IR 05000298/1985018

Insp Rept 50-298/85-18 on 850601-0731.No Violation or Deviation Noted.Major Areas Inspected:Operational Safety Verification,Monthly Surveillance & Maint Observations, Licensed Operator Training & IE Info Notices

ML20138C572
Person / Time
Site:	Cooper
Issue date:	10/08/1985
From:	Dubois D, Jaudon J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20138C557	List: IR 05000298/1985018 ML20138C551
References
50-298-85-18, NUDOCS 8510220464
Download: ML20138C572 (29)

Text

V: - .

'

-

y-r

,.

APPENDIX-U. S. NUCLEAR REGULATORY COMMISSION

.4

.

REGION IV

~

NRC Inspection Report: 50-298/85-18 License: DPR-46 Docket: 50-298 Licensee: Nebraska Public Power District- (NPPD)

P. O. Box 499 Columbus, Nebraska 68601 Facility Name: Cooper Nuclear Station (CNS)

Inspection At: Cooper Nuclear Station, Nemaha County, Nebraska -

Inspection Con' ducted: June 1-July 31, 1985 Inspector: U E 7//.7/[$

D. L. DuBois, Senior Resident Inspector, (SRI) Date '-

Other Accompanying Personnel: F. N..Carlson

.

Approved: ( l6/ 'A J./P. paudon/ Chief. Project Section A,

/d [ d Date'

[eactorPrpjectBranch 8510220464 851011 PDR ADOCK 0 % p G

%_.__________________-.__._=__ _---j

'

,

_2_

Inspection Summary Inspection Conducted June 1-July 31,1985 (Report 50-298/85-18)

Areas Inspected: Routine, unannounced inspection of operational safety verification, monthly surveillance and maintenance observations, licensee action on previous inspection findings, survey of licensee's response to selected safety issues, IE information notice, notification of unusual event, safety-related HFA relay replacement, design changes and modifications, facility modifications, tests and experiments program, startup testing--new or modified systems, surveillance--containment local leak rate testing--type B &

C tests, preoperational test procedure review--recirculation system flow control test, plant procedures, licensed operator training, inservice inspection--reactor coolant system hydrostatic test, refueling, and spent fuel pool activities. The inspection involved 410 inspector-hours onsite by one NRC inspector and one consultan Results: Within the 19 areas inspected, no violations or deviations were identified.

l l

t

'

,,

-3-DETAILS 1. Persons Contacted Principal Licensee Personnel

• V. Thomason, Division Manager of Nuclear Operations

• Brungardt, Acting Operations Manager

• Norvell, Maintenance Manager

• M. Mace, Plant Engineering Supervisor

• Ballinger, Reactor Engineering Supervisor

• C. R. Goings, Regulatory Compliance Specialist

• J. L. Peaslee, Surveillance Coordinator

• M. Unruh, Maintenance Planner D. Overbeck, Purchasing / Materials and Accounting Supervisor G. Hansen, Engineering Specialist D. A. Whitman, Technical Staff Manager V. L. 'Wolstenholm, Quality Assurance Manager J. M. Meacham, Technical Manager L. L. Rader, Administrative. Services Manager .

J. Flaherty, Assistant to the Plant Engineering Supervisor J. Scheuerman, Lead Reactor Engineer R. Windham, Emergency Planning . Coordinator R. Black, Acting Operations Supervisor R. Deatz, Engineering Specialist

.

R. Krause, Electrical /I&C Engineer The NRC inspector also interviewed other licensee and contractor personnel including operation _s, maintenance, and administrative personne * Indicates presence at exit meetin . Licensee Action on Previous Inspection Findings

~

(Closed) 8102-03 (0 pen-Item). This item concerned an apparent lack of full controls associated with material requisitioning. The NRC inspector determined that the controls in effect at the time this item was identified placed the responsibility for assuming correct material requisitioning'with each individual worker who withdrew materials from storage for use in the plant. The present controls. for material requisitioning are identified in the following CNS administrative, engineering, and maintenance procedures:

. 1.4, Revision 1 " Requisitioning"

. 1.5, Revision 0, " Receiving"

! _]

'

.

-4-

. 1.6, Revision 0, " Marking and Tagging"

. 1.7, Revision 0, " Storage"

. 1.8, Revision 1, " Warehouse Material Issuance and Return"

. 3.4, Revision 1, " Station Design Changes"

. 7.0.1, Revision 2, " Work Item Tracking-Corrective Maintenance"

. 7.0.2, Revision 0, " Work Item Tracking-Preventative Maintenance" Also, it was noted that the licensee has implemented the Equipment Spare Parts Inventory (ESPI) computer tracking system. The licensee is approx-imately 60% complete with the spare parts inventory and updating the ESPI to reflect that inventor The NRC inspector determined during his review of the above procedures and the ESPI computer tracking system that material requisitioning controls are extensive and consist of many checks and balances that did not previously exist. The responsibility -for assuring' correct material requisitioning no longer rests on one particular individual, but is shared by warehousemen, engineering personnel, and individual workers al ike. Further, the ifcensee is encouraged to complete their materials and spare parts inventory and to update the ESPI computer tracking system in a timely manner. The NRC inspector will review the licensee's progress in this area during future inspection This item is close (Closed) 8404-03 (Unresol ved) . This item concerned the licensee's failure to write Nonconformance Reports (NCR) and/or Maintenance Work Requests (MWR) for plant equipment that was not functioning as designe Specifically, annunciators labeled " Core Spray System 2 Differential Pressure Header Top Of Core" and "Drywell Iodine Activity High" were in an alarm condition. Plant conditions were normal; thus the annunciators were not required to be in the alarm state. .Also identified with this unresolved item was a failure of the high pressure coolant injection (HPCI) gland seal condenser condensate pump to start automatically as required by system condition The present CNS Technical Specification (TS) definition of OPERABLE is,

" Operable means a system or component is capable of performing its intended function in its required manner." No reference is made to attendant instrumentation or support system .

The licensee submitted Proposed Change No.16 to the TS in a letter from Mr. J. M. Pflant (NPPD) to Mr. D. B. Vassallo (NRC-NRR), dated February 22, 1985, in which the _ definition of OPERABLE was changed to I

i

\ .h

'

. 4-5-include all necessary attendant instrumentation. CNS Procedure 2.0.1, Operations Department Policy, Revision 1, includes the proposed definition of OPERABLE which removes any ambiguities concerning this definition from similar annunciator problems that may occur in the future. An MWR-was not initiated for the HPCI condenser condensate pump failure because the pump started before corrective maintenance commenced; however, NCR 002870 was written February.3,1984, to document the occurrenc During his review of this item,- the SRI noted that the licensee had used a surveillance procedure to adjust annunciator actuating switches to within-stated instrument setpoints. The SRI reminded licensee management that the MWR provides necessary management controls, oversite, and historical .information for all plant equipment requiring corrective maintenance. - Although surveillance procedures are good " troubleshooting" mechanisms, they should not be used in lieu of the MWR, but rather be included within the MWR FAIL / WORK Sectio This item is close (Closed) 8501-02 (Violation) . This item concerned the licensee's'

failure to correct a drawing error and to make drawing changes to conform with the current "as-built" configuratio The NRC inspector reviewed the following licensee documentsi

~

. Minor design change (MDC) packages 83-30 and 84-100, including the respective MDC Completion Report . Document change notices (DCN)85-246,'85-256, and 85-257 dated-March 16, March 18, and March 18, 1985, respectively.1 DCN 85-246 applied-to MDC 84-100_ whereas the remaining two DCNs applied to MDC 83-3 The NRC inspector verified that the DCNs were completed, applicable drawings were revised and/or corrected, and the MDC Completion Reports were updated to reflect the issuance of the DCN This item is close (Closed) 8504-04 (0 pen _ item). This item concerned the lack of a formal licensee corporate radiation protection policy statemen The NRC inspector verified that a corporate. radiation protection program

policy statement was issued and that it was transmitted to the Station l

'

Operations Review Committee (SORC) and the NPPD Safety Review and Audit Board (SRAB) in Inter-District Memorandum CNSS850313, dated June la,

, 1985. The policy statement will be included in the next revision of l Health Physics Procedure 9.1.1.1, " Radiation Protection at Cooper Nuclear Station."

This item is closed.

I l

L a

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _

'

.

-6-(Closed) 8504-06 (0 pen item). This item concerned the lack of a formal licensee corporate ALARA policy statemen The NRC inspector verified that a corporate ALARA policy statement was issued and transmitted to the 50RC and SRAB in Inter-District Memorandum CNSS850313 dated June 14, 1985. The policy statement will be included in the next revision of Health Physics Procedure 9.1.1.2, "ALARA Program."

This item is close (Closed) 8513-01 (Violation) . This item concerned the licensee's failure to review the detailed requirements of welding procedure specifications with welders prior to production welding. The NRC inspector reviewed the following licensee corrective actions:

. Formal instructions were presented by construction management to all Greater American Pipe Company (GAPC0) welding crews, including crew supervisory personnel .

. Information copies of all approved welding procedure specifications were posted on the drywell shield wall for easy access by welding Crew . Controlled copies of all procedure specifications were maintained in the drywell .

. Both GAPC0 and Chicago Bridge and Iron Company (CBI) welding and supervisory personnel were instructed as to the location of the welding precedure specifications inside the drywel . GAPC0 welding operators were provided with weld parameter data sheets at each welding console. The data sheets specified initial welding machine settings for each weld joint root consumption operatio . CBI employed additional welding supervisors in order to increase supervisory coverage of welding operation This item is close . Followup of NRC Headquarters Requests - Temporary Instruction (TI)

2515/67, " Survey of Licensee's Response to Selected Safety Issues."

The purpose of this TI was to determine the actions that the licensee has taken with respect to selected safety issues, including mispositioned control rods and steam binding of auxiliary feedwater pumps. These issues were identified in IE Information Notices 83-75 and 84-06 respectively, and were also the subject of Institute of Nuclear Power Operation (INP0) Significant Operating Experience Reports (50ER). The i CNS does not have an auxiliary feedwater system; therefore, the following l information only applies to the area of mispositioned control rod l l

J

'

.

-7-The NRC inspectors conducted interviews with plant reactor engineering, technical staff, and training department personnel, and also performed a review of the following CNS procedures applicable to the issue of

.aispositioned control rods:

. 2.1.1, Revision 38, " Cold Startup Procedure"

. 2.1.1.2, Revision 5, " Technical Specifications Pre-Startup Checks"

. 2.1.2, Revision 21, " Hot Startup Procedure"

. 2.1.3, Revision 9, " Approach to Critical"

. 2.1.4, Revision 19, " Normal Shutdown From Power"

. 2.1.5, Revision 6, " Emergency Shutdown From Power"

. 2.1.10, Revision 8, " Station Power Changes"

. 2.1.19, Revision .6, " Nuclear Fuel Preconditioning"

. 2.2.8, Revision 24, " Control Rod Drive System"

. 2.3.2.27, Revision 11, " Alarms Procedure, Panel 9.5 Annuciator 9-5-1"

. 2.4.1.1.1, Revision ll, " Stuck or Inoperable Control Rod"

. 2.4.1.1.2,' Revision 3, " Uncoupled Control Rod"

. 2.4.1.1.3, Revision 6, " Failure of Drive to Latch"

. 2.4.1.1.7, Revision 8, "CRD floduit Solenoid Valve Failure"

. 2.4.1.3, Revision 8, " Unexplained Increase in Reactor Power"

. 4.2, Revision 7, " Rod Worth liinirrizer"

. 4.3, Revision 10, " Reactor fianut.1 Control"

. 4.4, Revision 6, " Rod Sequence Control System With Group Notch Control"

. 5.2.2, Revision 6, " Inability to Shutdown With Control Rods"

. 10.13, . Revision 12, " Control Rod Sequence and flovement Control"

a

'

, ,.

r -8-1 Other documents reviewed by.the NRC inspectors included:

i

. - General Electric Company's NEDE-21493,- Section 5.5, " Recovery From j an Inadvertent Control Rod Insertion" l

. -IE Information Notice 83-75

, . INPO SOER 84-2 The NRC' inspectors _ verified that the licensee has met all of the requirements of TI 2515/67 listed below:

{ . Procedural requirements relating to' notification and concurrence of

, shif t supervisory, reactor engineering, and management personnel concerning rod control problems or special operational ~ conditions.

,

. Implementation of plant procedures which establish controls,

'

i guidelines, .and conditions for bypassing the rod worth minimizer (RWM).

. Procedural guidance: (1) that prohibits the use of scram timing

>

equipment except for testing and emergencies, and; (2) for the

appropriate use of the " emergency-in" mode of rod insertion and the t

use of the notch override switch for continuous rod withdrawal

,

manipulations.

j

. Prccedure requirements and guidance are in conformance with

Technical Specification requirements.

! . Training was provided in the areas of normal, abnormal, and - '

! emergency movement of control rods, including the consequences.of l _

improper movement and operating with mispositioned rod (s). Training

,

also included the descriptions and functions of the RWH, Rod

Sequence Control System, and the scram test switches.

i

! . CNS has a long-standing policy that requires reactor engineers to be present in the control room during plant startup, shutdown, and special operational conditions. Reactor engineers are promptly notified, whether they are on or offsite, of rod control problems or operational abnormalities.

i

. Procedures. address recovery from mispositioned control' rods and verification of rod position when the normal position ~ indication system is lost.

This item is closed.

.i

!

i I

-

__ d

'

.

--9-4. IE Information Notice Followup IE Information Notice 85-17, "Possible Sticking of ASCO Solenoid Valves," dated March 1,1985, discussed a failure of the- ASCO type HTX 8323-20V solenoid valve used to provide main steam isolation valve (MSIV)

fast closure. The SRI held discussions with the plant engineering supervisor and maintenance planner concerning this item and concluded

"lat CNS does not use the specified ASCO solenoid valve in the CNS MSIV control system. The CNS used AVC0 solenoid valves for this applicatio . Notification of Unusual Event 4 At 5.55 a.m. on July 19, 1985, the CNS emergency director informed the SRI that all meteorlogical (MET) instrumentation had been rendered inoperable due to a loss of power to both 10-meter and 100-meter MET towers. CNS Emergency Plan Implementation Procedure (EPIP) 5.7.1,

_

Attachment "C," " Classification Guide," Section 6.1, requires- the licensee to classify and initiate a Notification of Unusual Event (NOVE)

when a significant loss of meteorological assessment capability occurs (e.g., a complete loss of meteorological instrumentation).

The MET system lost power at approximately 4:53 a.m. on July 19, 1985, due to a lightning hit on the MET power supply line resulting in a blown main line power supply fuse. Normally, there are two independent power lines feeding the MET system but only one temporary supply was availabic at the time of this event. All MET permanent power supply and instrumentation cables were being rerouted as directed by MDC 85-4 MDC 85-47 was developed, approved, and implemented to reroute applicable cables from beneath the area where the licensee is planning to construct a new training building. Upon completion of MDC 85-47, all MET cables will be contained in a rigid concrete ductbank which will pass around the site of the proposed new training facilit The SRI was onsite during this occurance and verified that licensee actions were performed as required by plant emergency procedures. Also, the SRI observed that the licensee terminated refueling operations as a precautionary measure during the event. The NOUE was terminated by the emergency director at 10:50 a.m. on July 19, 1985, following restoration of the MET power supply and verification that all instrumantation was restored and operating in a stable manne The SRI performed a review of the fcilowing licensee logs, procedures, and reports:

. Control room logs for complete at d timely entry of significant information pertaining to the evel . Procedures and checklists applicable to the declaration and termination of the event.

t ,

_ -

,

-U

'

.

..

.

,;

.

,

'

,.

-10-

. Licensee followup report to the'NRC -in a letter-from Mr. P.- Thomason (NPPD) to Mr. R. D. Martin (NRC RIV) dated July 19, 198 . 5.7.1 Rev. 4, " Emergency Classification" ,

. 5.7.2 Rev. 4, " Notification of Unusual Event"

. 5.7.6 Rev. 5, " Notification"

. 5.7.22 Rev. 6, " Communications" 5.7.28 Rev. 1, " Emergency' Director"

.. ' MDC 85-47, " Underground Cable Reroute Around Training Center" The SRI held the above discussions and performed initial and followup reviews and determined that licensee personnel performed all actions required by the CNS emergency procedures, TS and Emergency Pla No violations or deviations were identified in this are . Safety-Related HFA Relay Replacement IE Bulletin (IEB) 84-02 was issued March 12, 1984, as a result of an increase failure rate of General Electric (GE) type HFA relays. One requirement stated in IEB 84-02 was that all holders of operating licenses were to replace HFA relay coils or entire relays within 2 years from the date of the bulleti The SRI ~ verified that all GE type HFA 120V AC relays used in safety-related applications were replaced and satisfactorily tested during the'present outage. The replacement,was accomplished by implementation of MDC 83-6 . Design Changes and Modifications

~

The NRC inspector conducted a followup review of the design ch'anges and modifications inat were initially inspected and reported in NRC Inspec-tion Report- 60-298/85-16. The review included 20 packages consisting of-16 minor design changes (MDC), 2 special' procedures (SP), and 2 special test pro'.edures (STP).

This review-verified that:

. Design change packages and' test procedures were approved-and implemented in accordance with program ^ requirements,. established procedures, and Technical Specification requirement ~

. Appropriate-QA/QC hold points, reviews, and approvals were included in packages as require ' '

- .

,.

)

.

, u

,

,J

f~

c-11-

. Special hazards analysis were performed and precautions and controls were specified when necessar . Personnel involved with the development, review, approval, implementation, and inspection of completed design changes and test procedures were qualified and knowledgeabl . Amendments to or revision of original approved documents were accomplished in accordance with established procedures and received appropriate SORC review and approval .

. Postmodification testing was sufficient to demonstrate system / component operabilit . Acceptance test deviations were resolved and necessary retesting was accomplished when require . Documents affected by an HDC such as vendor technical manuals, as-built drawings, control room prints, operating procedures, and surveillance tests, were temporarily revised and/or adequately identified as undergoing revisio . When an MDC or test package was completed, a Design Change Completion Report, Attachment "G" to CNS Engineering Procedure 3.4, or an SP/STP Completion Report, Attachment B" to CNS Engineering Procedure 3.5, was completed and attached to the package. The completion reports provided a description of completed work or testing and indicated completion of administrative requirements such as the document revisions indicated above, final engineering disposition, and management review and approval .

. In some cases, a completion report was. not administratively closed because special tests or operational checks could not be performed due to the shutdown condition of the plant. In those cases, the NRC inspector verified that the packages contained a status report that served to document completion of the package to the point of readiness for special testing or operational checks. The status report also noted the documents that were in the process of final printin . By use of completion reports and status reports, licensee site / corporate safety committees and senior management determined the readiness of the affected component / system readiness for plant startup. Subsequent to this reporting period, the SRI observed special tests, plant startup, and systems testing to insure that the licensee adequately proved operability of those affected systems prior to declaring them fully tested and operational.

i I

.

I

'

.

-12-

. Operator training was accomplished for major work items or MDCs in either a formalized classroom environment or self study situation as determined by the significance of the work effort or plant

~

modificatio . MDCs were placed on 'a tracking system that would insure their inclusion in the annual report to the NR The adequacy of the licensee's program and controls of temporary modiff-cations, lifted leads, and jumpers was discussed in NRC Inspection Report 50-298/85-15 and resulted in a violation (8515-01). This area will be the subject of a future NRC inspectio No additional violations or deviations were identified in this are . Facility Modifications The NRC inspector reviewed facility mod' fications directly associated with the replacement of reactor recirculation system piping affected by intergranular stress corrosion cracking (IGSCC). This review specifi-cally covered the adequacy of the associated modification packages, including their content, control of work activities, test, and acceptance requirement The following MDCs, associated with the piping replacement modification, were reviewed:

MDC.84-150, "IGSCC Piping Replacement" MDC 84-150 A, " Removal of Interferences for IGSCC" MDC 84-150 B, " Reactor Building Interior Wall Penetrations" MDC 84-150 C, " Jet Pump Instrument Small Bore Piping Reroute" MDC 84-225, " Welded Attachments to the Drywell for Piping Replacement" MDC 84-150, Amendment 1, "Reinsta11ation of Interferences for IGSCC Piping Replacement" The NRC inspectcr's review verified that:

. Applicable . industry codes, standards, regulatory guides were addresse . Work activities were performed in accordance with applicable governing procedures, k a

'

.

-13-

. All HDC stIps completed to date were reviewed and found acceptabl . Permanent changes to plant procedures, vendors manuals, Updated Safety Analysis Report (USAR), TS and training materials were identified and in progres NRC inspector observations of work and nondestructive examination (NDE)

activities and reviews of related documentation were conducted and documented in previous NRC inspection report Observations of test activities and a review of completed test records will be performed by the SRI during his next inspection perio These reviews and observations were conducted to ensure that f acility modifications were approved by the NRC as required; and that the modifications were completed in accordance with license requirements, licensee commitments, TS,10 CFR, and applicable industry codes and standard No violations of deviations were identified in this are . Tests and Experiments Program The NRC inspector reviewed the licensee's program and implementation procedures in the area of tests and experiments (T&E) controls. The NRC inspector verified that:

. A formal system was approved and used to prepare, review, approve, and conduct necessary T&E. Provisions were included which assured that responsibilities for the conduct of T&E were assigned and that T&E engineers were qualified to perform and review tests. The USAR was reviewed to determine if the test was described, and if not, that a 10 CFR 50.59 safety evaluation was conducted to insure that the T&E did not involve an unreviewed safety question or change to the Technical Specification. Provisions were established for reporting required T&E to the NRC on a scheduled basi . The T&E program was implemented by approved procedures. T&E results were fully documented, reviewed, and approve . The licensee's QA program requires periodic audits of SP and STP The audits are performed using QA checklists which provide satisfactory review of T&E program status and adequac The NRC inspector reviewed the following SPs and STPs:

. SP 83-06, " Hydrostatic Testing of New Fire Protection System"

. SP 83-12. " Removal of Board D From Control Room" l

l

,

l

. J

.

-14-

. SP 83-13, " Visual Examination of Drywell Mechanical Snubbers"

. SP 83-18, " Repair Weld of RR Loop "A" 4" Bypass Base Metal"

. SP 84-02, "RHR-SW Shutdown Cooling Operation Without SWBP's Operating"

. SP 84-09, " Instrumentation of RPV Annulus Level"

. SP 84-10, " Installation of RPV Shroud Level Indication"

. ~SP 84-17, " Post-Decentamination Flush"

. STP 83-04, "Testino of HFA & CR120 Relays in the Safety Systems"

. STP 83-09, " Fuel Inspection"

. STP 83-10, " Pressurizing Yarway NBI-LIS-83A to Reactor Pressure"

. STP 83-20, "TEC Pump Mechanical Seal"

. STP 84-07, " Scram Valve Reset Testing"

. STP _84-09, " Identification, Location, and Inspection of Terminal Boxes in Reactor Building"

. STP 84-10, VOIDED

. STP 84-11, " Pressure Switch FP-PS-651F Operability Test: Remote

. Pressure Drop Simulation" The NRC inspector noted that, of the 16 procedures reviewed, I was voided,-

11 were closed out, including completion reports and SORC reviews, and the remaining were open pending satisfactory completion of preoperational tests and the compilation of required documentation. The following CNS j procedures were reviewed by the NRC inspector in this area:

. 3.3 Rev.1, " Station Safety Evaluations"

. 3.5 Rev. O, "Special Test Procedures /Special Procedures"

. QAP-1700 Rev. 6, " Quality Assurance Plan-Design Changes (Major and Minor)"

. SC-1700-4 Rev. O, QA Surveillance Checklist, "Special Test Procedures /Special Procedures"

. SC-1700-6 Rev. O, QA Surveillance Checklist, "STP/SP Completion Review"

. SC-1700-7 Rev. O, QA Surveillance Checklist, "STP/SP In-Progress Review" J

'

w

, 1 '

,

(

-15-

,

These reviews were conducted to ensure that' the licensee was implementing a QA p'rogram relating to the control of tests and experiments that conforms _with regulatory requirements, . licensee commitments, and industry guides and standard No violations or deviations:were identified in this are .- 'Startup Testing-New or Modified Systems *

The NRC inspector reviewed the licensee's startup testing program and observed initial systems startup activities to ascertain that the program conformed to the regulatory requirements of ANSI U18.7-1976 and 10 CFR 50, Appendix B, and that_ the tests were performed in accordance with approved pl. ant procedures and administrative controls. Test procedures were reviewed and verified to contain.the following infor-mation:

. Appropriate management reviews and approva . Scope and objective . References to the TS drawings, codes and standard . Precautions, limitations, setpoints, and acceptance criteri . Provisions for recording general comments concerning the conduct of the test and for providing details such as test data, deficiencies, resolution of problems or deviations, and retest informatio . Prerequisites were identifie . . Modifications to a component / system were installed and removed within the body of the test procedur . Provision for the identification of test personnel, including appropriate acceptance and approval sign-offs. Spaces were provided to' log required test dat . QA/QC hold points were established as appr priat .- Temporary or permanent choses were accomplished as. required by plant administrative procedures, including safety committee review .

Test results met established acceptance c,riteria.-

. Test deviations were resolved and further testing accomplished as appropriate.

!

s

.

' -

( -

,

s

T

'

.

-10 .

The startup test program is still in progress and will continue to be monitored by the SRI during the next report perio The NRC inspector observed the startup program to be well managed, organized, scheduled, and controlled. Specific test procedure reviews and observations conducted by the NRC inspector are noted in various paragraphs within this report. The startup program was still in progress at the end of this reporting period and will continue to be monitored by the SR '

,

.

No violations or deviations were identified in this are . Surveillance--Containment Local Leak Rate Testing--Type B&C Tests '

The NRC inspectors reviewed the following licensee procedures that were used for controlling local leak rate testing (LLRT) of primary and secondary containments pentrations: ,

'

. 6.3.1.1 Revision 16, " Primary Containment Local Leakage Tests"'

. 6.3.10.8 Revision 5, " Secondary Containment Leak Test" None 'of the MSIVs failed initial LLRT. The NRC inspector observed portions of the "B" outboard MSIV stem replacement and reviewed LLRT test data

' following that maintenance. Other LLRT test data reviewed included:

. RWCU-H0-15

. RWCU-M0-18

. Primary containment personnel airlock outdoor *

. Primary containment control rod drive equipment hatch *

. Secondary containment leak test *

• The performances of these tests were observed by the NRC inspecto The NRC inspector's reviews of the above procedures and tests verified that:

. .LLRT procedural controls were adequat . TS requirements were identifie . Calculation methodology was included in the procedures as require . Testing was accomplished according to procedure and test results were complete, accurate, and received adequate management review and approval.

l

f ( .

"

. '

.

'

,. ,

i-17-

. Plant QA personnel performed independent' checks o'f the leak rate calculations and basic calculational methodelogie ,

. Testing was performed and reviewed by qualified personne . Penetration isolation valves were restored to their correct position and independently verified following testin These reviews,and observations were performed to verify that type B&C LLRTs were performed in accordance with the TS, facility procedures, and industry standard No violations or deviations were identifie '

12. Preoperational Test Procedure Review--Recirculation Flow Control System -

The NRC inspector reviewed CNS Special Test Procedure 85-15,

" Recirculation Flow Control System," dated April 10, 1985, which was approved on June 24, 1985. The NRC_ inspector verified that the procedure contained the following information:

. Appropriate reviews and approval's as required by the CNS controlling procedure 3.5, "Special Test Procedures /Special Procedures."

. A 10 CFR 50.59 Reportability Analysi . Test objective . References including the USAR, TS, CNS Operating Manual, and appi fcable drawing ~

. Precautions and limitation . Test equipmen . Prerequisite . Acceptance criteri . Step-by-step procedural steps including' hold points-as require Other systems to be tested in conjunction with a particular -

procedure were also addresse . Appropriate data sheet .

'

. Test completion checklis '

-

- .

~ , .. .

. The procedure was reviewed against the requirements noted in "

Regulatory Guide 1.68, " Initial' Test Programs!for Water-Cooldd

_ , Nuclear Power Plants," and- found to 'adequatelyfaddress, those , ,

,

requirement '

'

! -

'

, ,

, .."* v $

J e

'

& b

_

_

.

-18-The ifcensee is scheduled to begin plant startup and implementation of a formal power ascension test program in August 1985. Prior to startup, preoperational tests will be performed on those systems which have not been in service during the outage, have undergone substantial modification, or are located in an area where major work activities occurred. The SRI is presently reviewing the preoperational test schedule, procedures, and power ascension program. The SRI will observe

~t he performance of various tests prior to startup and during the power ascension progra These reviews were conducted to verify that licensee procedures, associated with components / systems preoperational tests, were consistent with the requirements established in the CNS Operating License, TS, and regulatory guidanc No violations or deviations were identifie . Plant Procedures The NRC inspector reviewed the licensee's system for procedure control and performed a review of selected procedures to assure conformance with the requirements of.the TS, ANSI 18.7--1976, Regulatory Guide 1.33, and 10 CFR 50.59 parts a and b. Forty-three procedures were selected for review from the following areas:

. Administratio . General plant operation . Individual system operation '

. Annunciator respons . Abnormal and emergency condition . Maintenanc . Surveillanc Through his review, the NRC inspector determined that:

. The licensee had a viable system for developing, reviewing, approving, and controlling plant procedures and subsequent temporary or permanent revisions. The controlling system also containe provisions for meeting record storage and retention requirement . Procedure changes reflected current revision to the TS.

l l

i

!

- - - _ .

- J

.

-19-

. New procedure development or revision met 10 CFR 50.59 parts a and b requirements, including appropriate level of site and/or corporate management revie . Selected procedures were comparabic to those listed in ANSI 1 Overall procedure content was also simila . Complex procedures contained checklists to assure a step-by-step progression or to note hold points or data entry condition Checklists contained provisions for adequate management level review and approva . Technical Specification requirements were included in the reference i srction and/or within the procedure body to preclude operating the sistem(s) outside of established license conditions or limiting ccrditions for operation. Notes or warning statements were included in the body of the procedures to alert person (s) performing the steps of possible sensitive areas of operation or limiting equipment or license condition . Procedures appeared to provide sufficient direction to accomplish necessary evolution . Instructions exist for exercising or rotating operational equipment which are normally idle and are not run frequently during surveillance testin . Documents are distributed according to current distribution lists, guidance is available prescribing the documents use in performing an activity, and controls specify usage of only the most recently dated or revised cop , Temporary procedure changes were initiated and approved according to approved administrative procedures. Also, the temporary changes did not change the " intent" of the original procedure. . Additionally, permanent and temporary procedure revisions were distributed in a timely manner to' operating areas of the plant. Temporary revisions were changed to permanent procedure revisions in the time required by administrative procedure The NRC -inspector randomly selected procedures from the categories listed above and. verified that the current revisions of. the procedures were located in the master file and in the operating areas. Weaknesses have been identified previously in this area and are presently being tracked as NRC open items pending-further review. The following is a list of applicable open items:

i . Open Item, 8326-01, " Procedures taking exception to 10 CFR 50.59 requirements."

l

.

^

. _ - . .. . - - __ . . ._ _ . _ . -

~

.

-

, [

-20-

,

+

. Violation , 8411-01, " Failure to have procedures for maintenance of safety-related equipment." {

'

. Viol ation , 8515-03, " Failure to have procedures 'for activites !

affecting quality." l

. Violation, 8519-01, " Lack of procedures for controlling, voided, deleted, or superseded drawings."

'

In NRC Inspection Report 50-298/84-21, NRC inspectors documented various'

inadequacies of CNS Procedures. The licensee responded to that report in

.

a letter from Mr. L. G. Kunc1 (NPPD) to Mr. J. N. Grace (NRC), dated April 9,1985. The response stated that an independent consultant, KHC, i INC., had been contracted to assess station procedures with regard to i compliance with regulations and. industry codes and standards. The  ;

assessment was to include content, control of, usability, and forma l j The assessment was expected to be completed by June 1,198 j

4 The SRI verified that the assessment of CNS procedures was complete i

The consultant reported its findings in a letter to Mr. P. V. Thomason, i J

dated May 24, 1985. The licensee is presently evaluating.those finding ;

The SRI will followup licensee actions in this area during a subsequent i

inspection.

.

.

The.,e reviews and observations were performed to verify the technical adequacy of licensee procedures and to verify that the procedures .,

conformed with the TS, Operating License, and regulatory guidanc j i

No additional violations or deviations were identified in this are t i .

14. Licens6d Operator Training I

'

The NRC inspector reviewed licensed operator training in the area of facility design changes which were implemented during the present outag .This review was conducted in conjunction with his reviews of the facility !

- design changes and modifications discussed in paragraphs 7 and 8 of this !

inspection report. 'The purpose of this review was to determine if the -!

licensee had a viable system for ensuring licensed operators received ,

adequate and timely training in this are !

CNS Procedure 3.4, " Station' Design Changes," Revision 1,-is the f

. controlling document for ensuring, in part, that licensed operators are -

made aware of facility changes and modifications. Procedure 3.4,.

Section'IV.A., steps 12 and 13 placed the operations and training groups into the design change review and approval process which alerted those groups to the impending plant changes. Also, Attachment "G," " Station

- Design Change Completion Report,"' provided the final management' review,

-

/

_

.

I-21-approval, and closecut of the MDC, which included the receipt of the updated and formally printed copies of affected drawings, procedures, and other documentation. The completion report was then distributed to various plant and corporate departments including the training and operations departments. Once the various departments received the completion report, they assummed the responsibility for insuring this information and approved prints and drawings were distributed and training was conducted as require l The NRC inspector found that the above program was being follwed. Each MDC completion report was reviewed for complexity and significance and was then assigned to be placed into operator requalification training sessions  :

and/or placed into the operators' required reading list, whichever was  !

appropriate. The inspector randomly selected many of the MDCs performed  !

during the present outage and determined that adequate formalized training was conducted as required. The required reading list was also sampled, and it was determined that assigned reading was being accomplished and properly l documented. Additionally, MDCs were observed to be filed in the control l room so they were readily available to operations personnel .

The NRC inspector noted that plant personnel were not being trained or in receipt of the most up-to-date procedures and drawings affected by the MDCs because the MDC completion report was not distributed, with attached up-to-date documentation, until it was officially closed ou In many cases, postmodification testing could not be accomplished until plant conditions, or the time required to formally revise, print, and ,

distribute affected procedures and drawings delayed the final closecut of l the completion report. In order to expedite interim procedures, drawings, and other important documents being provided to the training and operations groups, the licensee was in the process of revising Procedure "L," Proposed Revision 2 to Procedure 3.4 would implement an attachment

" Design Change. Status Report," which would require that the training l and operations supervisors be notified that MDC work was complete l Interim drawings, procedures, and Training Manual revisions would be l

.

l attached to that status repor Proposed Revision ? to Procedure 3.4 was l l

reviewed by the NRC inspector and he determined that the revision would l l

excedite the transmittal of MDC related documentation to the training and l

!

operation groups in a more timely manne l These reviews were performed to assure that licensed operators were l

i l trained in the area of facility modifications as required by 10 CFR 55,

' - Appendix A, and regulatory guides and standard !

l No violations or deviations were identified in this area.

l 1 Inservice Inspection--Reactor Coolant System Hydrostatic Test The SRI performed procedural _ reviews, performed independent inspections, and held discussions with lead test engineers and operations personnel  ;

!

!

---__. _ ._.. ..._. ._._ _.-. ._. _ .._____

. .

,

-22-concerning a planned hydrostatic test of the reactor coolant system. His review encompassed the following plant procedures:

. Maintenance Procedure 7.0.8, Revision 0, " Hydrostatic Leak fest"

. Special Test Procedure 85-1L, Revision 0, "ASME Class 1 Hydrostatic Test" The SRI noted that Procedure 85-1L was developed and approved to replace Procedure 7.0.8 for this specific test. Procedure 85-1L scope and inspection effort was much larger than normally experienced during previous hydrostatic tests because the complete reactor coolant system piping was replaced during the present outage. Also, provisions were made to hydrostatically test portions of other systems that are directly attached to the reactor coolant system and of which new replacement piping was also installed. Additionally, this STP was to fulfill postmodification tests requirements associated with numerous MDCs, STPs, and SP The SRI included STP 85-1L in his review of the new or modified systems startup testing program discussed in paragraph 10 of this repor Procedure 7.0.8 testing criteria was incorporated into STP 85-1L. No procedure descrepancies were identifie The SRI performed inspections and independent systems lineups prior to commencement of the hydro including:

. Reactor recirculation system valve lineu . Hydro pressure boundary valve position verifications of various systems including the reactor water cleanup, residual heat removal, core spray, main steam, control rod drive, reactor feedwater, high pressure coolant injection, reactor core isolation cooling, and main stea . The primary containment buildin . License preparations for the hydro, including .the completion of systems walkdowns, ready identification of the test boundary on the control room panels, test prerequisites, and test equipment calibration and installatio These reviews, observations, independent verifications, and discussions were performed to ensure the licensee's readiness to perform the hydrostatic test and that the test would meet requirements established in the Technical Specification, CNS Operating License, and the ASME Boiler and Pressure Vessel Code Section XI,1974 editio No violations or deviations were identified in this area.

L a

'

.

-23-16. Refueling Activities The SRI observed fuel handling activities and reviewed licensee documentation associated with the preparation, performance, and verification of actions required for refueling operations. The SRI verified the following:

. Surveillance testing was performed prior to and periodically during i fuel handling operation . Secondary containment integrity was maintaine . Precautions were taken t.o prevent foreign. objects from falling into the reactor vessel and refueling cavity. - A control point log was established and maintained to provide accountability of all tools and equipment used above the refueling cavity and reactor vesse . Good housekeeping practices were maintained in the refueling are '

. Shift manning responsibilities and requirements were me . . Continuous communications were maintained between the refueling bridge, refueling area, and the control roo . Radiological precautions were observe '

. - Fuel accountability status boards and records were maintained during refueling operation . The reactor conirol mode selector switch remained in the REFUEL position.-

. Con' trol room operators monitored nuclear instrumentation and other core parameters for verification'of reactor condition . Fuel- bundles, single and double blade guides, and control rods, were positioned'in accordance with established procedures.-

. Liceasee management verified the correct core locations of all fuel

. assemblies fo11'owing the: completion of refueling activitie The SRI reviewed .the following refueling related procedures:

. 2.L.31, Revision 7, '" Fuel Handling-Refueling Platform"

. ~10.25,L Revision .3,'" Refueling"

. 10.26, Revision 0, " Working Over or in Reactor Vessel Requirements"

. 5.3.5, Revision 5, " Refueling Floor High Radiation"

-

.

-24-The SRI observed preparations for and the conduct of fuel loading operations during the period of July 17-24, 1985. Following the completion of fuel loading, he performed an _ independent verification of the fuel assemblies locations in the reactor vessel. Also, he verified that' licensee fuel accountability records were in agreement with the approved design fuel loading plan and the final core configuratio The reviews and observations were conducted to verify that refueling activities were performed in accordance with the requirements in the CNS Operating License and T No violations or deviations were identified in this are . Spent ~ Fuel Pool Activities-The SRI observed that'the following TS requirements were met during fuel handling activities in the spent fuel pool ~(SFP):

. SFP water level anil cooling system operatio . SFP ventilatio . Secondary containment integrit . Reactor building negative pressure.-

. Refueling floor radiation monitoring syste . SFP activities were supervised and conducted by qualified personnel in accordance with approved procedure . Fue11 assemblies location status boards and accountability records were properly documented and maintaine , On July' 24, 1985, the SRI'was present in the control room while fuel

' bundle dechanneling evolutions were being conducted in the SFP. During '

' that period, ~the refueling floor supervisor reported that a dechanneled

-

Efuel assembly was'iin the, fully raised position and was not engaged by the

! '

grappl e. Rather, the fuel assembly bail was caught and held in place by

~

an outside ccrner ~of the fuel grapple head. Movement of the affected

! ' fuel assembly was' stopped pending the completion 'of the following j . precautionary measures:

l

. The refueling floor and entire reactor building was cleared of all unneccessary personnel, and personnel accountability was performed and maintained.

L . The ~ point of contact between the' fuel assembly and fuel grapple head

was ' observed with binoculars and-was also video taped.,

.

!-

'

a

-

)

-25-

. Since the fuel assembly was already fully supported by the fuel grapple head assembly in the full up position and no alternate method for grabbing and supporting the fuel bundle could be devised, the decision was made to lower the assembly down into the SFP storage position located directly beneath the fuel bundle. The fuel assembly was subsequently lowered into that storage position without inciden The controlling procedure for the dechanneling evolution,10.25,

" Refueling," was immediately reviewed to determine if a temporary change should be made that would aid in preventing further occurances of this kind. As a result, an additional precaution was added as step Q to Procedure 10.25,Section VI, which stated, " Visual confirmation must be made that the grapple is properly engaged prior.to raising a fuel assembly."

The licensee initiated NCR No. 004688, dated July 24, 1985, to document this occurance. The SRI will followup further licensee actions and reviews in this area during a subsequent inspectio The reviews and observations were conducted to ver.ify that SFP actitities were performed in accordance with the requirements in the CNS Operating License and T No violations er deviations were identified in this are . Operational Safety Verification The SRI observed control room operations, instrumentation, controls, reviewed plant logs and records, conducted discussions with control room operators, and performed system walk-downs to verify that:

. Minimum shift manning requirements were me . TS requirements were observe . Plant operations were conducted using approved procedure . Plant logs and records were complete, accurate, and indicative of actual system conditions and configuration . System pumps, valves, control switches, and power supply breakers were properly aligne . Licensee systems lineup procedures / checklists, plant drawings, and as-built configurations were in agreemen . Instrumentation was accurately displaying process variables and protection system status to be within permissible operational limits

for operation.

[ 7,

!

'

~

T

- d

,

-26-

. Plant equipment that was discovered to be inoperable or was removed from service for maintenance was properly identified, redundant equipment was verified to be operable, and applicable limiting conditions for operation were identified and maintaine . Equipment safety clearance records were complete and indicated that affected components were removed from and returned to service in a correct and approved manne . Maintenance work requests were initiated for equipment discovered to

~

require repair or routine preventive upkeep, appropriate priority was assigned, and work commenced in a timely manne ~

.

. Plant equipment conditions such as cleanliness, leakage, lubrication, and cooling water were controlled and adequately maintaine . Areas of the plant were clean, unobstructed, and free of fire hazards. Fire suppression systems and emergency equipment were maintained in a condition of readines ~ ~

. Security measures and radiological controls were adequat The SRI performed lineup verifications of those portions of the following systems which are located in the drywell and steam tunnel areas:*

. Reactor Recirculation

. High Pressure Coolant Injection

. Reactor Core Isolation Cooling

. Core Spray

'

. Residual Heat Removal

. Main Steam

. Reactor Water Cleanup

. Reactor Head Vents

. Standby Liquid Control

~

• These areas are High Radiation Areas during power operation and are nonnally inaccessable.

l l

I t

A

.

-27-The tours, reviews, and observations were conducted to verify that facility operations were performed in accordance with the requirements established in the CNS Operating License and T No violations or deviations were identified in this are . Monthly Surveillance Observations The SRI observed TS required surveillance tests. These observations verified that:

. Tests were accomplished by qualified personnel in accordance with approved procedure . Procedures conformed to TS requirement . Test prerequisites were completed including conformance with applicable limiting conditions. for operation, required administrative approval, and availability of calibrated test equipmen . Test data was reviewed for completeness, accuracy, and conformance with established criteria and TS requirement . Deficiencies were corrected in a timely manne . The system was ' returned to servic The following surveillance . tests were selected and observed:

2.1.18, " Control Rod Drive Friction Test" 6.1.2, "IRM Functional Test (Mode Switch Not In Run)"

6.1.4, " Main Steam Line Process Radiation Monitor Calibration and Functional Test" 6.1.21, "SPJi Calibration and Functional Test (Reactor Not In Run)"

6.1.27, Attachment "D," " Refueling Equipment Test"

"E," "SPJ4 Alarm and Response Check"

"F," " Refueling Requirements Test" 6.1.33, " Backup Scram Valves Functional Test" 6.2.4.1, " Daily Surveillance (Technical Specifications)"

6.2.7.1, " Mech'anical Vacuum Pump Isolation Logic-Functional Test" l

-

.

-28-6,3.10.7, " Primary Containment Isolation Valve Closure Timing" 6.3.10.13, " North and South SDV Vent and Drain Yalves Cycling, Open Verification, and Timing Test" 6.4.5.3, " Fire Pump Weekly Operability Test" 6.4.6.3, " Control Room Vent Monitors Calibration and Functional / Functional and Logic Tests" 6.4.8.16, " Dose Assessment Pocket Computer Functional Test" 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 T No violations or deviations were identified in this are . Ilonthly Maintenance Observation The SRI observed preventive.and corrective maintenance activities on portions of the following systems / components:

. 11SIVs

. Drywell Piping Insulation

. Reactor Water Cleanup

. Service Water Booster Pumps

. Reactor Equipment Cooling Pumps

. 480V AC Breakers ,

. Station 125V DC Batteries

. Primary Containment Systems Restorations

.

The observations were conducted to verify that:

, . Limiting conditions for operation were met.

. Redundant equipment was operabl . Equipment was adequately isolated and safety tagged.

! . Appropriate administrative approvals were obtained prior to

'

comencement of work activities.

'

.

,

..

-29-

. Work was performed by qualified personnel in accordance with approved procedure . Radiological controls, cleanliness practices, and appropriate fire prevention precautions were implemented and maintaine . Quality control checks and postmaintenance surveillance testing were performed as require . Equipment was properly returned to servic These reviews and observations were conducted to verify that facility maintenance operations were performed in accordance with the requirements established in the CNS Operating License and T No violations or deviations were identified in this are . Exit !!eetings Exit meetings were conducted at the c~onclusion of each portion of the inspection. The NRC inspector summarized the scope and findings of each inspection segment at those meetings. .- J