ML20141C474
| ML20141C474 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 03/27/1986 |
| From: | Dubois D, Jaudon J, Plettner E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20141C424 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-1.A.1.3, TASK-1.C.1, TASK-TM 50-298-86-02, 50-298-86-2, GL-82-33, NUDOCS 8604070254 | |
| Download: ML20141C474 (18) | |
See also: IR 05000298/1986002
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APPENDIX
U. S. NUCLEAR REGULATORY COMMISSION
REGION IV
NRC Inspection Report: 50-298/86-02
License: DPR-46
Docket: 50-298
Licensee: Nebraska Public Power District (NPPD)
P. O. Box 499
Columbus, NE
68601
Facility Name: Cooper Nuclear Station (CNS)
Inspection At: Cooper Nuclear Station, Nemaha County, Nebraska
Inspection Conducted: January 1-February 28, 1986
Inspector:
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E. A. Plettner, Resident Inspector, (RI)
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Inspector:
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D. L. DuBois, Senior Resident Inspector, (SRI)
Date
Approved:
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J. /P. Qaudorf, Chief, Project Section _ A,
Date
Merctor Prbject Branch
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. Inspection Summary
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Inspection Conducted January 1-February 28,1986'(Report 50-298/86-02)'
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Areas Inspected: ' Routine, u'nannounced inspection of followup of TMI action
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plan requirements, licensee managemcnt changes', cold weather preparation, .
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preparation for refueling, spent, fuel shipments, plant trips - safety system
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challenges,' operational _ safety verification, and mo'nthly surveillance and
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- maintenance-observations.
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Resultisi LWithin the nine areas-. inspected, one violation was identified
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(inadequateprocedure--paragraph 2).
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DETAILS
1.
Persons Contacted
Principal Licensee Personnel
+* Jerry Sayer, Acting Technical Staff Manager
+R. Brungardt, Operations Manager
+C. Goings, Regulatory Compliance Specialist
+V. L. Wolstenholm, QA Manager
+D. M. Norvell, Maintenance Manager
+*J. M. Meacham, Technical Manager
+*G. R. Horn, Division Manager of Nuclear Operations
- P. V. Thomason, Senior Nuclear Advisor
The NRC inspectors also interviewed other licensee operations,
maintenance, and administrative personnel.
NRC Personnel
- J. P. Jaudon, Chief, Reactor Project Section A
- W. M. McNeill, Project Inspector
- R. E. Baer, Radiation Specialist
+*D. L. DuBois, Senior Resident Inspector
- E. A. Plettner, Resident Inspector
+ Indicates presence at exit meeting held January 17, 1986.
- Indicates presence at exit meeting held February 28, 1986.
2.
NUREG-0737, TMI Action Plan Requirements
a.
(Closed) Item I.A.I.3, Section 2.A, " Shift Manning - Minimum Shift
Crew Complement"
In a letter from Mr. D. Eisenhut (NRC) to All Licensees of Operating
Plants, dated October 31, 1980, the NRC forwarded to each licensee a
copy of NUREG-0737, " Clarification of TMI Action Plan Requirements."
This document provided a comprehensive and integrated plan to improve
safety at power reactors.
Initial TMI related actions were
identified and included in previously issued NUREG-0660. NUREG-0737
included only those action items from NUREG-0660 that were approved
for implementation at that time.
Item I.A.1.3(2.A) of NUREG-0737 provided the licensee with interim
shift staffing requirements pending the approval and cublication of
future rulemaking in that area.
The interim require:nents included
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Meet the stated guidance for interim shift staffing by July 1,
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1982.
Establish administrative procedures which set forth the policy,
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objectives, and manning criteria of the requirement.
Change the CNS Technical Specification to reflect minimum shift
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crew manning requirements.
The NRC promulgated final shift manning requirements on July 11,
1983.
Those requirements were published in 10 CFR 50.54(m) on
January 1, 1984.
An NRC inspection was performed in this area during the period
Novemicer 1-30, 1982; this inspection was documented as NRC Inspection
Report 50-298/82-33. That inspection determined that licensee
actions met the interim shift manning requirements specified in
NUREG-0737 with the exception of necessary administrative procedure
and Technical Specification changes.
During this inspection period the SRI reviewed the requirements _of
10 CFR 50.54(m), CNS Technical Specification Section 6.1.3, and CNS
Administrative Procedure 2.0.3.
The present Technical Specification
revision and Proposed Change No. 16 to the Technical Specification
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dated February 22, 1985, were found to meet the requirements of
10 CFR 50.!4(m), with the exception of the following deficiencies:
The recuirement that a senior reactor operator (SR0) be in the
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control room at all times when the unit is operating is not
stated.
It is noe stated that a licensed reactor operator (RO) or SRO be
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present at the controls at all times when there is fuel in the
reactor.
The term, 'at the controls," is not defined.
The CNS Technical
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Specificat3on used the term, "in the control room." The control
room inclucis not only the operating consoles but also the
process cab' nets, reactor protection system panels, secondary
control panels, and a lunch area, which are located behind the
operating consoles.
The CNS Conduct o' Operations Procedure 2.0.3, " Control Room Conduct
and Manning," Revision 1, dated December 26, 1985,Section II.F,
delineates control com manning and watch requirements.
Procedure 2.0.3 rep; aced a portion of previous CNS Administrative
Procedure 1.4, " Station Rules of Practice," that was discussed in NRC
Inspection Report 50 298/82-33.
Procedure 2.0.3,Section II.F, does
not state the requirenents of 10 CFR 50.54(m) as delineated
below:
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Subsections 1 and 2.a of the procedure state that manning
requirements are detailed / described in the Technical
Specification. This statement is unsatisfactory because of the
Technical Specification deficiencies noted above.
Subsection 2.c of the procedure does not include the requirement
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that one SRO must be in the control room during unplanned or
unscheduled shutdowns or during a significant reduction in
power, and while refueling operations are in progress.
Subsection 2.e of the procedure states that an RO or SRO must be
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in view of the front control room panels at all times instead of
"at the controls" at all times when there is fuel in the
reactor.
10 CFR 50.54(m) specifically provides separate and distinct manning
requirements for the facility, control room, and "at the controls."
The CNS Technical Specification and Procedure 2.0.3 do not clet*1y
state these requirements. By repeated observation, the SRI noted
that the licensee's practice appeared to meet the requirements of
10 CFR 50.54(m). Thus, it was concluded that the licensee's
Procedure 2.0.3 did not implement the regulations and that the
Technical Specifications did not clearly reiterate the requirements.
of 10 CFR 50.54(m). The inadequate procedure is an apparent
violation (8602-01).
b.
(Closed)
Item I.C.1, Section 2.8, " Guidance for the Evaluation and
Development of Procedures for Transients and Accidents - Revise
Procedures for Inadequate Core Cooling."
On December 17, 1982, the NRC published Generic Letter 82~33,
" Supplement 1 to NUREG-0737-Requirements for Emergency Response
Capability." That document provided additional clarification of the
NUREG-0737 requirements in the area of emergency operating procedures
(EOP) upgrade and implementation. Also, the licensee was requested
to furnish a proposed schedule for completing the specified
requirements by April 15, 1983. The licensee was reqeired to perform
the following actions in this area:
Upgrade E0Ps in accordance with NRC requirements, technical
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guidelines, an appropriate procedure writers guide, and licensee
commitments.
Verify that generic emergency procedure guidelines (EPGs) are
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revised to incorporate plant-specific data, tables, and
instrumentation information.
Validate the initial procedure drafts to verify the technical
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adequacy of the E0Ps text and procedure steps.
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Provide formal training of operating personnel on the use of the
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E0Ps prior to implementing the E0Ps.
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Approve the upgraded, plant-specific E0Ps in accordance with the
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requirements found in the CNS Technical Specification.
Implement the E0Ps.
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In a letter from Mr. D. Eisenhut to All Boiling Water Reactor
Licensees of Operating Reactors, dated February 4,1983, the NRC
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The subject of the SER was a safety evaluation of " Emergency
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Procedure Guidelines, Revision 2," NED0-24934, June 1982. That SER
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indicated the NRC's acceptance of those guidelines for implementation
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The licensee responded to Generic Letter 82-33 and the'aboy'e SER in a
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letter from Mr. J. Pilant (NPPD) to Mr. D. Eisenhut, dated April 15,
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1983.
The licensee estimated that the E0Ps would be--implemented on
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or about December 31, 1984. On December 30, 1984, the licensee
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submitted Revision 4 to their response to Generic Letter 82-33, which
. changed the E0P implemention date to September 30, 1985.
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Subsequently, the NRC issued an Order Modifying License on August.29,.
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1985, which ordered the licensee to meet their proposed
implementation date.
The SRI verified through discussions with personnel, review of
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documentation, and observation, that the licensee completed the
following actions:
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Symptomatic E0Ps were developed in accordance with the
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requirements and guidance of:
(1) NUREG-0737
(2) NUREG-0737, Supplement 1 (Generic Letter 82-33)
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(3) NUREG-0899, " Guidelines for the Preparation of Emergency
Operating Procedures," Revision 5, June 4, 1982
(4) General Electric Topical Report NE00-24934, " Emergency
Procedure Guidelines," Revision 2, June 1982, and later
errata.
(5) Boiling Water Reactors Owners Group (BWROG) Emergency
Procedure Guidelines (EPG), Revision 31, March 1984,
generic procedures for BWRs 1 through 6.
(6) CNS plant ~ specific EPGs, Revision 1, May 3, 1985.
(7) ' Emergency Operating Procedures Writing Guidelines, INP0
82-017, July 1982.
(8) CNS Guideline for Preparation of Emergency Operating
Procedures, Revision 2, May 3, 1985.
.(9) NUREG/CR-2005, " Checklist for Evaluating Emergency
Operating Procedures Used in Nuclear Power Plants,"
(10', CNS Emergency Procedure Validation Program, April 23, 1984.
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Generic EPGs were revised to incorporate plant-specific
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information. Deviations from the recommendations of the BWROG
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EPG, Revision 31, were documented on a CNS " Step Documentation"
status form.
Revisions to the plant specific EPGs were documented on " Summary
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of Changes to CNS EPG," Revision ', May 3, 1985.
E0P procedure drafts were validated for technical adequacy
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during the initial procedure development and review process, at
a simulator by licensee SR0s and training instructors, and by
CNS control room walkdowns by licensed operations personnel.
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CNS Emergency Procedure 5.8, " Emergency Operating Procedures,"
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was developed from approved plant-specific EPGs.
Emergency
Procedure 5.8 consists of five sections which include:
(1) E0P/C - Operator Precautions
(2) E0P - 1 - Reactor Pressure Vessel (RPV) Control
(3) EOP - 2 - Primary Containment Control
(4) E0P - 3 - Secondary Containment Control
(5) E0P - 4 - Radioactive Release Control
CNS Emergency Procedure 5.8, Revision 1, was reviewed by the
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Station Operations Review Committee (SORC) and approved for
implementation August 1, 1985.
The SRI observed that approved E0Ps were used by licensee
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personnel during the annual emergency preparedness exercise
conducted on October 16, 1985.
A review of formal licensee training on the use of E0Ps is
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presently in progress. The results of that review will be
documented in NRC Inspection Report 50-298/86-10.
This item is considered closed.
c.
(Closed)
Item I.C.1, Section 3.B, " Guidance for the Evaluation and
Development of Procedures for Transients and Accidents - Revise
Procedures for Transients and Accidents."
This item was inspected in conjunction with item I.C.1, Section 2.B,
above. This item is considered closed.
3.
Licensee Management Changes
a.
On January 1, 1986, Mr. Guy R. Horn replaced Mr. Paul V. Thomason as
the Division Manager of Nuclear Operations (DMNO) at Cooper Nuclear
Station.
Mr. Thomason was appointed to the position of Senior
Nuclear Advisor for the licensee's Nuclear Power Group.
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Mr.LHorn has-accunnlate[24 years of service with*NPPD. - Past
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experience included' operator's duties at the Hallam Nuclear Power
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facility, licensed SR0 nhift supervision and operations departmenta
management. at the CNS, nanager, of. hydroelectric and fossil electric
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. generation sites construction' activitics, and site manager of the BWR
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pipe replacement project at CNS during 1984-1985. ' Mr.' Horn appears
to meet the requirements.of ' ANSI 18.1(1971) except for the fonnal
academic requirements, .but his principal assistantstmeet these
requirements. This appears to satisfy the requirements of ANSI 18.1
(1971),
b.
NPPD changed the titles' for senior management positions as follows:
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fsam President of the NPPD Board of Directors to Chairman of the
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from Vice Presidents of the Board to Vice Chairmans
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The purpose of the title changes was to more closely match equivalent
position titles used in other-utilities. -The licensee ~has submitted
a revision to Proposed Change No. 28 to the CNS Technical
Specification, Figure 6.1.1, which will revise the title changes as
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applicable. .That revision was submitted _to the NRC in a letter from
,Mr. J. Pilant to Mr. D. Muller (NRR) dated January 13,~1986.
4.
Cold Weather Preparation
The purpose of this inspection was to verify that adequate protective
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measures had been taken to prevent safety-related process, instrument, and:
sampling lines from freezing during extremely cold weather.
The SRI reviewed logs, perfanned app. icable systems walkdowns, and
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conducted discussions with licensee' personnel to determir.e adequacy and
continuing implementation of the licensee's protective measures. The SRI
verified the following:
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Individual plant systems operating procedures identified heating
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requirements and equipment including power supplies, temperature
controls 'and settings, indication circuits, > insulation requirements,
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heat tracing, and space heaters as required.
Backup freeze protection was provided during extended plant.shu'tdown
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in areas that were norma'lly kept warm by heat losses from operational
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' Plant procedures used during maintenance or modification of existing
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systems provided reasonable assurance that cold weather protective
measures were reestablished following completion of those activities.
Plant preventive maintenance requirements associated with cold
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weather preparation were completed.
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The following documents were reviewed:
Preventive Maintenance (PM) routines 01271 and 04047.
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General Operating Procedure (GOP) 2.1.11, " Station Operators Tour,"
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Revision 39, dated October 30, 1985.
Attachment "C" to GOP 2.1.11, " Station Operators Tour - R/W, SOG, ARW
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Areas and Outside."
Attachment "A" to GOP 2.1.11, " Station Operators Tour - Turbine
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Generator Area."
System Operating Procedure (SOP) 2.2.30, " Fire Protection System,"
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Revision 24, dated October 17, 1985.
The discussions, reviews, and walkdowns were performed to verify that the
licensee has maintained a program of cold weather protective measures for
safety-related components and systems.
No violations or deviations were identified in this area.
5.
Preparation for Refueling
The SRI held discussions with fuel handling personnel, observed new fuel
handling activities, and reviewed licensee documents concerning the
receipt and interim storage of new fuel assemblies.
Prior to receipt of
the new fuel, the SRI reviewed the following related procedures for
technical adequacy:
Nuclear Performance Procedure (NPP) 10.21, "Special Nuclear Materials
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Control and Accountability Instructions," Revision 0, dated
October 2, 1984
NPP 10.22, " Receiving and Handling Unirradiated Fuel," Revision 0,
dated October 2, 1984
A total of 152 new fuel bundles were received during January and
February 1986. That fuel will be used for core reload No. 10 which will
be performed during the October 1986 refueling shutdown.
The new fuel was
transported to the site by semitrailer in six shipments.
Shipments one
through four consisted of 26 fuel bundles each and shipments five through
six each contained 24 fuel bundles. The fuel bundle inner shipping
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containers were lifted to the refueling floor for temporary storage
pending future inspections, channeling, and placement in the spent fuel
storage pool.
The SRI observed the receipt and handling of all six shipments and also
performed independent radiation and contamination surveys which included
the transport vehicle and outer shipping containers.
The SRI's
observations also included the general condition of the transport vehicle
and outer shipping containers.
The SRI also verified that shipping
containers were properly sealed and placarded.
The following shipping
papers and licensee documents were also reviewed for completeness and
accuracy:
GE shipping order
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Straight bill of lading
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Preshipment radiation and contamination survey sheets
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GE domestic memo of shipment
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Radioactive materials packaging and shipping record
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GE product quality certification
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Licensee special nuclear material transfer form, Attachment "A"
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NPP 10.21
New fuel and channel handling - material and equipment checklists,
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NPP 10.22, Attachment "A"
Radiation survey of incoming fuel shipment and vehicle, NPP 10.22,
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Attachment "B"
Radiation survey of metal shipping containers, NPP 10.22,
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Attachment "C"
Radiological survey of empty outer shipping containers, NPP 10.22,
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Attachment "D"
Radiological survey of empty fuel truck bed, NPP 10.22,
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Attachment "E"
Fuel storage area survey - stored fuel stack, NPP 10.22,
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Attachment "F"
Background radiation survey - fuel storage area, NPP 10.22,
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Attachment "G"
The reviews and observations were performed to verify that the licensee
used and adhered to approved and technically adequate procedures during
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the conduct of fuel handling and storage activities. The SRI also
verified that documentation of the above activities were complete and
accurate.
No violations or deviations were identified in this area.
6.
Spent Fuel Shipment
The SRI inspected the licensee's activities associated with two shipments.
of spent fuel from CNS.
Included in those inspections were observations
and reviews of applicable procedures, documentation, surveys, inspections,
and shipping document preparation.
The SRI verified by review of licensee documentation, through discussions
with responsible personnel, and by independent inspection that the
licensee completed the following:
Receiving inspection of railcars and shipping casks.
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Shipping documents.
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Advance notification of and approval by affected state and federal
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agencies.
Proper placarding of the transport vehicles.
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Appropriate labeling of the spent fuel shipping casks.
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Establishment of provisions for response by escorts and local law
enforcement agencies.
Training of escort personnel.
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Testing of communications systems.
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Continual manning of the licensee's communications center (Movement
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Testing of fuel and cask handling cranes, hoists, and tools.
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Proper loading and sealing of the spent fuel shipping casks.
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Surveillance of area radiation monitors, ventilation systems, and
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spent fuel pool water level and chemistry.
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Update of fuel location and accountability records.
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Applicable quality assurance audits and inspections.
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U.S.. Department of Energy and U.S. NRC " Nuclear Material Transaction
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Report," DOE /NRC Form 741.
Bill of lading.
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CNS Health Physics Procedure 9.5.3.7, " Cask IF-300 Shirment,"
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Revision 3, dated December 26, 1985.
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CNS Nuclear Performance Procedure 10.27, " Cask IF-300 Handling and
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Shipping," Revision 3, dated November 12, 1985.
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CNS HP-138, " Contamination Survey - Sample Count Data Sheets."
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CNS HP-141, " Contamination Survey - Railroad Car for IF-300
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Irradiated Fuel Shipping Cask."
CNS HP-142, " Contamination Survey of IF-300 Shipping Casks."
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CNS HP-143, " Radiation Survey of IF-300 Shipping Cask."
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CNS HP-608, " Spent Fuel Shipment Checkoff Sheet and Certificate of
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Compliance of Number 9001 Conditions for Shipping Spent Fuel."
CNS HP-14a, " Radioactive Material Shipment Record."
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The following independent radiation and contamination surveys were
performed by the SRI and verified to be satisfactory:
Contact radiation surveys of the shipping casks.
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Radiation surveys at a distance of 2 meters from the cask transport
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vehicles.
Contamination surveys of the shipping casks surfaces.
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Ccntamination surveys of the cask transport vehicles.
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The SRI reviewed CNS Procedure 10.27, Revision 3, dated November 12, 1985.
The licensee incorporated into Procedure 10.27 specific handling
instructions for the GE Type IF-300 spent fuel shipping cask.
Also
included vithin Procedure 10.27 was Attachment "A," " Handling and Loading
of IF-300 Spent Fuel Shipping Cask Checkoff Sheet." The checkoff sheet
provided two functions:
it identified important steps used in the
receipt, inspection, preparation, movement, loading with fuel, leak
testing, decontamination, loading of the cask onto the transport vehicle,
and final preparation for shipping; and it provided a checkoff list
including spaces for signatures and/or initiais of personnel who performed
or witnessed the performance of key steps of the procedure.
The SRI
verified that Attachment "A" of Procedure 10.27 was properly completed,
signed, and dated.
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The spent fuel shipments left the CNS on January 9,1986, and February 18,
1986.
Each shipment consisted of 2 spent fuel shipping casks, each of
which contained 18 spent fuel bundles. The shipments were transported to
the GE Morris Operation Complex, Morris, Illinois.
The spent fuel casks
identification numbers were:
Shipment No. 1 - Casks IF-301 and IF-302
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Shipment No. 2 - Casks IF-302 and IF-304
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The observations, reviews, and independent measurements were conducted to
verify that spent fuel handling and shipment operations were in
conformance with the requirements established in the CNS Operating Li:ense
and Technical Specifications.
No violations or deviat' ions were identified in this area.
7.
Plant Trips - Safety System Challenges
The URC inspectors held discussions with operations shift personnel and
reviewed control room records including log entries, recorder traces, and
computer printouts associated with an unscheduled reactor scram that
occurred on February 27, 1986, at 4:17 a.m.
The reactor was at 69% of
rated power and steady state operating conditions prior to the scram.
On February 25, 1986, a control power fuse was inadvertently blown that
provided power to the "A" reactor feedwater pump (FWP) electrical
overspeed trip circuit. The licensee decided to delay replacement of the
fuse until the development and approval of a Special Procedure (SP)86-005, "RFPT-1A Return to Auto Control," which would provide necessary
guidance and controls to replace the fuse without inadvertently causing
unnecessary feedwater system transients and a possible plant shutdown.
Approximately 5 minutes prior to the scram, operations personnel had
reduced reactor power to 69% in preparation for manually removing the "A"
FWP from service. While reducing "A" FWP speed, the operator noticed that
reactor water level was decreasing and the "B" FWP appeared to respond
more -slowly than expected to increase total feedwater flow.
Before
adequate corrective action could be initiated to correct the low feedwater
flow condition, the reactor scrammed on low reactor water level.
The high
pressure coolant injection (HPCI) and reactor core isolation cooling
(RCIC) pumps started and returned reactor water level to normal.
Both
emergency diesel generators automatically started but were not required to
supply power to their respective vital 4160V AC busses because of the
successful transfer of those busses to their emergency source.
Primary
containment isolation systems groups II, III, VI, and VII actuated as
required by the low reactor water level condition.
No other safety
systems were required to operate.
The licensee made notification of the
event to the NRC.
Following the scram, the plant operators maintained reactor water
inventory using the condensate booster pumps and reactor pressure using
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the main steam bypass system. The reactor was subsequently cooled down
and depressurized to perform maintenance on the HPCI gland steam
condenser, to repair various minor steam leaks and to perform necessary
pre.;tartup surveillance tests.
The plant was restarted and returrad to
power on February 28, 1986.
The NRC inspectors attended a licensee plant management post-shutdown
review meeting conducted on February 27, 1986. The SRI reviewed Station
Operations Review Committee (SORC) meeting minutes No. 458, dated
February 27, 1986. Those meeting minutes provided assurance that the
licensee had thoroughly reviewed the scram and that those reviews
indicated that plant startup could be authorized.
The failure of the "B"
FWP to adequately increase speed and assume required feedwater flow was
attributed to (a) insufficient operator guidance in SP 86-005,
(b) marginal capacity of one FWP to supply sufficient feedwater flow for a
69% power condition, and (c) lower FWP steam supply pressure from the main
turbine moisture separators at other than a 100% main turbine power
condition.
The SRI reviewed the following completed plant procedures that were
performed as a result of the scram:
GOP 2.1.2
Scram Recovery Checklist, Attachment A
GOP 2.1.4 Normal Shutdown from Power, Revision 21
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GOP 2.1.5 Emergency Shutdown from Power, Attachment A
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GOP 2.1.10 Station Power Changes, Revision 9
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The inspections, reviews, discussions and observations were conducted to
verify that the plant responded as designed, plant personnel performed
immediate and followup corrective . actions, and that no unreviewed safety
questions existed. Also, the NRC inspectors verified that facility
operations were in conformance with the requirements established in the
CNS Operating License and Technical Specification.
No violations or deviations were identified in this area.
8.
Operational Safety Verification
The NRC inspectors observed control room operations, instrumentation,
controls, reviewed plant logs and records, conducted discussions with
control room personnel, and performed system walk-downs to verify that:
Minimum shift manning requirements were met.
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Technical Specification requirements were observed.
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Plant operations were conducted using approved procedures.
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Plant logs and records were complete, accurate, and indicative of
actual system conditions and configurations.
System pumps, valves, control switches, and power supply breakers
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were properly aligned.
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Licensee systems lineup procedures / checklists, plant drawings, and
as-built configurations were in agreement.
Instrumentation was accurately displaying process variables and
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protection system status to be within permissible operational limits
for operation.
Plant equipment that was discovered to be inoperable or was removed
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from service for maintenance was properly identified, redundant
equipment was verified to be operable, and applicable limiting
conditions for operation were identified and maintained.
Equipment safety clearance records were complete and indicated that
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affected components were removed from and returned to service in a
correct and approved manner.
Maintenance work requests were initiated for equipment discovered to
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require repair or routine preventive upkeep, appropriate priority was
assigned, and work commenced in a timely manner.
Plant equipment conditions such as cleanliness, leakage, lubrication,
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and cooling water were controlled and adequately maintained.
Areas of the plant were clean, unobstructed, and free of fire
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hazards.
Fire suppression systems and emergency equipment were
maintained in a condition of readiness.
Security measures and radiological controls were adequate.
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The SRI performed a lineup verification of the following systems:
4160V AC Electrical Power Distribution
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Service Water Booster Pumps
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The NRC inspectors witnessed a reactor startup and heatup on February 28,
1986. The reactor achieved criticality at 8:08 a.m. and the main
generator was loaded at 8:00 p.m. on that date.
Those activities followed
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the unscheduled plant shutdown discussed paragraph 7 of this report. The
following areas were observed or verified prior to, during, and following
that startup:
Operable status of required systems
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Completion of required surveillance tests
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Crew shift manning
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Usage of and adherence to approved procedures
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Reactor instrumentation response
Management authorization for startup
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The NRC inspectors observed performance of the following plant procedures:
GOP 2.1.1, " Cold Startup Procedure," Revision 41
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GOP 2.1.1.2, " Technical Specifications Pre-startup Checks,"
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Revision 6
GOP 2.1.3, " Approach to Critical," Revision 9
.
GOP 2.1.15, " Reactor Recirculation Pump Startup and Shutdown,"
.
Revision 15
GOP 2.2.56, " Main Steam and Turbine Bypass System," Revision 19
.
NPP 10.1, "ApRM Calibration," Revision 15
.
The tours, reviews, and observations are conducted to verify that facility
operations were performed in accordance with the requirements established
in the CNS Operating License and Technical Specification.
No violations or deviations were identified in this area.
9.
Monthly Surveillance Observations
The SRI observed Technical Specification required surveillance tests.
Those observations verified that:
Tests were accomplished by qualified personnel in accordance with
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approved procedures.
Procedures conformed to Technical Specification requirements.
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Test prerequisites were completed including conformance with
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applicable limiting conditions for operation, required administrative
approval, and availability of calibrated test equipment.
Test data was reviewed for completeness, accuracy, and conformance
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with established criteria and Technical Specification requirements.
Deficiencies were corrected in a timely manner.
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The system was returned to service.
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The following completed surveillance tests and surveillance
procedures (SP) were reviewed to verify completeress, accuracy,
performance interval requirements, and appropriete management review:
SP 6.1.1, "SRM Functional Test (Reactor Not in Run)," Revision 13
.
SP 6.1.17, "IRM Calibration and Functional Test (Mode Switch Not in
.
Run)," Revision 10
SP 6.3.3.1, "HPCI Test Mode Surveillance Operation," Revision 21
.
SP 6.3.9.3, " Main Steam Isolation Valves Partial Closure Test,"
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Revision 9, dated July 19, 1977.
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SP 6.4.1.2, " Withdrawn Control Rod Operability," Revision 13, dated
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February 13, 1986.
As an additional followup to a previo;s failure of the licensee to
adequately demonstrate operability of station batteries (reference NRC
Report 50-298/84-26 and 50-298/EA 84-132), the SRI conducted interviews
with three electricians and the electrical department supervisor.
The
purpose was to make an up-to-date assessment of the following:
Employee background and experience
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Onsite training; e.g. , formal classroom and on-the-job training (0JT)
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Technical knowledge and understanding of battery technology and CNS
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battery surveillance procedures
The SRI determined that technical knowledge, craftsmanship, previous
experience, and DJT were strong attributes. However, onsite formalized
training of electrical personnel specifically applicable to their trade
and generally in the area of overall plant knowledge was minimal.
The reviews and observations were conducted to verify that facility
surveillance operations were performed in accordance with the requirements
established in the CNS Operating License and Technical Specification.
No violations or deviations were identified in this area.
10. Monthly Maintenance Observation
The SRI observed preventive and corrective maintenance activities on
portions of the following systems components:
Service Water Pumps
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Service Water Booster Pumps
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The coservations were conducted to verify that:
Limiting conditions for operation were met.
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Redundant equipment was operable.
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Equipment was adequately isolated and safety tagged.
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Appropriate administrative approvals were obtained prior to
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commencement of work activities.
Work was performed by qualified personnel in accordance with approved
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procedures.
Radiological controls, cleanliness practices, and appropriate fire
.
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prevention precautions were implemented and maintained.
Quality control checks and postmaintenance surveillance testing were
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performed as required.
Equipment was properly returned to service.
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Those reviews and observations were conducted to verify that facility
maintenance operations were performed in accordance with the requirements
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established in the CNS Operating License and Technical Specification.
No violations or deviations were identified in this area.
11. Exit Meetings
Exit meetings were conducted at the conclusion of each portion of the
inspection. The NRC inspectors summarized the scope and findings of each
inspection segment at those meetings.
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