number of years.
Ice would not cause complete or even substantial blockage of the screens or racks due to its buoyancy and the geometry of the intake structure.
Any ice large enough to damage the traveling screens would have to be l drawn under the submerged intake structure wall, a difficult concept to i accept due to the buoyancy of the ice. If the river level were low enough to allow the ice to float under the intake structure wall, the size of ice would be limited by the top of the weir wall being at an elevation higher than the bottom of the intake structure wall. In this case, the weir wall would also prevent ice, in the main stream of the river, from entering the intake structure. Any ice which did manage to get under the intake structure wall would be small and would be stopped by the trash grid or the trash racks. The depth of the ice pileup on the trash racks would be limited by the buoyancy of the ice together with the low flowrate associated with the operation of two service water pumps, and thus would not be able to cause any significant blockage or damage to the trash racks or traveling screens. Hence, only small particles of ice would be present to cause a hazard to the traveling screens.
It is considered that a failure of the Screenwash and Sparger System would not result in a blocked service water pump suction. During initial plant start-up, pre-operational testing was performed on the service water pump bay (E bay) to assess the quantity and location of silt build-up in the bay. The test showed the highest level of silt build-up occurred immediately in front of the trash rack. The silt level then dropped in. the narrow part of the cell, raised again just downstream of the traveling screen where.the cell started to widen out, and then diminished to less than 2-3 inches in the remainder of the cell. The test results showed that the build-up is not rapid. District personnel would have at least seven days to either get the sparging system back in service, open the emergency crosstie from the Service Water System to the E bay spargers (CNS Station Operating Procedure 2.2.3) or manually open the sluice gate between the D and E bays to supply water from the D bay.
An earthquake would not alter the silt loading in the river, since the river is always considered to be in equilibrium so far as silt loading is concerned. A statement will be included in the CNS procedures to pru.npt management and operators to open the emergency cross-tie from the Service Water System to the E bay spargers in the event of a loss of power to the screenwash and.sparger systems.
In the light of the above comments, the District considers that the concern stated in the Inspection Report about the possible blockage of !
the service water pump suction is resolved. l
- 5. Functionality Concern 2.1.2(1)
The start-up and emergency transformers may not be properly sized to provide adequate voltage to start all the emergency core cooling system loads as designed [3.1.l(1)] (50-298/87-10-05).
Response
- 1. The AC Voltage Drop Analyses have been performed by Burns and Roe and NPPD personnel. The analysis covering the segment of the distribution system from the off-site power lines to the MCC terminals on the 480 VAC system has been performed using Burns and Roe's Computer Program ELO 110. NPPD personnel are performing the voltage drop analysis for the remainder of the AC Distribution j System at CNS.
- a. The Burns and Roe study indicates the following:
(1) For the 161 kV line supplied via the 345/161 kV transformer at CNS, the electrical distribution system powered via the start-up transformer can withstand a
" block" start of Emergency Core Cooling System (ECCS) equipment with the normal auxiliaries (essential and non-essential) remaining energized for all possible 161 kV bus voltages.
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(2) For the' 69 kV line - supplied from the Omaha Public Power District _(OPPD) system, the emergency off-site supply, the j electrical distribution system powered via the emergency J station transformer can withstand a sequential starting of ECCS equipment as well as a phased start of two service q f
water pumps with all non-essential equipment de-energized. l l
The above cases meet the design requirements of the off-site power supply criteria for CNS; two independent power supplies in addition to the emergency diesel generators.
Calculations performed by NPPD demonstrate that the ECCS motors would start and accelerate to speed without actuating the undervoltage relays. This calculation assumed maximum potential transformer (PT) and relay errors and a sufficient margin was demonstrated to exist between the calculated low voltage and relay pick-up point.
- b. Although the existing configuration of the Emergency and Start-up Transformers is adequate to provide preferred power in accordance with GDC 17, NPPD',' performing the following work to increase the safety margin and improve the voltage control for the Auxiliary Power Distribution System.
(1) A modification of the start logic of the ECCS motors will be performed during the 1988 outage to sequential start the ECCS motors onto the start-up transformer upon a LOCA and no loss of off-site power. This will '.ncrease the safety margin and give better voltage control on the 4160/480 system while the motors are starting.
(2) NPPD has performed a voltage drop study on the 480 VAC and 120 VAC systems in NPPD Calculation 87-132. This calculation analyzed the Division Il essential loads and certain Division I loads. Division II is considered to be the worst case with longer cable runs. The circuits analyzed to date are adequate to pe-form their safety related functions under postulated v>1tage conditions. In addition, voltage and current measurements were taken to establish the margin between the calculation and the field condition for the 120 VAC component analyzed. The readings taken indicated there was a considerable margin of conservatism in the calculation. Analysis will continue on the remaining circuits and will be completed prior to the plant start-up after the 1988 outage.
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(3) A Transformer Tap Change and Load Study is being performed by NPPD to determine the optimum tap settings for the various transformers in the supply and distribution systems. The existing system configuration is adequate per the Burns and Roe studies; however, further studies are being instigated to attain further improvement in the voltage control. The tolerance of grid voltage control could be relaxed by minor tap changes at each voltage level. Any tap changes emanating from the studies will be implemented during the 1988 outage.
(4) Since the 69 kV emergency line is now part of the Mid-Continent Area Power Pool (MAPP) system, it is maintained and controlled by MAPP regulations which specifies system voltages between 90 and 100 percent, assuming two separate contingencies. Preliminary discussions with OPPD indicate that no modifications are expected, between the present time and the 1988 outage, that would affect the voltage control on the 69 kV emergency line. Further discussions will be held with OPPD on the contract associated with the 69 kV line.
Studies will also be undertaken on completion of contractual negotiations to assess the feasibility of installing a low voltage annunciator circuit for the 69 kV line.
NPPD considers that when all the actions above are completed the Commission's concerns, described in Section 2.1.2.(1) will have been resolved.
- 6. Functionality Concern 2.1.2(2)
The station battery may not be properly sized to provide adequate voltage to the closing coils for the output breakers of the emergency diesel generators and the emergency transformer. [3.1.l(2)) (50-298/87-10-06)
Response
The DC Voltage Drop Study and the DC Load Study have been completed. The studies were performed by CYGNA Energy Services. NPPD has completed a review of both studies. The review of the results of the studies confirm the statements made in NPPD's letter dated August 14, 1987 (Reference 1).
Further, to the statements made with regard to the closing coils for the 4160 kV Breakers 1FS, 1GS, 1FE, and 1GE, the vendor has recently supplied documentation which corroborates the statements made in the referenced correspondence.
In addition, the voltage drop to the DC motor-operated valves (MOVs) were also checked and the results show that adequate voltage exists at each MOV to enable it to perform its cafety function. The voltage study verified that the MOVs will develop sufficient torque to open or close the valve under worst case conditions. Assumptions cuch as full voltage locked rotor current (which provides a conservative voltage drop), margin between actual motor locked rotar torque and nameplate torque, and a battery at 70 F with a capacity at least 3 percent below existing capacity were included in the calculation.
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- 7. ~ Functionality Concern 2.1.2(3) J i
l The heating, ventilation,.and air conditioning system may'not be able to h'
. provide adequate cooling to the AC switchgear, DC switchgear, and station
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battery rooms. Excessive temperatures could prevent proper operation'of cs
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essential electrical equipment and systems in the rooms. '[3.1.2(1))-
(50-298/87-10-97)- i l
Response
In response to the concerns raised during the inspection, Nebraska Public j Power District initiated an evaluation of the Cooper Nuclear Station's ventilation systems located in the Control Building together with the essential electrical equipment cooled by this system.' The results of
'this evaluation were given in Reference 1. . The evaluation was performed for Battery Rooms lA and 1B, DC Switchgear Rooms 1A and 1B, and Critical ,
Switchgear Rooms IF and 1G, to develop the bounding transient room '
temperature profiles, resulting from postulated abnormal events which include loss-of ventilation or loss of off-site power. In addition, an ;
assessment was performed of the essential electrical equipment located in these rooms and determined that the equipment was capable of performing
'its intended safety functions when subjected to the bounding event temperature profilen. As a result of.the study, the District developed plant operating procedures which, when invoked, would put in place dedicated auxiliary ventilation equipment capable of being powered from an essential bus, to assure that the _ battery and switchgear room temperatures remain at acceptable levels during the summer months.
In response to ' the concern about minimum ambient temperatures ,, NPPD initiated the ' following action to ensure the battery. room temperatures ' l will not fall below 70*F during a loss of heating from the Control Building air units. Calculations have . been performed to size the
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equipment and procedures are in the course of review to implement an auxiliary heating system-for the Control Building Battery Rooms 1A and 1B should the heating systems associated with the air units fail. Power will be supplied via a critical MCC. The equipment will be stored in close proximity to the battery rooms for expedient implementation. These actions will be completed before December 4, 1987. The shortcomings in the Station's HVAC Systems in the Control Building were recognized by the District before the inspection. The long-term study into the HVAC Systems at CNS, implemented in May 1987, has been expanded to encompass the concerns voiced by the inspection teams. The District has appointed a senior project manager to oversee and direct the task. Once the ,
recommendations from the study have been assessed, modifications will be j performed as appropriate. The study is expected to be completed by May 1988.
The diesel generator room ventilation is annunciated for such items as failure to start and high DP across the filters. The District will investigate additional annunciation of the diesel generator HVAC units and will incorporate any recommendations into the District's overall annunciator upgrade program currently under development as part of the Detailed Control Room Design Review project.
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In regard to the air supplies for the components in the ventilation system for the diesel generator control panels, a design change has been written and is in the process of approval to upgrade the control system to add an essential air supply onto the existing pneumatic control system for HVAC Units 1C and 1D. The modifications will ensure that on an auto emergency start of the diesel generator, the controls would be automatically switched to the essential air supply which controls the subject valves and damper. This work will be completed by December 18, 1987.
In response to the concerns about direct indication to the Control Room for failure of the battery room exhaust fans, the District draws attention to the fact that the fans have two flow supervision sencors.
The 125 VDC and 250 VDC battery room exhaust fans are supervised by flow sensors which are annunciated to the Control Room at Vertical Board R-1, Annunciator Panels 3-1 and 3-2. These flow sensors will alarm on low flow from either Room 1A or 1B. This direct indication of flow is in addition to the exhaust fan motor interlocks which alarm on motor failure and on failure of the standby fan to start.
- 8. Functionality Concern 2.1.2(4)
There was no analysis to demonstrate that the 120 VAC electrical system would function as intended during accident conditions [3.1.1(1)]
(50-298/87-10-08).
Response
See answer provided to Functionality Concern 2.1.2(1) given in Item 5 above.
- 9. Functionality Concern 2.1.3 The 4160 VAC electrical system did not appear to be adequately designed to accommodate emergency diesel generator testing. The 4160 VAC switchgear appeared to be undersized for short circuit conditions that could occur during test configurations. The circuit breaker overload settings to protect the emergency diesel generator during testing were also set above the stall rating of the diesel generator [3.1.1(3) and 3.1.1(4)] (50-298/87-10-09).
Response
The 4160 VAC Momentary Fault Study commissioned by NPPD during the audit time period indicates that while performing the monthly surveillance test on the emergency diesel generator, a phase-to-phase momentary fault, if it occurred, could be as high as 63,800 A. The switchgear is certified to withstand 60,000 A and the breakers 58,000 A faulted current. The District gave an interim response in paragraph A2 of Reference 1.
Additional studies and discussions are currently being pursued with vendors to clarify the District's response given in Reference 1. These studies will be completed by December 11, 1987.
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In regard to.the overcurrent relays, it is intended to replace the relays I for, Diesel _ Generator Overload ~ Protection with directional overpower L- relays during the.1988 outage'. _The_ relays, 51/IFE and 51/ IGE overcurrent relays, will be : replaced with directional- overpower solid-state ' relays set: at ' 110' percent of- the - electrical rating of the diesel generator. ]
This setting' allows the diesel generator combination to provide an output o up to 110 percent of its rating and still allow for instantaneous
! tripping before the diesel reaches stall speed at 128 percent of its rating.
- 10. Programmatic Concern 2.2.1 The' team identified'several instances where events were not reported to-the NRC and inadequate corrective actions .were' taken for significant deficiencies with essential equipment [3.5.1 and 3.5.2] l (50-298/87-10-10).
Response
This. concern identified' apparent deficiencies with the implementation of the nonconformance .and corrective action program covered under Administrative Procedure 0.5.1, "Nonconformance And Corrective Action".
-In reviewing the concerns noted in Sections 3.5.1, 3.5.2,' and 3.5.3 of the SSFI Report 50-298/87-10 enclosure, the deficiencies identified can be grouped into three general areas of concern: deportability, operability, and timeliness. The balance of this response will be directed at addressing actions taken or planned to rectify. deficiencies in these three areas.
- 1. . Deportability: The following Nonconformance Reports ~ (NCRs) were identified as having deportability concerns:
NCR 4759 - 10/23/85 NCR 5227 - 12/26/85 NCR 5056 - 8/22/86 NCR 4600 - 9/22/86 NCR 6383 - 10/31/86 ;
NCR 6392 - 11/28/86 On October 6, 1986, organizational changes were directed by senior management that created a Technical Staff Support Group (TSSG) at ;
CNS. As of March 9, 1987, this support group was fully staffed, i consisting of the Technical Staff Supervisor (new position), two senior level engineers (new positions), and the Regulatory Compliance Specialist (existing position). A primary function of this group is to evaluate and recommend event deportability.
Additionally, this group was delegated the responsibility to administer the Nonconformance Program and draft all Licensee Event Reports for the station. These functions have allowed this group to obtain the experience required to provide consistent and accurate deportability evaluations based on the guidance provided in the CFR and in NUREG 1022.
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In addition to the formation of the TSSG to provide an upgraded level of expertise and to focus this expertise on deportability, the District has committed in response to Inspection Report 50-298/87-18 to assess the availability of training materials and/or services on 10CFR50.72 and 10CFR50.73 deportability requirements. If suitable materials and/or services can be located, the upgrading of the knowledge level of required personnel in this area will be completed by February 16, 1988, as previously committed.
Independent of the District's commitment to pursue deportability training, one of~the two senior level engineers in the TSSG is currently in Senior Reactor Operator (SRO) Qualification Training.
This dedication of manpower and resources will further enhance the knowledge level and expertise of the TSSC in the area of technical and operational plant concerns.
The above stated actions implemented / initiated by the District will assure consistent and accurate deportability evaluations based on i the guidance provided in the CFR and in NUREG 1022. In order to assure that all nonconformances will be evaluated for deportability concerns, Administrative Procedure 0.5.1, "Nonconformance And Corrective Action", will be revised prior to January 1, 1988, to require a documented deportability review for every NCR. ,
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- 2. Operability: The following NCRs were identified as having operability concerns:
NCR 4759 - 10/23/85 NCR 5056 - 8/22/86 NCR 4600 - 9/22/86 On May 14, 1987, Administrative Procedure 0.27. " Component Operability", was implemented to provide station personnel with guidance on operability evaluations. This procedure utilizes a ,
step-by-step evaluation process to assist in the determination of component / system operability. The procedure requires that if found ,
I necessary, an engineering evaluation justifying continued plant operation will be performed and submitted for SORC review and ,
approval. This procedure will continue to be used and refined as an d ongoing commitment to maintain operational safety.
- 3. Timeliness: The noted timeliness concerns included initial NCR disposition, as well as implementation of corrective actions. The following NCRs were identified as having timeliness concerns:
NCR 4759 - 10/23/85 NCR 5056 - 8/22/86 NCR 4600 - 9/22/86 NCR 5034 - 4/8/86 NCR 3543 - 1/24/85 NCR 3264 - 2/14/85 NCR 3288 - 3/21/85 NCR 2988 - 6/14/84 NCR 6379 - 10/27/86
On February 20, 1987, Administrative Procedure 0.5.1, "Nonconforn.ance and Corrective Action", was revised to require the assignment of due dates for the initial disposition of NCRs.
Subsequently, the commitment and action item trac' king system was modified . to facilitate the tracking of NCRs during the various phases of disposition. All NCRs generated since. the revision of Administrative Procedure 0.5.1, and all prior NCRs still awaiting initial disposition, were input'to the ' tracking system. These actions, in conjunction with increased management attention, have greatly reduced the time to initially disposition NCRs and the number of NCRs awaiting initial disposition.
The February 20, 1987 revision to Administrative Procedure 0.5.1 also addressed the timeliness of corrective action implementation, stating:
" Corrective actions identified during the NCR disposition should be, as a minimum, initiated (e.g., Nuclear Engineering Department Work Request generated, Procedure Change in routing, CNS Work Item initiated, etc.) prior to cubmittal for close-out review."
Although this statement addresses the timely igir.iation of corrective actions, it does not assure the timely completion of !
corrective actions that could be critical to continued plant operation. Administrative Procedure 0.27, " Component Operability",
provides a methodology for evaluating failed critical components and requires that a justification for continued plant operation be provided, when required. Therefore, Administrative Procedure 0.5.1 will be revised such that nonconformances which require that corrective actions be completed at some time in the future (due to plant conditions required, availability of parts, etc.) will not be submitted for close-out review until the specific evaluation justifying continued plant operation (per Administrative Procedure 0.17) is re-validated. This revision will be implemented prior to January 1, 1988.
! In summary, the District had recognized and addressed the deficiencies noted in the concern prior to the Safety System Functional Inspection. As a result of the Safety System Functional Inspection, however, additional actions have been initiated to further enhance the nonconformance and corrective action program.
- 11. Programmatic Concern 2.2.2 Examples of deficiencies were noted in the design analyses performed by the licensee, including the use of incorrect calculation methods, assumptions, and design inputs. Additionally, drawings and design bases l
were not always updated to reflect station modifications. [3.1.3) 1- (50-298/87-10-11) 1
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i Following the 1986 Outage the District performed a review of the design change process. As a result, steps were initiated early this year and j are still underwsy, that when completed will correct. the programmatic l concerns expressed by the inspection team. Responsibility for the design of the plant was recently transferred to one organization within the District, the Nuclear Engineering and Construction Division. As a part ;
of this reassignment of responsibility, the design change control and - J configuration management process is being reviewed ' and upgraded. The concerns expressed by the evaluation team in this area are being evaluated closely to assure the new programs will resolve these concerns and strengthen the design change process for the enhancement of safety. i These programs to review and revise procedures will be completed by May 1988.
NPPD acknowledged the discrepancies in some of the calculations reviewed by the inspection team at the time of the inspection. In each case the ;
calculation has either been amended or is in the course of review. As stated in previous paragraphs the ' revised engineering procedures being produced in association with the changes in responsibility for design control will ensure a ' complete and thorough review of all calculations will be undertaken by the District personnel or consultants i before a modification is considered complete. In all instances the revised calculations relating Lo deficient calculations identified in the report, as well as any associated modifications, will be completed before j plant startup following the 1988 Outage.
In regard to concerns expressed over Special Test Procedures, it should be noted that STP 85-007 which was cause for concern was implemented under Revision 0 of "Special Test Procedure /Special Procedure", CNS Engineering Procedure 3.5. The current revision of CNS Engineering Procedure 3.5 requires the design engineer to perform a 10CFR50.59 Deportability Analysis and Safety Evaluation. In addition, Revision 2 now also reduces the possibility that temporary change on an essential system or component would remain in effect beyond the duration of the test without being properly evaluated, documented, and approved. Again the District had identified the issue, described above, in March 1987 and already had taken significant steps to resolve the issue. The revised procedures, now in place, call for an evaluation of the system .
functionality and related problems to detect potential modifications that !
may degrade plant safety.
In regard to the concerns over Equipment Specification Changes (ESCs),
the recent revision to the Design Change Procedure, Rev, 5, upgraded the ESC requirements. The requirements in those areas of concern are now the same as that required for a formal design change (DC) and satisfy the requirements of ANSI N45.2.ll-1974 (Ref.Section III.A.3 of CNS '
Engineering Procedure 3.4 " Station Design Change, Rev. 5). Discussions on the control and direction given in procedures with regard to ESCs had taken place following the 1986 outage. CNS Engineering Procedure 3.4 was being rewritten at the time of the inspection to give more direction and guidance with regard to the content and format of ESCs. The concerns voiced by the Inspection Team are considered to have been resolved.
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In regards to the concerns expressed over design basis documentation, update, and control, the District had recognized the potential issue in 1986 and authorized the formation of a new department, the Nuclear Configuration Management Department. This group was formed in February 1987 and is currently implementing policies which will establish a " Configuration Control and Data Base" for CNS in a form readily accessible and usable by the engineers in the District. The majority of the problems identified by the evaluation team in the design analysis and ,
I design inputs area are due to the current difficulty of the engineer to readily locate the design basis documentation and related assumptions, !
This also led to the inspection team's concern that the design base may not always be updated to reflect changes. The retrieval and re-establishment of the design basis documentation in a form readily accessible and usable by the engineer, along with programmatic changes being made in the design change process, will mitigate the current concern with design assumptions, design inputs, and maintaining an updated design basis.
The current plan is for a pilot program to be imp 3 emented for an essential NSSS System. The system selected is HPCI. It is intended to have the study complete by May 31, 1988. A schedule covering all the essential systems will then be produced based upon the experiences encountered during the production of the HPCI documents. The age of the design of CNS makes a pilot study imperative so that realistic estimates in terms of n.oney, man power and schedules can be formulated and approved. As part of the production of the design basis documentation, a review and update of documentation will be performed to account for any possible changes resulting from design changes during 14 years of operation. l The District, acting on the results of SSFIs conducted at other nuclear facilities instigated a program of vertical audits prior to the inspection by the Commission. One audit on the Core _ Spray was completed.
A second, on the HPCI System, was halted once the Commission's SSFI started due to manpower commitments associated with the inspection. The program of internal audits together with the full assessment of the results of these audits has been delayed until the responses to the Commission's Inspection, 87-10, have been clarified, submitted, and accepted by the Commission and any subsequent changes completed.
In regard tc drawing discrepancies, in 1986 the District identified a number of minor discrepancies in the drawings held in the Control Room.
A pilot project was implemented to determine the extent of the problems with the Control Room drawings. Following the verification of the pilot program after the 1986 outage, the project was expanded to encompass all the essential Control Room drawings. Steps are alao being taken to put in place a program to cross reference the drawings to assist the design engineer in identifying drawings when implementing the design change procedure. A large effort has been underway since February 1987 to verify these essential drawings. The overall effort should provide a more accurate method for an engineer to identify what drawings are affected when he is working on an essential design change.
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I A root cause evaluation of drawing discrepancies determined that: 1) it is probably due to the accuracy to which the original as-built effort was performed by the A-E when 'CNS was built and 2) the engineer not identifying all the drawings that are affected by a design change.
Although the original as-built errors have not been found to be of safety .
significance, they are more numerous than desirable. The second source l of drawing errors is much less in number. No error has been found where !
functional operability of a system could not be justified.
In' addition to.the above, it should be noted that in 1984 the District identified a possible area of concern in regard to ' the authenticity of conduit and cable tray hanger drawings. The decision was made to embark upon an as-built verification of the conduit and cable tray hangers.
Since 1984, two additional as-built verification programs have been started; verification of wall and floor penetrations and verification of the essential drawings held in the Control Room at CNS. The drawings listed in Section 3.1.3(5) of the report have been included in the programs described above. The District is also considering a further expansion of.the drawing verification program to encompass an additional 2,000 drawings related to essential drawings for CNS.
l In conclusion, the programs described in preceeding paragraphs indicate i the District's determination to improve its overall performance and quality of work in regard to the areas of configuration, design, and-modification control. In addition, the District wishes again . to emphasize that in all these instances, the deficiency was recognized, the commitment made, an overall, and in many cases a detailed plan and ,
schedule were in place prior to the Commission's SSFI. The District committed considerable resources in terms of money, manpower, and
, material to resolve the programmatic issues raised in the inspection report prior to the audit. NPPD considers the overall action plan will further improve the District's aim of operating CNS to its full capacity in terms of output and operating excellence.
- 12. Programmatic Concern 2.2.3 Instances were identified where inadequate post-maintenance testing occurred af ter maintenance on essential systems. In one case, an incorrectly sized residual heat removal pump
- gland seal water pump remained installed for ten months because inadequate post-maintenance was conducted after pump installation.
Note *: This pump is correctly titled RHR Service Water Booster Pump. J
Response
In order to address >:h e SSFI Inspection Team's concerns incregard to post-maintenance testing, Maintenance Procedure 7.0.1, " Work Item Tracking-Corrective Maintenance", has been revised to explicitly define responsibility for post-maintenance testing. In addition to clarifying responsibility, preassigned post-maintenance testing on Ess'ential or EQ j class components will now also be reviewed by the systea engineer to I verify adequacy prior to starting work. Additionally, he will be responsible for reviewing acceptability of the post-maintenance testing l
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results to verify adequacy of the original test requirements for the actual work performed. This review will be performed and documented '
prior to closure of all Maintenance Work Requests requiring post-maintenance testing.
To ensure proper coordination and close-out of Maintenance Work Requests and Design Changes, the size of the CNS Maintenance Planning staff has been increased from three to six permanent NPPD personnel. The noted organizational change was implemented just prior to the Safety System Functional Inspoction. The Maintenance Planning staff reorganization included a Maintenance Plannirg Supervisor, one additional Maintenance Planner / Scheduler, and a Data Clerk. One of the primary purposes of the Maintenance Planning staff upgrade was to ensure that the required reviews and signatures are completed on all Maintenance Work Requests and Design Changes prior to their acceptance and closure.
In addition, a Maintenance Self-Assessment is in progress at Cooper Nuclear Station. This self-assessment is the result of an INPO initiative to accelerate maintenance performance improvement in the nuclear power industry. Ten selected chapters of INPO Guideline 85-036, which includes post-maintenance testing, are being self-assessed against the requirement of this guide by a team of District personnel. Each team member was selected based on expertise and independence to ensure that the Maintenance Self-Assessment resulted in the District's overall objective of Maintenance excellence. A consultant with many years of direct maintenance management experience will also be used to provide independent review of the Maintenance Self-Assessment program implementation and results.
The goal for completion of the Maintenance Self-Assessment is December 31, 1987. Presently, the proj ect is on schedule. All recommendations resulting from the Maintenance Self-Assessment in regard to post-maintenance testing, as a result of the SSF1 concerns, will be given first priority for implementation.
To ensure a proper understanding of the Post-Maintenance Testing requirements a training session by 1NPO has been arranged for pertinent station personnel and will be completed by December 31, 1987. Similar training sessions are being evaluated for other Nuclear Power Group personnel.
- 13. Programmatic Concern 2.2.4 The trending program for the Service Water System inservice test data appeared inadequate. The team identified instances where service water pumps were operating in the alert range without the increased monitoring or corrective actions being accomplished as required [3.2.l(5)]
(50-298/87-10-13).
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Response
The District agrees that the existing procedural controls for trending of' plant IST data are inadequate. As such, the District will revise Engineering Procedures 3.9, " Inservice Testing of Pumps General Procedure", and 3.10, " Inservice Testing of Valves General Procedure", to
' establish specific trending guidance, where necessary, As a minimum, these ' revisions will require graphic superimposition of normal, alert, and required action ranges (where applicable) for inservice test quantities to further enhance early detection of equipment degradation and promote timely preventive maintenance. These procedure revisions will be completed by January, 1988.
De;viations noted by the inspection tean regarding inservice testing of service water pumps have been corrected by revising Procedure 6.3.18,3,
" Service Water Pump Surveillance". Furthermore, a review of all Inservice Test procedures for pumps will be conducted to verify compliance with ASME Code Section XI to fully address this programmatic concern. This review and any necessary procedure changes will be completed by January, 1988.
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