IR 05000267/1988200
| ML20235X579 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 03/07/1989 |
| From: | Callan L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | Robert Williams PUBLIC SERVICE CO. OF COLORADO |
| References | |
| NUDOCS 8903130533 | |
| Download: ML20235X579 (2) | |
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In Reply Re'fer To:
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Docket:
50-267/88-200
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Public Service Company of Colorado ATTN:
Robert 0. Williams, Jr., Senior Vice President, Nuclear Operations P.O. Box Ba0 Denver, Colorado 80201-0840 Gentlemen:
Thank you for your letter of January 27, 1989, in response to our letter and Notice of Violation dated December 7, 1988. We have reviewed your reply and find it responsive to the concerns raised in our Notice of Violation.
Your response regarding the unresolved items and enhancement of communication between management and field personnel appears to be acceptable. We will review the implementation of your corrective actions during a future inspection to determine that full compliance has been achieved and will be maintained.
Sincerely,
b L. J. Callan, Director Division of Reactor Projects cc:
Fort St. Vrain Nuclear Station ATTN:
C. Fuller, Manager, Nuclear Production Division 16805 WCR 19)
Platteville, Colorado 80651 Fort St. Vrain Nuclear Station ATTN:
P. Tomlinson, Manager Quality Assurance Division 16805 WCR 191 Platteville, Colorado 80651 Colorado Public Utilities Commission ATTN: Ralph Teague, P.E.
l 1580 Logan Street OL1
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Denver, Colorado 80203 l
Colorado Radiation Control Program Director
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D:DRP RIV:0Pc C:0PS C:PSB DRHuntM/lb JEGag iYrdo TFWestermanhLMi dan LJCallan
$/W/89 4/y/89 3 /9-/89 g/h/89 J /7 /89 I,l 8903130533 890307 ADOCK0500qg7 (
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Public Service Company of Colorado-2-bec to DMB (IE01)
bec distrib. by RIV:
RRI R..D. Martin, RA.
SectionChief(DRP/B)
Project Engineer, DRP/B RPB-DRSS-Lisa Shea, RM/ALF MIS System.
RSTS Operator K. Heitner,'NRR Project Manager (MS:
13-D-18)
RIV File DRS F
DRP J. E. Gagliardo W. C. Seidle D. R. Hunter G. Sanborn
.T. F. Westerman J. E. Cummins, NRR_
J. Calvo,'NRR'
G. Holahan, NRR C. Haughney, NRR B. Grimes,_NRR l
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a Company of Colorado P.O. Box 840 Denver, CO 80201 0840 January 27, 1989 R.O. WILLIAMS. JR.
Fort St. Vrain SENioA viCE SmES cEV
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Unit No. I P-89024
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U. S. Nuclear Regulatory Commission ATTN:
Document Control Desk
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Washington, D.C.
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Docket No. 50-267 SUBJECT: NRC Operational Sa'ety Team Inspection Report 88-200 REFERENCES:
1) NRC letter, Callan to Williams, dated December 7, 1988 (G-88495)
2) NRC letter, Holahan to Williams, dated September 16, 1988 (G-88372)
Gentlemen:
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This letter is in response to the Notices of Violations (Reference 1) received as a result of the review and evaluation of the unresolved items identified during the Operational Safety Team Inspection (OSTI) conducted by Mr.
J.
E.
Cumins and other NRC
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personnel during the period May 9-20, 1988.
Also, addressed (Reference 2) are unresolved items from G-88372.
By telephone on January 3,
1989, Mr. Westerman agreed that the response to this inspection could be delayed until January 27, 1989.
The followino j
responses (Attachment 1) to the five items included in the Notices of I
Violations and statements (Attachment 2) regarding the emaining unresolved items are hereby submitted.
In addition, actions taken or planned to improve the communications between manaaement and field
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personnel to assure full implementation of quality activities are
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addressed (Attachment 3).
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P-89024-2-January 27, 1989
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In general, Public Service Company haI1mplemented procedure rewrites and training to improve performance of its personnel and equipment.
We believe considerable progress has been made.
We also have committed to use these and other inspection results to continue improvement of the Fort St. Vrain operations.
Should you have any further questiers, please contact Mr. M. H. Holmes at (303) 480-6960.
Sincerely, 2'W W R. O. Williams, Jr.
Senior Vice President, Nuclear Operations R0W:DLW/pjb Attachments cc: Regional Administrator, Region IV ATTN: Mr. T. F. Westeman, Chief Projects Section 8 Mr. Robert Farrell Senior Resident Inspector Fort St. Vrain
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A.
Inadeouate Emergency Ooerating Procedures Technical Specification, Section AC 7.a.a.
" Procedures, Administrative Controls", states, in part, that procedures shall be established, implemented, and maintained as recommended in Aopendix A to Safety Guide 33, Novemoer 1972, including Secticn F, " Emergency Procedures".
Contrary to the above, emergency operating procedures were not adequately maintained as described in the folicwing examples:
-1.
Attachment I to SSC-03,
" Recovering From a Noncongested Cable Area Fire Resulting in an Interruption of Forced Circulation", Revision 1,
provided a table for defeating interlocks associated with valves, controllers, or other
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plant equipment which may become disabled, requiring the I
pulling of fuses or confirming the integrity of fuses, installing jumpers at terminal. blocks, or removing a grill cover and actuating relays located behind the grill.
For Valve SV-2106, the attachment required the operator to pull Fuses F-254 and F-134 at Panel I-05.
Fuse F-254 did not exist.
A dedicated supply of replacement fuses for replacing blown-cut fuses had not been provided.
The grill at the bottom of Panel I-10, Bay 800 could not be removed by use of a screwdriver, since one of the screws had been replaced by a hex head bolt.
Attachment 7 to SSC-03, provided an operator aid for the ecuipment operator to read thermocouple tempera tures from the'
temperature transmitter located in the auxiliary electrical room. The thermocouple reader was not located in the Shift Supervisor's office, as referenced in the attachment, but was located in the control room.
The standard screwdriver required to connect the thermocouple reader to the temperature transmitter was not located with the instrument.
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3.
Attachment 5 to SSC-04, " Recovery From SLRDIS", Revision 1, provided instructions for installing a through flange between the firewater system and the emergency feedwater header. Drain Valve V-45947 did not have a tag and could not be verified as the correct valve.
The mechanical
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spreader referenced in Steps 7 and 10 of the procedure could not be found in the sealed toolbox dedicated to performing this procedure.
This is a Severity Level IV violation.
(Supplement I)(267/88200-01)
(1) The Reason For The Violation If Admitted:
The violation is admitted.
The Safe Shutdown Cooling (SSi')
procedures, SSC-01 through SSC-05, are relatively new procedures at Fort St.
Vrain (FSV).
The procedures were
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implemented in mid-year 1987 to satisfy, in part
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' Appendix R. comitment..The procedures also were implemented to accomodate the newly installed Steam Line Ruoture Detection / Isolation System (SLRDIS) and improve the overall procedural approach to mitigating emergencies at the plant.
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The SSC's were implemented following an intense inter--
organizational effort to develop, prepare, ve ri fy, and validate the new procedures.
The implementation effort included a high level of input from operators.
The procedures were verified through operator table top reviews and walkdowns.
They were further validated by additional walkdowns and formal engineering evaluation.
The procedures were test run, and timed, under simulated emergency-ccnditions.
SSC-01 through SSC-05 were also processed through the normal procedure review and approval process at FSV.
Offficulties in maintaining the currency of SSC-03,
" Recovering From a Noncongested Cable Area Fire Resulting in an Interruption of Forced Circulation", were experienced. A typographical error occurred when the procedure was first issued; fuse F-254 should have been identified as fuse F-234 The procedure development team failed to recognize the need to dedicate fuses which are normally readily available to the control room staff. A standard screw fastener. was replaced with a hex head bolt during an undetermined maintenance activity.
Changes were made to standard operating practices which were not reflected in Attachment 7 to SSC-03.
The thermocouple reader noted was originally. assigned to the Shift Supervisor's Office.
The reader was relocated to the control room in the interests of security. Operations personnel were aware of the actual location of the thermocouple reader but the procedure had not been revised accordingly.
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Hardware discrepancies associated with Attachment 5 to SSC-04, " Recovery From SLRDIS", were due to several factors.
The tag for drain valve V-45947 was removed or lost at an j
undetermined time. The tag was in place when SSC-04 was
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walked down and test run prior to its initial issue. When SSC-04 was first developed, Public Service Company of Colorado (PSC)
intended to detemine if it would be necessary to procure a special
" mechanical spreader" to facilitate installation of the through flange previously mentioned.
It was detemined, by testing, that a special tool was not needed.
A wooden handle was used as the
" mechanical spreader" when the procedure was test run and the flange was actually replaced.
The wooden handle was located in the sealed tool box dedicated to perfoming this procedure, i
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All of the deficiencies noted with the maintenance of the SSC procedures can be attributed to the lack of a single I
group being assigned responsibility fer the procedure set.
(2) The Corrective Steos Which Have Been Taken and The Results Achieved:
SSC-03,
" Recovering From a Noncongested Cable Area Fire Resulting in an Interruption of Forced Circulation",
has been revised.
The reference to fuse F-254 was corrected to reference fuse F-234.
The dedication of a supply of replacement fuses was initiated concurrent with a revision to station material staging requirements. The hex-head bolt on. the grfil at the bottom of Panel I-10, Bay 800 was replaced with a standard screw.
Attachment 7 to SSC-03 was revised to reflect the control room location of ' the thennoccuple reader.
A standard screwdriver has been located with the instrument.
Drain valve V-45947 was retagged in accordance with the station's plant signage program.
Steps 7 and 10 of Attachment 5 to SSC-04 have been revised to correctly reference the use of a wooden handle to physically separate the flange faces.for through flange installation.
Ongoing maintenance of the SSC procedure set has been assigned to personnel involved in FSV's Emergency Operating Procedure (EOP) Rewrite Program.
(3) The Corrective Steps Which Will Be Taken To Avoid Further Violations:
Ongoing maintenance of the SSC procedures will remain the responsibility of the E0P rewrite program.
Effective interface between operations personnel and procedure writers has been established.
Feedback mechanisms to initiate procedure revision, as deficiencies are identified, have
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been developed in the areas of operator training and operations experience.
An additional revision to the SSC procedure set has been completed to maintain currency of the emergency procedures.
The OSTI team reviewed Issue 1 of SSC-03 and SSC-04. OSTI concerns were addressed in Issue 2 of the procedures. Additional operator input was solicited and subsequently addressed with another revision effort.
Issue 3 of each procedure was issued effective October 31 1988.
The remaining SSC procedures have received the same level of review and subsequent revision.
(4) The Date When Full Compliance Will Be Achieved:
Full compliance was achieved on October 31, 1988, with the issuance of Issue 3 of SSC-03 and SSC-04.
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Failure to Control Soecial Processes 10 CFR 50, Appendix S, Criterion IX, and the approved quality assurance program recuire that special orocesses, including welding, be controlled and accomplished using qualified procedures.
Contrary to the above, on May 11, 1988, the licensee performed a special process (welding) without a
qualified procedure.
Specifically, an unauthorized welding technioue was used to
remove broken drill bit parts from a reheat steam system l
thermocouple well.
A welding rod was placed down inside the thermowell and was momentarily energized to fuse the rod to the
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broken parts for retrieval. Station Service Request 88502797-03 authorizing removal of broken thermocouple by drilling, did not authorize the above activities nor reference any site welding procedure applicable to the activities.
Neither an engineering review nor application of any approved welding procedure were evident.
This is a Severity Level IV violation.
(Supplement I)(267/88200-06)
(1) The Reason For The Violation If Admitted:
The violation is admitted.
Although the special process noted in the violation received significant attention prior to being performed, personnel involved did not solicit the involvement of appropriate engineering personnel. The plan for addressing the work activity was not documented prior to implementation.
The violation is attributable to a lack of communication between personnel who perform maintenance, supervisors, planners, and personnel who support maintenance activities.
(2) The Corrective Steos Which Have 8een Taken And The Results
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Acnieved:
This issue has been discussed in detail with maintenance supervisory personnel.
Management expectations regarding prior documentation of work-plans and the need to acquire engineering assistance when needed were stressed. Also, all maintenance personnel were given training in the requirements of P-7, " Station Service Request Processing",
in June of 1988.
During this training, a general discussion of P-7 requirements, G-2 procedural requirements, and OSTI team criticisms was held.
Program requirements and nuclear standards for work documentation were emphasized.
The daily work control process at FSV has been improved
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through the institution of regular Plan Of The Day (POD)
meetings.
The POD meetings are coordinated by the Central Planning and Scheduling organization.
The following individuals attend this meeting:
Central Planning and
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o Scheduling Manager, Superintendent of Operations, Systems Engineering Supervisors, Maintenance Superintendents and Supervisors, Material Managerrent Supervisor, Health Physics Supervisor, Chemistry Supervisor, Nuclear Engineering Supervisor, Maintenance QC Supervisor and Licensing Supervisor.
Objectives of the POD meeting are to review current work activities, identify and coordinate support
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activities for work in progress, and to ensure good communications through all levels of management.
The meeting promotes awareness among all disciplines involved in supporting maintenance activities. Appropriate expertise is also made available through the inter-organizational makeup of the group. Also,' SMAP-29, " Station Service Request Work Planning", has been revised to emphasize requirements for work documentation when work-plans require revision during the course of work.
(3) The Corrective Steos Which Will Be Taken To Avoid Further Violations:
A high level of awareness regarding this issue is necessary in the area of first line supervision which has the responsibility of interfacing with support groups and craft personnel.
In addition to the prior documentation of work-plans noted above, the Maintenance Manager will conduct a femal training seminar for all Maintenance supervisory personnel to increase their understanding of the reoufrements for proper documentation of work-plans prior to job performance.
The seminar will also address the need for appropriate Engineering involvement prior to the conduct of work activities which are not covered by established procedures or which are not routine in nature.
(4)
The Date When Full Como11ance Will Be Achieved:
POD meetings will continue and be enhanced as necessary.
Full compliance will be achieved following ecmpletion of the
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supervisory seminar presented by the Maintenance Manager.
The seminar will be completed by February 1, 1989.
C.
Inadeouate Procedure for Presetting a Valve Position. Controlling StoD Work, and Controlling Nonconformances 10 CFR 50, Appendix 8, Criterion V, requires that safety-related activities be accomplished in accordance with procedures appropriate to the circumstances, and that the procedures include appropriate quantitative or qualitative acceptance criteria for determining that the activities have been satisfactorily accomplished.
Contrary to the above the following procedural deficiencies were found:
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1; The disassemb1'y, reassembly, and. repositioning'of the Steam
' Generator B-2-6 Trim. Valve TV-2228-6, authorized'
in accordance with. Station Service Recuest '88502035, 'was r
. accomplished without. benefit of either 'a procedure or-
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acceptance criteria.
Plant design recuired that the trim
' valve stem be set.so that, with the ' valve fully. shut,' it would pass a minimum 20 percent full rated feedwater flow.
The valve was' disassembled and reassembled using scribe marks on the. valve body to preposition' thel stem; no i
procedure nor engineering information 'was available to either substantiate.the source and accuracy of-the scribe
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marks or to provide alternate assembly instructions and criteria.
(267/88200-08)
2.
Procedure P-12, " Plant Maintenance", Issue 5, paragraph 4.4, and Procedure MPRM-13. "Stop Work", Issue-2, regarding "stop work" activity, did not provide adequate guidance or
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necessary authority to enable the-OC.
inspectors and.
supervisor to exercise "stop work" actions when actua'l field conditions warranted.
(267/88200-12)
3.
Administrative Procedures. Q-15,
" Nonconforming Reports",
Issue -7; Q-16, " Corrective Action System", Issue 8, dids not-provide adequate guidance for evaluation and documentation -
of the cause, and the corrective. actions to preclude repetition regarding Nonconformance Reports NCR 85-042.85-043, 87-607,88-002, and~88-603.
(267/88200-13)
This is a Severity-Level IV violation.
(Supplements)
(1) The-Reason For The Violation If Admitted:
The violation is admitted.
,The. violation was due to inadequate procedures.
1.
RP-900,
" Calibration. and Maintenance of Annin Domotor Actuators and Positioners Sizes A',~8, C, D and DD",-
is a, generic calibration procedure for these types of
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valves. These types of valves are in service in varying applications throughout the plant.
When applied as steam generator trim valves, the valve's'
service is modified by limiting valve stem travel so that the' valves cannot close completely.
The valve calibration procedure system. utilized at FSV accommodates.special applications for generic valve types b Sheet (y providing a service _ specific Valve Calibration VCS) as an attachment to the procedure. The VCS is issued concurrently with the generic calibration procedure Service specific information. is maintained in the Flant Component Data 8ase.
For the valve in question, the VCS issued to document the work underway did not include available service specific information specifying valve adjustments to limit valve stroke.
However, other valves in identical service had VCS i
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maintenance technicians performing the work were familiar with the service specific requirements of:this
- valve. in this service.
The maintenar.ce technicians proceeded to accomplish work activities which they knew-to be' necessary and correct, but which.were~ not properly. supported by the-procedures or work instructions' issued to them.
2.
Procedures P-12 and MPRM-13 were not written adeouately to describe, for the.' QC inspector, the stop work-process.
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Administrative procedures,0-15 and 0-16 do address the
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evaluation and trending of -NCR's..
Q-15, Issue 7,
states that during Engineering disposition, the root:
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cause evaluation as well as operability per design shall be documented for QA approval.
Further, the NCR
.shall be evaluated for repetitive nonconformances that may require further cor'rective actions.
Q-16, Issue 8, specified that the QA Operations Manager is responsible for -accumulation and analysis 'of data for quality trends.
In addition, 0-16 provided a description of the Quality Trends Analysis Report.
These procedures, at the time of inspection, did not provide complete guidance for evaluation and corrective actions.
'(2) The Corrective ' Steps Which Have Been Taken And The Results Achieved:
s Enhancements made to the daily' work control process as discussed in the response to: Notice of Violation
(267/88200-06) will also serve to prevent the deficiencies noted above.
1.
All-trim valves, TV-2227-1 through TV-2227-6 and TV-2228-1 through TV-2228-6 were' evaluated for the correct
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placement of punch marks used for limiting valve stroke in this application. The valves were disassembled for the perfomance of preventive maintenance, and as a followup to this violation during the recent circulator repair. outage.
During reassembly, the existing punch
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marks were checked.
It was detemined that all punch marks were correctly located to provide for a valve stroke. limit to keep the valve discs 1 inch off of the valve seats when the valves were closed. Data Base Change Request (DBCR)88-317 was initiated on May 13, 1988, to verify-and correct the information in the Plant Component Data Base concerning all twelve trim
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valves.
It was detemined that the VCS data base was
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lacking reference to the punch marks in the cases of two of the twelve trim valves.
The proper information I
was included in the data base in these two cases.
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Attachmeng 2
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go Pr89024 dated
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2.
Procedures MOCIM-1,
" Maintenance Ouality Control Inspection Manual".
Issue 8,
and OCIM-7,
"Cuality
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i Assurance Inspection Manual",
Issue 5,
have been revised to provide specific guidance and criteria fo r Maintenance Quality Control (M0C) personnel and Quality i
Assurance / Quality Control (OA/0C)
inspectors to
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exercise "stop work" authority.
These procedures also j
provide guidance for follow up documentation i
requirements following a
"Stop Work" Order.
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Administrative Procedure 0-10, "!nspection",
Issue 9,
now provides adequate authority to CA personnel to
"stop work".
The Manager, OA or his designated alternate has the authority to order that work be stopped when, in his opinion, repair or modification work does not comply with the approved controlling documents.
Additionally, Administrative Procedure P-
" Plant Maintenance" Issue 5, has been revised to provide general guidelines for the stop work process.
3.
The responsibility for data evaluation and corrective actions has been established through a revision to
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Administrative Procedure 0-16.
0-16 identifies trending programs as a method of identification and initiation for corrective action.
The procedure states that NCR's are initiated and processed per Administrative Procedure 0-15
" Quality Assurance Operations Procedure".
" Quality Trending Analysis Program", provides trending) analysis of NCR's, Quality Deficiency Reports (QDR's and corrective Action Reports (CAR's),
i (3) The Corrective Steos Which Will Be Taken To Avoid Further Viciations:
1.
An elevated level of management attention is being directed to the issues of procedure adecuacy and procedural compliance at FSV.
These areas were
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addressed during the recent Maintenance Self Assessment (MSA)
performed under INPO guidelines.
Specific, ongoing actions to address these deficiencies were developed as a result of the MSA. Actions include:
ongoing reviews and procedure enhancements, verification and validation efforts for revised procedures, and development of craft ownership of procedure adequacy.
2.
No additional actions are required for improvement of
"stop work" authority procedures.
Instruction on the revised procedures was completeo May 16, 1988.
3.
Additional revisions to 0-15 have been initiated. A new Level II procedure is being developed to provide additional guidance to Quality Assurance Personnel for the evaluation and documentation of NCR cause(s) and
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actions taken to. preclude recurrence; The trendino system is also being revised to improve NCR : trending
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. analyses and adequately _
identify repetitive nonconforming conditions.
Training seminars. will be used within the Quality Assurance organization to train personnel that will be implementing the newly revised procedures.
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(4) The Date When Full Comoliance Will Be Achieved:
1.
Management attention to improvement. of procedure adequacy. and procedaral. compliance. continues.
The specific findings have been corrected and no further action is required to verify the punch mark locations on the trim valves.
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2.
-The procedures relating to "stop work" authority have been improved.
No further action is recuired at this time.
-3.
Final revisions to-Q-15 and the trending system, as well as issuance of the new level II quality procedure, will be accomplished by May 31, 1989.
D.
F,ailure to implement Cleanliness and Material Control Procedures 10 CFR 50, Appendix 8 Criterion V, requires that safety-related activities be accomplished in accordance with procedures of a type appropriate to the circumstances.
Maintenance Administrative Procedure MAP-8, " System and Component Cleanliness Requirements During Performance of Maintenance Activities",
Issue 3,
required that.the work area be cleaned
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after each operation that generated potential contaminants.
Maintenance Administrative Procedure MAP-7, " Parts Identification and~ Control", Issue 1,
required that maintenance personnel
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performing the work shall ensure that components and parts were packaged and identified - as they were removed from plant equipment.
The-procedure further required that removed components and parts receive the required degree of protection while they were removed from the system.
Contrary to the above, the following examples of failure to accomplish the stated procedural requirements were identified:
1.
During performance of Station Service Request 88502490 (for repair of Loop 1 Turbine Water Header Isolation Valve HV-21243) debris from lagging removal was not cleaned up at the job site until several days after the system had been opened I
with the debris present.
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During the.above' repairs, valve HV-21243 internals'were removed from the valve but were left unprotected and J
untagged on the floor in the vicinity of_the valve.
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W This is a. Severity Level IV violation.
(Supplement I)(267/88200-
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(1)' The Reason For The, Violation If Admitted:
The violation is admitted. Maintenance personnel did not.
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take care in maintaining appropriate cleanliness around the
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work area whfie the system was open for maintenance.
Parts
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removed from the system were not appropriately handled while being : staged _ at the work area.
MAP-8 contained sufficient administrative controls but its requirements were not met by workmen.
The administrative controls present in MAP-7 were
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not sufficiently clear to give workmen proper guidance.
(2) The Corrective Steps Which Have Been Taken And The Results Achieved:
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Maintenan'e Administrative Procedure MAP-7,
" Parts c
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Identification and Control", has been revised _ for clarity.
Issue 2 of MAP-7 was issued effective January 5,1989.
Maintenance Supervisors are being required to ' spend more time at the work sites observing the work.
They also have l-been instructed to increase their attention to work site cleanliness,and materials. control during observance of ongoing work.
Supervisors are' to ensure a continuing-awareness of the need fore adherence to established administrative controls at the craft level.
(3) The Corrective Steps Which Will Be Taken To Avoid Further violations:
A training seminar is planned for all maintenance personnel.
The seminar will address and reinforce the requirement for
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increased attention to this area.
(4) The Date When Full Compliance Will Be Achieved:
Full compliance will be achieved when the noted training seminar.has been completed.
The training will be complete by February 28, 1989.
<
E.- Failure to Perfom Technical Specification Inservice Testing i
Technical
. Specification 5.0,
" Surveillance Requirements",
J specifies, in part, the test, calibrations, and inspections necessary_ to verify the perfomance and operability of equipment essential.to safety during operation and abnormal situations be conducted.
The following are examples of failure to conduct required surve111ance.
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Technical Specification SR 5.2.7,
"Wa ter Turoine Drive Surveillance", requires, in part, that, "a) One circulator and associated water supply valving in each loop will be functionally tested by cperation on water turbine drive using feedwater,... annually".
Contrary to the above, Surveillance Procedure SR 5.2.7al-A,
Issue 3 " Loop I/ Loop II Valves ano Circulator Drive Tests",
included an exemption for Valve V-22371, emergency feedwater (EFW) header check valve.
The completion of the annual testing of Valve V-22371 to verify its operability was not evident.
2.
Technical Specification 5.3.4,
" Safe Shutdown Cooling Valves", requires, in part, that valves (pneumatically, hydraulically, or electrically opera ted valves; normally closed check valves; and manually operated valves
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including normally closed power operated valves) shall be tested for operability by partial stroking every 92 days unless they cannot be operated during normal operation, and a full functional test shall be performed annually, or at the next scheduled shutdown, not to exceed the 18 month surveillance interval.
Contrary to the above Surveillance Procedures SR 5.3.4bl-A (2-A), " Loop I (II) Safe Shutdown Cooling Power Operated Valve Tests",.did not include the testing of the bearing water filter isolation valves (HV-2153-1 and HV-2153-2).
The valves' had been exempted from testing by the licensee and had not been tested annually prior to May 12, 1988, when the licensee tested the valves after the identification of the deficiency by the NRC.
This is a Severity Level IV violation.
(Supplement I)(267/88200-10)
(1) The Reason For The Violation If Admitted:
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The violation is admitted.
Because the performance and operability of the valves noted in the two parts of the Notice of Violation is regularly demonstrated through plant operations or by the performance of testing on other aspects of their associated systems, the valves were exempted from the formal surveillance test procedures as noted.
Survef11ance Procedure SR 5.2.7al-A, " Loop I/ Loop II Valves and Circulator Drive Test",
listed V-22371 as " exempt
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i nomally open when emergency feedwater header is in service".
V-22371 is a check valve in the emergency feedwater line to the Back Up Bearing Water (BUBW) header.
The BUBW header supplies bearing water surge tank makeup for nomal bearing water.
The BUBW header also supplies backup bearing water in the case of loss of normal bearing water.
Water to the pelton drives on the helium circulators is also
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.supp1 fed by th'e BUBW system.
During normal plant operation, the BU8W header' is in service with V-22371 in the.open position to provide a flow of water to maintain bearing water surge tank level. During normal operation, V-22371 is in the correct position for safe shutdown' cooling via the emergency feedwater header. Although not verified by direct observation, SR 5.2.7al-A/5.3.4al-A does demonstrate that V-22371 is in the open position by the flow of water from-the emergency feedwater header to the circulator 1 pelton drive wheels.
Adequate flow is demonstra ted by comparing circulator speed obtained in this configuration with design values.
Another Surveillance Procedure, SR 5.2. 7a2-A/5.2.23-A1/5.3.4a2-A verifies operability of V-22371 in the closed position.
During this test, operation of the circulators on the pelton wheel drive is verified with the emergency feedwater header being supplied by simulated.
boosted firewater.
The testing performed by SRS.2.7al-A/5.3.4al-A and SR 5. 2. 7a2-A/5. 2.23-A1/5. 3. 4a 2-A demonstrates operability of the valve, in accordance with the methods of ASME code Section XI, Div 2. Art.
IGV-3412.
Surveillance Procedures SR 5.3.4bl-A(2-A), " Loop I (II) Safe Shutdown Cooling Power Operated Valve Tests", listed valves HV-2153-1, 2153-2, 2154-1 and 2154-2 as exempt due to the fact that. these valves are routinely stroked during normal plant operations.
These valves are the inlet block valves for Loop I and Loop II normal bearing water filters.
The basis for. the exemption is that the valves are stroked every time that the on-line filter is isolated for cleaning.
A search of Station. Service Requests -(SSR's) initiated to accomplish cleaning of the normal bearing water filters indicates that these valves were stroked 40 to 60 times during the last two years.
(2) The Corrective Steps Which Have Been Taken And The Results Achieved:
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Valves HV-2153-1, 2153-2, 2154-1 and 2154-2 were tested and found to be operable. Other Surveillance Procedures were reviewed to detemine if they contained similar inappropriate exemptions.
This review identified the
"examption" of V-22371 on May 18, 1988, prior to the completion of the OSTI inspection.
PSC has subsequently evaluated avaffable test data and determined that the perfomance and operability of V-22371 has been adequately demonstrated as discussed above. Surveillance Procedures SR 5.2.7al-A/5.3.4al-A, SR 5.3.4bl-A and SR 5.3.4b2-A have been revised to formally test and document the operability and performance of the noted valycs.
(3) The Corrective Steos Which Will Be Taken To Avoid Further
_VI'oTa tions:
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Surveillance Procedures SR 5. 2. 7al-A/5. 3. 4al-A, and SR 5.2.7a2-A/5.2.23-AI/5.3.4a2-A will be revised to show that V-22371 is tested by these procedures.
Since SR 5.3.4bl-A and SR 5.3.4b2-A have been issued with the corrections necessary to include HV-2153-1, HV-2153-2, HV-2154-1, and HV-2154-2, no further action is required.
(4) The.Date When Full Comoliance Will Be Achievec:
Full compliance will be achieved following the issuance of the revised Surveillance Procedures by February 28, 1989.
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Attachment 2
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to Pr89028 dated
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, January 27, 1989 Page 1 (1) Unresolved Item (267/88200-02):
Investigate provision for control of checklists for complex activities.
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I PSC RESPONSE l
PSC agrees that several System Operating Procedures (50P's)
cortain valve line up changes and activities without associated checklists.
Checklists have already been incorporated into critical operations procedures as determined by internal assessments of the needs of the operating staff.
Certain critical operations activities are performed in accordance with formal procedures which do include checklists.
Overall Plant Opera ting Procedures (0 POP's) include checklists for the performance of critical integrated operations of plant systems.
They have been divided into categories including:
Plant Start-up, Technical Specifications and Shutdown.
Checklists have also been incorporated into procedures governing processes where the operating experience at Fort St.
Vrain has warranted that it be done.
Foi example: Nuclear Production Administrative Procedure, NPAP-19, " Radioactive Gaseous Effluent Releases", was revised to include checklists to document and coordinate complicated activities performed inter-departmentally during the process of making gas waste releases at fort St.
Vrain.
Plant management routinely initiates complete integrated valve line-up walkdowns, with independent verification, for systems affected by significant maintenance or construction activities.
The need for these walkdowns is determined by the Superintendent of Operations.
A general revision to the existing plant SOP's is not planned. A revision of the 50P's for the purpose of including checklists and independent verification would be of little benefit, would be
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time consuming, and would be resource intensive.
Inclusion of checklists with SOP's would also represent a major change for our operating staff and would require an extensive retraining effort.
(2) Unresolved Item (267/88200-03): Develop a lesson plan for auxiliary tender trainees on the subject of valve operations.
PSC RESPONSE A new lesson plan for Valve Operations and Indications has been developed.
Training Department Request (TDR) No. 051388-1 has been completed.
This lesson plan (AT-105) has been incorporated into the Outside Auxiliary Tender Program which is the initial training program for auxiliary tender trainees.
Training in accordance with the new lesson plan started with the outside auxiliary tender class which began on May 8, 1988.
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January 27, 1989
'Page 2 (3) Unresolved Item (267/88200-04): Licensee action to ensure correctness of the TCR log and reduce the backlog of old open TCR's.
PSC RESPONSE The methodology used for maintaining the TCR index has been reviewed. A checklist for the control room TCR book has been i
formulated and is being reviewed prior to permanent incorporation into administrative controls governing the TCR process.
The new checklist will provide additional assurance that the recuired update of the TCR fndex will occur whenever a TCR is initiated.
Steps taken in the initiation and implementation of a TCR will be documented by the checklist.
The " Checklist for Control Room TCR Book" will be incorporated into Station Managers Administra tive Procedure SMAP-18,
" Processing of Temporary Configuration Reports".
The revision will be accomplished by February 1, 1989.
The backlog-of TCR's has been reduced significantly at FSV.
Strategies have been developed to further reduce the backlog.
Administrative controls have been established which ensure that the implementation of new TCR's are kept to an absolute minimum and that existing TCR's are constantly monitored for necessity.
The TCR backlog as of January 13, 1989, stands at 48 TCR's. Of these 48 TCR's, 28 were created prior to 1988 and are installed in non-safety related applications.
Plans for removal of the non-safety related TCR's had been developed and were to be accomplished during the fourth refueling outage.
However, removal of the non-safety related TCR's is being reassessed
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relative to announced plans to cancel the fourth refueling outage and end operations at FSV.
Fifteen of the 48 TCR's are new TCR's which were implemented
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under strict administrative controls. New TCR's must receive an additional review by the Station Manager and a fomal Engineering Evaluation, including review by the Manager, Nuclear Engineering Division, if they are to remain installed for more than 90 days.
Additional approval by the Sr.
Vice President of Nuclear Operations is required if the TCR is to remain in service for more than 180 days.
Five TCR's were associated with the original series 8 testing which, to date, has not been completed. A decision on these five TCR's has not Deen reached.
However, alternatives are being evaluated.
(4) Unresolved Item (267/88200-05):
Inadeouate training, minimum operations staff required for remote shutdown concurrent with a fire, and whether an integrated test of
.? mote shutdown
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capability should be performed.
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PSC RESPONSE This concern has been analyzed by operations management, nuclear engineering, and personnel in the Operator Training Unit of the Nuclear Training Department.
The Operator Training Unit has determined that training for specific components of the remote shutdown activity is provided as a part of position specific training. As noted in the OSTI inspection report (Reference 2), licensed operators are trained on remote shutdcwn every two yea rs.
Equipment operators and auxiliary tenders are trained in specific activities, which are a part of remote shutdown, as components of the ongoing training they receive.
For example: Equipment operators are trained in manual boiler feed pump operation but they are not trained in that task specifically during remote shutdown.
The Operator Training Unit will develop team training specific to the remote shutdown task. A tentative schedule for the training is included in the early rotations of the 1989/1990 requalification program.
Remote shutdown training will consist of academic lesson plans and simulated crew perfomance in the field. The training will include both licensed and non-licensed personnel as a team. There will be a minimum of two instructors available for the evaluation and implementation of this training.
Further evaluation of the need for an integrated test of remote shutdown capabilities is planned for the first quarter of 1989.
Representatives from Operations, Nuclear Engineering, Systems Engineering, Licensing, and Nuclear Training will address this issue with regards to existing tests, equipment and training.
This group will recomend a course of action to Plant Management by April 30, 1989.
(5) Unresolved Item (267/88200-07): Licensee actions to correct the other maintenance program and implementation weaknesses in paragraphs 2.2.3.1 (1),
(2), (3), (5), (ti), (7) and (8) of the OSTI inspection report (Reference 1).
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PSC RESPONSE The corrective actions identified in response to Notice of Violation 8 (267/88200-06) will also correct most of the problems identified in this unresolved item.
Considerable effort has been and will continue to be expended toward improving work control. Numerous sessions have been held with the craftsmen and supervisors to sensitize them to the issues.
Some disciplinary actions have been initiated in accordance with the station progressive discipline policy for cases of failure to follow procedure.
Improvements have been realized.
However, further management
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attention will continue to be directed toward additional
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An approach to addressing these problems is to l
increase the presence of first line supervisors in the field to i
allow the craftsmen additional resources for proper resolution of work related problems. Management direction is available when supervisors require specific instructions for proper courses of action if there is doubt.
(6) Unresolved Item (267/88200-11): Licensee actions to improve and ensure the OIAG's continuing effectiveness.
PSC RESPONSE As part of the reorganization of Nuclear Operations in May 1988, the Nuclear Regulatory Affairs Department assumed responsibility for the operating experience program at Fort St. Vrain.
The
group responsible for review of operating experience has been
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restructured and renamed as the Operating Experience Board (OEB).
OEB functions are defined in Administrative Procedure G-24,
" Operating. Experience Program", and further clarified in Nuclear Licensing Procedure NLR-13. " Operating Experience Program".
The OEB is composed of a membership drawn from Operations, Nuclear Training, Systems Engineering, Maintenance and Nuclear Regulatory Affairs.
OEB meetings are held weekly.
The activities of the OEB are documented and distributed to all levels of management in the Nuclear Operations organization.
Meeting minutes are also distributed to the membership.of the Nuclear Facility Safety Consnittee (NFSC).
OEB members meet once a month with representatives from Licensing, Quality Assurance and Nuclear Engineering to review OEB actions.
The OEB screens operating experience documents received from the Nuclear Regulatory Commission and the Institute for Nuclear Power Operations.
Fort St. Vrain generated issues are also reviewed.
The activities of the OEB fully address the weakness identified in this unresolved item.
The group is effective and receives the
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full support of the Nuclear Operations organization.
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