IR 05000298/1984010

IE Insp Rept 50-298/84-10 on 840401-30.No Violations or Deviations Identified.Major Areas Inspected:Operational Safety Verifications,Monthly Surveillance & Maint Observations,Ler Followup & Plant Trips

ML20197G937
Person / Time
Site:	Cooper
Issue date:	05/24/1984
From:	Carpenter D, Dubois D, Jaudon J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20197G934	List: IR 05000298/1984010 ML20197G928
References
50-298-84-10, NUDOCS 8406180216
Download: ML20197G937 (14)

Text

. .

APPENDIX U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

NRC Inspection Report: 50-298/84-10

Docket: 50-298 '

License: DPR-46 Licensee: Nebraska Public Power District (NP D) /

P. O. Box 499 3 Columbus, Nebraska 68601

'

Facility Name: Cooper Nuclear Station (CNS),

I Inspection At: Cooper Nuclear Station, Nemaha County, Nebraska Inspection Conducted: April 1-30,1984 Inspectors:

lli k ^ ch C/79 /N Date o

D. Q. DuBois, Senior Resident Inspector (SRI)

41u n we f/24/iM hg .' ' Carpenter, Resident Inspector (RI) Date

Approved:

h ?%b J. Pl./Jaudon, Chief, Project Section A, s7n/s Date Rdactor Project Branch 1

Inspe,ction Summary Inspection Conducted April 1-30, 1984 (Report 50-298/et R )_

Aree; Inspected: Routine, announced inspection of operational safety verifica-tion , monthly surveillance and maintenance observations, licensee event followup, planc trips - safety system challenges, declaration of unvaual event, independent

~

inspection effort, and followup of previously identified nems. The inspection involved 121 inspection hours onsite by two NRC inspector l Results : Within the areas inspected, no violations or M/f ations were identifie PDR ADOCK 05000298 G PDR

- . -.

. . - . _ _ . - . - -

p

. .

-2-

, Persons Contacted Principle Licensee Employees

• P. Thomason, Division Manager of Nuclear Operations

• K. Wire, Operations Manager .

• D. Whitman, Technical Staff Manager

'J. Meacham, Technical Manager ,

P. Ballinger, Reactor Engineering Supervisor J. Sayer, Staff Assistant

"G. Mace, Plant Engineering Supervisor ,

'

• C. Goings, Regulatory Compliance Specialist M. Wolken, Outage Coordinator .

• R. Brungardt, Operations Supervisor R. Black, Assistant to Operations Supervisor R. Windham, Emergency Planning Coordinator

'

.

In addition to the above employees, the SRI held discussions with other i licensee employee * Denotes presence at exit meeting . Operational Safety Verification The SRI observed control room operations, instrumentation, controls, reviewed applicable logs, and conducted discussions with control room operators. The SRI verified operability of: P

. Offsite Electrical Power System

. 250 VDC Electrical Power System

. 125 VDC Electrical Power System

. Service Water System (Intake Structure)

. Reactor Protection System Power Supply System The SRI reviewed safety clearance records, including verification that affected components were removed from and returned to service in a correct and approved manner, that redundant equipment was verified operable, and that limiting conditions for operation were adequately identified and *

,

maintained. The SRI also verified that maintenance requests had been initiated for equipment discovered to require repair or routine preventive upkeep, appropriate priority was assigned, and maintenance commenced in a

' timely manner commensurate with assigned prioritie Tours of accessible areas of the facility were conducted to verify that minimum shift crew requirements were met, to observe normal security practices, plant and equipment conditions including cleanliness, radiological controls, fire suppression systems, emergency equipment, potential fire hazards, fluid leaks, excessive vibration and instrumen-tation adequacy.-

, ._

- _ . . - . -- - . - _ . . - - .

.

-3-The SRI observed a normal plant shutdown conducted on April 7,1984. The planned shutdown was performed to place the plant in a cold shutdown con-dition for an approximate one week maintenance outage. The main turbine was off-loaded at 4:30 a.m. and the reactor shutdown at 5:3/s a.m. During the outage, the licensee installed a new station service transformer, performed inspections and measurements of the reactor recirculation system and other piping and equipment in the primary containment to gather information needed prior to the fall 1984 major outage, inspected contain-ment vent header piping and penetrations associated with or located adja-cent to the Npinerting system piping, and performed other minor repairs, tests, and inspections. The SRI observed the performance of the following procedures during the plant shutdown evolution:

2. Normal Shutdown From Power 2.1.10 Station Power Changes 2.1.15 Reactor Recirculation Pump Operation 2. Circulating Water System 2. Condensate System 2.2.14 22 KV Electrical System 2.2.28 Feedwater System The SRI observed a reactor startup conducted on April 14, 1984, that followed the completion of the spring outage discussed above. The reactor achieved criticality at 2:39 a.m. and the main turbine was loaded at 3:49 p.m. Performance of the following procedures was witnessed by the SRI during the startup:

2. Cold Startup Procedure 2. Approach to Critical 2.2.56 Main Steam and Turbine Bypass System 6.4. Withdrawn Control Rod Operability The SRI reviewed the following completed procedures which were applicable to the April 14, 1984, startup:

2.1. Technical Specifications Pre-Startup Checks 6. SRM Functional Test (Reactor Not In Run)

6. IRM Functional Test (Mode Switch Not In Run)

6.1.22 APRM System 15% High Flux and Inop Trip Functional Test 6.1.24 Rod Worth Minimizer Functional Test For Startup 6.1.26 Rod Sequence Control System Functional Test For Startup The SRI observed a plant shutdown conducted on April 19, 1984. The main turbine was off-loaded at 2:01 p.m. and the reactor made subcritical at ,

2:00 The plant was subsequently cooled down and depressurize The shutdown was required by the CNS Technical Specification as a result of the j simultaneous loss of both Standby Gas Treatment Systems (SGT). The  :

circumstances surrounding the loss of SGT is discussed in paragraph ,

e

- , - - - -

. . _ . . - - , ,, ,

i

>

. .

r

, ,

-4-

+

t Ferformance of the following procedures was witnessed by the SRI during the plant shutdown evolution:

2. Normal Shutdown From Power 2.1.10 Station Power Changes 2. Circulating Water System 2. Condensate System 2.2.14 22 KV Electrical System 2.2.28 Feedwater System The SRI observed a reactor startup that was performed on April 20, 198 The reactor achieved criticality at 9:56 p.m. and the main turbine loaded

,

at 7:29 a.m. on April 21, 1984. The performance of the following procedures was witnessed by the SRI during the reactor startup and heatup: ,

2. Cold Startup Procedure 2. Approach to Critical 2.2.56 Main Steam and Turbine Bypass System ,

The SRI reviewed the following completed procedures applicable to the April 20,1984, startup:

2.1. Technical Specifications Pre-Startup Checks 6. SRM Functional Test (Reactor Not In Run)

6. IRM Functional Test (Mode Switch Not In Run) -

6. APRM System Excluding 15% Trip Functional Test 6.1.22 APRM System 15% High Flux and Inop Trip Functional Test 6.1.24 Rod Worth Minimizer Functional Test For Startup 6.1.26 Rod Sequence Control System Functional Test For Startup 6.3.10.13 North and South SDV Vent and Drain Valves Cycling, Open Verification, and Timing Test ,

The tours, reviews, and observations were conducted to verify that facility operations were in conformance with the requirements established in the CNS Operating License and Technical Specificatio I No violations or deviations were identified in this are . Manthly Surveillance Observations

.The SRI observed Technical Specification required surveillance tests to t verify that test prerequisites were completed, testing was performed in  :

accordance with approved procedures, test instrumentation was in cali-bration, limiting conditions for operation were met, removal and subsequent i

. restoration of affected components was accomplished, test results conformed with Technical Specification and procedure requirements, tests were reviewed by personnel other than the person directing the tests, and deficiencies identified during testing were properly reviewed and resolved by appropriate management personnel.

.

- -- -- - - - - . -e ,-, e - - -~ ,

..,,-,-,,,m,e-cy -- ,,_, , , -m,- g -em,. .-- , - - ..,_ - - - - . w

.

-5-

, These reviews and observations were conducted to verify that facility operations were in conformance with the requirements established in the CNS Operating License and Technical Specificatio No violations or deviations were identified in this are . Monthly Maintenance Observations The following clearance orders were independently verified for proper placement / restoration of affected components:

84-166 "A and C" Service Water Zurn Strainer 84-192 High Pressure Coolant Injection Pump Auxiliary Oil Pump 84-221 "D" 5crsice Water Booster Pump Incluceu with the above were checks for availability of redundant equip-ment, adequate safety isolation' and clearance, work was accomplished by

. qualified personnel in accordance with approved procedures and Technical Specification requirc; rents, verification that QC checks were performed as required, cleanliness controls and health physics coverage were adequate, and post-maintenance surveillance testing was performed to prove operability of the affected component and/or syste These reviews and observations were conducted to verify that facility operations were in conformance with the requirements established in the CNS Operating License and Technical Specificatio No violations or deviations were identified in this are . Licensee Event Report Followup (LER)

The following LER is closed on the basis of the SRI's inoffice review, review of licensee documentation, and discussions with licensee personnel:

LER 84-004 Failure of the Reactor Core Isolation Cooling Pump Speed l Control System Plant Trips - Safety System Challenges The licensee initiated two manual reactor scrams from low reactor power levels to facilitate timely entry into required outages. The scrams occurred on April 7,1984, and April 19, 1984, (Scram Reports 84-03 and 84-04). All syshms responded to both scrams as designed and appropriate followup actions were performed by plant personnel. The outages were required in order to perform scheduled and unscheduled maintenance; details are provided in paragraphs 2 and 7.

l No violations or deviations were identified in this are .

-

-6-

.

7. Declaration of Unusual Event At 10:45 a.m. on April 19, 1984, Mr. P. Thomason, Division Manager of Nuclear Operations, informed the SRI that the licensee had declared both trains of the Standby Gas Treatment System (SGT) inoperable. As a result of the loss of both trains of the SGT, the licensee entered the Limiting Condition for Operation (LCO) established in Section 3.7.B.4 of the CNS Technical Specification which required the plant to be shutdow The CNS Emergency Plan required the licensee to declare an UNUSUAL EVENT when entry was made into the Technical Specification LC0 that resulted in the shutdown of the plan Technical Specification Section 3.7.B.1, states , in part, ". . . standby gas treatment system . . . shall be operable at all times when secondary containment integrity is re"uired."

,

Technical Specification Section 3.7.C.1, states, " Secondary containment integrity shall be maintained during all modes of plant operation except when all of the following conditions are met: 1) the reactor is sub-critical . . . , b) the reactor water temperature is below 2120F and the i reactor coolant system is vented, c) no activity is being performed which can reduce the shutdown margin below that value specified in Specification 3.3. A. , d) irradiated fuel is not being handled in the secondary containment."

Technical Specification Section 3.7.B.3 permits reactor operation for 7 days from and after the date that one circuit of the SGT is found to be inoperable. Technical Specification does not discuss permissable reactor power operation if two circuits of the SGT are found to be inoperabl Technical Specification 3.7.B.3 permits reactor operation for Technical Specification Section 1.0.J states, in part, "In the event an LC0 cannot be satisfied because of circumstances in excess of those addressed in the specification, the facility shall be placed in H0T SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following

,

30 h ou rs . . . . "

Conditions leading to the requirement to declare both trains of the SGT inoperable began at 10:15 a.m. on April 19, 1984, when a construction bulldozer sheared off a fire hydrant in the north yard of the CNS protected area. Fire pumps automatically started in an attempt to maintain fire-main pressure. The fire pumps were subsequently secured until the damaged hydrant was isolated. The fire pumps were retuned to service and the fire protection system repressurized by 10:30 a.m. Shortly thereafter, control room personnel received an alarm annunciation indicating high moisture in both SGT trains. An immediate inspection was performed in the SGT room, which verified that the fire deluge system had activated and flooded the charcoal absorbers in both trains, thus rendering both SGT trains inoperabl _

t

.

!

-7-Subsequent followup investigations appeared to indicate that the SGT deluge system clapper valves opened due to a water hammer effect created when the fire pumps were returned to service following isolation of the damaged hydrant. System design and automatic features of the SGT deluge system is being reevaluated to determine if the system should be modified or *

procedures revised to preclude reoccuranc Having determined by Technical Specification requirements that plant shut-down was required, the licensee commenced shutdown at 11:00 a.m. At 2:00 p.m., the reactor was scrammed from 34% power which placed the i facility in HOT SHUTDOWN. An orderly reactor cooldown was commenced  ;

innediately and at 11:24 p.m., the reactor reached COLD SHUTDOWN which ,

satisfied all Technical Specification LC0 requirements applicable to the loss of both trains of the SGT syste CNS Emergency Plan Implementation Procedure 5.7.1, Emergency Classification, Section 6.1, requires the licensee to declare an UNUSUAL EVENT if any  ;

Technical Specification LC0 results in plant shutdown. The licensee i declared a Notification of Unusual Event at 11:40 a.m. and terminated the

event at 11:25 p.m. when COLD SHUTDOWN conditions were established.

.

At 11:36 a.m., the SRI arrived in the control room to monitor licensee activities. At 11:41 a.m., the SRI notified Region IV of initial licensee actions and implementation of their Emergency Plan. He provided supplementary information to Region IV as necessary throughout the event.

The SRI observed the following licensee actions:

i

. Identification of Technical Specification LCOs applicable to the I loss of both trains of the SGT syste ["

. Reactor Shutdown and initiation of plant cooldown per plant procedure and Technical Specification requirement ;

. Implementation of the CNS Emergency Plan including appropriate event  !

- classification and notifications of the NRC and off-site agencie I J

"

The SRI performed followup inspections of the below listed data, procedures f

and events: '

.

. T(mperature recorder tracking of the plant cooldow . Initiation of the shutdown cooling mode of the residual heat renoval syste . Review of control room logs for complete and timely entry of tignificant plant operational events and mode change . Re>iew of procedures and checklists applicable to the declaration t of and termination of the UNUSUAL EVEN . Review of surveillance procedures associated with the SGT system following replacement of the charcoal bed . Review of licensee followup reports to the NR ;

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

. , , _ _

_- . _ _ _

,

. .

,

-8-

,

.

• EPIP 5. Attachment B Classification Checklist ,

'

• EPIP 5. Attachment C Classification Guide r

• EPIP 5. Attachment A Notification of Unusual Event Implementing Procedure Checklist i

• EPIP 5. Attachment A Nuclear Power Plant Incident Initial Report  ;

• EPIP 5. Attachment C Emergency Notification Call Checklist ;

• EPIP 5. Attachment D Station Internal Call List

'

,

6.3.1 SGT Operability Test 6.3.1 SGT Filter Differential Pressure and Heater Output Test  :

6.3.1 SGT HEPA Filters Leak and Housing Door Seal Leak Test  !

SGT Charcoal Filter Leak. and Fan Capacity Test

'

6.3.1 .27 Attachaent A Notification of Significant Events Checklist (Per 10 CFR 50.72)  !

, *EPIP Emergency Plan Implementing Procedure

'

The SRI performed the above observations, reviews, and followup and-  !

determined that the licensee performed all actions required by the Technical Specification, operating procedures, and the CNS Emergency Pla No violations or deviations were. identified in this are ; Independent Inspection Effort An SRI review of the CNS Technical Specifications has revealed the following anomalies: CNS Technical Specification 1.0, part J, states in part,

"

. . . entry into an operational condition shall not be made unless the conditions of the LC0 are met without reliance on the actions specified in the LC0 unless otherwise excepted. This provision shall not prevent passage through operational conditions required to comply-with the specified actions of an LCO."

The following are other definitions located throughout Section of the CNS Technical Specifications:

" Definitions The succeeding frequently used terms are explicitly defined so that a uniform interpretation of the specifications may be achieve Cold Condition - Reactor coolant temperature equal to or less than 212* [

- - , . .. .

. - - - - - . - - - - --. . - . -

. _ _

. .

-

9-

, Hot Standby Condition - Hot standby condition means

operation with coolant temperature greater than 212*F, system pressure less than 1000 psig, and the mode switch in Startup/ Hot Standb Mode - The reactor mode is established by the mode selector-switch. The modes include refuel, run, shutdown and startup/ hot standby which are defined as follows

Refuel Mode - The reactor is in the refuel mode when the mode switch is in the refuel position, the re-fueling interlocks are in servic . Run Mode - In this mode, the reactor system pressure is at or above 825 psig and the reactor protection system is energized with APRM protection (excluding the 15%

high flux trip) and RBM interlocks in servic ! Shutdown Mode - The reactor is in the shutdown mode when the reactor mode switch is in the shutdown mode positio . Stactup/ Hot Standby - In this mode the reactor pro-tection scram trips initiated by the main steam line ,

'

isolation valve closure are bypassed when reactor pressure is less than 1000 psig, the low pressure main steam line isolation valve closure trip is by-passed, the reactor protection system is energized with APRM (15% SCRAM) and IRM neutron monitoring system i trips and control rod withdrawal interlocks in servic i Reactor Power Operation - Reactor Power Operation is any operation with the mode switch in the Startup/ Hot Standby  ;

'

or Run position with the reactor critical and above 1%  ;

rated powe Shutdown - The reactor is in a shutdown condition when the moae switch is in the Shutdown or Refuel positio . Hot Shutdown means conditions as above with reactor coolant temperature greater than 212* . Cold Shutdown means conditions as above with reactor coolant temperature equal to or less than 212*F and the reactor vessel vented."

.

--wt a-e --

r-- m -- --

y-v--y -- ,,,t -- ,- y -e ----

. . . . _ .

.

-10-Item (1.0) above emphasizes the importance of Section 1.0, Definitions, to explicitly define the terms used throughout the Technical Specification for purposes of uniform interpretation of those terms.

>

However, the definitions provided in (C), (G), (R), and (W) above are spread throughout Section 1.0. Also, the word " condition" is not attached to the definition titles listed in (R) and (W), although the definitions are addressing specific plant condition Definition (L) above discusses modes which are synonymous with specific plant conditions as well as reactor control mode switch posi ti on Standard Technical Specifications for General Electri. Boiling Water Reactors , "NUREG-0123," Table 1.2, lists and defines five BWR operational conditions as indicated below:

" TABLE OPERATIONAL CONDITIONS MODE SWITCH AVERAGE REACTOR CONDITION POSITION COOLANT TEMPERATURE- POWER OPERATION Run Any Temperature STARTUP Startup/ Hot Standby Any Temperature HOT SHUTDOWN Shutdown # 2000 F COLD SHUTDOWN Shutdown # 2000 F REFUELING * Shutdown or Refuel ** 140 F

1. The reactor mode switch may be placed in the Run or Startup/ Hot Standby position to test the switch interlock functions provided that the control rods are verified to remain fully inserted by a second licensed operator or other. technically qualified member of the unit technical staf * Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head remove **See Special Test Exception 3.10.3" The CNS Technical Specification Section 1.0 should be revised to provide concise and grouped definitions of all plant operational conditions like that provided by Table 1.2 of the Standardized Technical Specification illustrated above. Also, the licensee should attempt to incorporate in the definitions those plant conditions which are defined by the reactor control mode switch (definition (L) above). Specific conditions listed with Technical Specification LCOs should be reviewed and revised as necessary to agree with such a revision to Section 1.0. The licensee has agreed .

_ _ _ - _ . _-- . _ _ _ . _ - - . _ - _ _ _ _ - . _ _ _ - _ _ , . ._-

- .

. .

-11-to perform a review of the above anomalies and provide recommenda-tions, if warranted, to the NPPD licensing department concerning possible revision to the Technical Specifications. This item will remain an open item pending completion of the licensee's revie (298/8410-01)

b. Table 3.1.1. of the CNS Technical Specification provides a listing of reactor protection system instrumentation requirements and the actions required to be performed by licensee personnel when equipment operability is not assure Page 30 of the CNS Technical Specification titled, " Notes For Table 3.1.1," note 1, states:

"There shall be two operable or tripped trip systems for each function. If the minimum number of operable instrument channels for a trip system cannot be met, the affected trip system shall be placed in the safe (tripped) condition, or the appropriate actions listed below shall be take " Initiate insertion of operable rods and complete insertion of

all operable rods within four hour " Reduce power to less than 30% or rate " Reduce power level to IRM range and place mode switch in the Startup position within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and depressurize to less than 1000 psi " Reduce turbine load and close main steam line isolation valves within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />."

There appears to be contradiction in the required action statements when the requirements of Table 3.1.1 cannot be met. Specifically, Table 3.1.1 states that if the required minimum number of operable APRM channels per trip system is less than the table specifies, then action is to be taken in accordance with notes A or C above. Note A specifies a 4-hour limit in which all operable control rods must be fully inserted and Note C specifies an 8-hour limit,to be in the IRM rang The SRI discussed the above anomalies with the licensee. The licensee has agreed to review the anomalies and make Technical Specification revision recommendations, if warranted, to the NPPD licensing department. This item will remain an open item pending completion of the licensee's review. (298/8410-02)

. .

-12- Followup of Previously Identified Items Open Item 8220-01 (Closed). Exposure Device Discrepancy Evaluatio This is a recurrent item. See paragraph 9.F for details. This item is considered closed, Violation Item 8320-01 (Closed). Failure to Follow Procedure - Fire Door Not Included in surveillance Inspectio The licensee performed an Appendix R review of doors P105 and P106 that provide access to the service water (SW) pump room and determined that neither door is a rated fire door nor are they re-quired to be rated fire doors. E.emption to Appendix R was granted for the SW pump room area in a letter from Mr. D. B. Vasssllo (NRR) to Mr. J. M. Pflant (HPPD) dated December 14, 1982. The RI reviewed surveillance procedures 6.4.5.1 and 6.4.5.2 and determined that doors P105 and P106 were properly designated as non-fire rated and that surveillance requirements were revised accordingly. It was roted that Admin-istrative Procedure 1.4, Attachment F, still identified door P105 as a fire door. The RI was informed that a procedure change re-quest had been initiated to correct the ciscrepancy in procedure This item is considered close Violation Item 8324-03 (Closed). Failure to Protect Quality-Related Records Against Destruction by Fir The licensee's approved quality assurance plan allows temporary storage of many records for up to 2 years and states that provisions of ANSI N45.2-9-1974 will be used for management retention and storage of quality-related records. However, an NRC inspector had noted that all temporary records storage was in thin-walled file cabinets of a type that did not have a fire rating. ANSI N45.2-9-1974 requires protection of records against destruction by fire by storage in cabinets having a two-hour fire ratin The licensee has purchased and distributed 29 Class 350, UL fire rated file cabinets having a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> rating which is the maximum fire rating available for Class 350 factory built fire cabinets presently on the market. The CNS Quality Assurance Policy Document, Section 8.9(b), has been revised to reflect the availability of the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> rated file cabinet The RI observed the presence of the new file cabinets in place and in use in several CNS office spaces. The RI performed spot checks of surveillance data sheets, preventative maintenance records,

__ _- ____ ____ _ __ _ ___ __________ _ - ______ __ _ _____ _ ______

. .

-13-

.

personnel radiation exposure records, and equipment calibration records and determined that none of the records were greater than

< 2 years old and that they were stored in the new cabinets. This item is considered close Open Item 8324-04(Closed). Equipment Calibration Frequency Control An NRC inspector had traced calibration records for specific metrology equipments and noted that gauges were routinely recalled for calibration on a quarterly basis. The NRC inspector had discovered that the licensee used gauges 1 to 2 weeks beyond the quarterly recall date and was informed that the licensee used an automatic 30% extension on the gauge recall schedul Licensee procedures used to control retrology did not include the practices of quarterly recall nor the 30% extension criteri The RI reviewed CNS procedure 1.7.2, tit!ed, " Work Item Tracking,"

Revision 3, dated August 30, 1983. Procedure 1.7.2 now includes the practices of quarterly recall and the 30% extension criteria. This item is considered close Violation Item 8325-02 (Closed). Leak Check of Radioactive Source This violation concerns the failure of the licensee to leak check 27 radioactive sources within the 6-month required frequenc CNS personnal misinterpreted the statement, "not to exceed 6 months " to include a 125% allowable extension. All sources were leak checked prior to the conclusion of the inspection. The licensee has modified the radioactive source leak check analysis schedule and the status board to clearly indicate that the 6-month frequency is not subject to an automatic 25% extension. All CNS personnel have been instructed on this item. The licensee's response was considered satisfactory. This item is considered closed, Violation Item 8325-03(Closed). Exposure Device Discrepancy Evaluatio This violation concerns failure of the licensee to evaluate discrepancies between thermoluminescent dosimeters (TLDs) and pocket chamber dosimeters exposure results when those results exceed specified limit . .. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .

_

. ..

-14-

,

CNS procedure 9.1.1.3, Personnel Dosimetry Program, was revised to institute a new monthly computer printout sheet titled,

"TLD/ Pocket Chamber Ratio Summary." The Printout lis'ts the names of zall personnel whose monthly TLD/ Pocket Chamber results rctio is greater than 1.2 or less than The printout provides signature spaces which are to be signed by a health physicist and the chemis-try and health physics superviso Space is also provided for comments and details of the evaluatien. The RI reviewed 4 months of printouts and did not find any discrepancies. This item is con-sidered closed, Violation 8325-04 (Closed). Failure to Maintain Records of i

Radiation and Airborne Radioactivity in The Same Units Used in 10 CFR Part 20 and Surveys Were Not Recorded on Form CNS-10 Since the areas in which the surveys in question no longer exist (acid neutralization area) or were unaccessible when the violation was observed (dry well), it was not possible to repeat the surveys, verify the results, and complete the proper documentation. The RI re-viewed the documented training record of instruction provided to 20 CNS HP personnel relevant to this item. The training class was 3/4 hour in duration and covered the NRC inspection report finding and the proper actions that should have been taken. The training

,

included the requirements of 10 CFR 20 and the use of Form CNS-HP-10 The licensee's response is considered satisfactory. This item is

~

considered close Violation 8325-06 (Closed). Failure of a Contract Employee to Wear Personnel Radiation Monitoring Devices (TLD and pocket chamber dosimeter) While Working In a Radiation Area.

!

The licensee conducted a special radiation survey of the area where l the contract employee had worked. Using the dose rate results versus time calculation and including a conservative safety factor, the i licensee calculated an estimated total radiation dose of 720 MR.

i The estimated dose was entered on the employee's Radiation Form NRC- The RI attended a typical new employee radiation worker training session which was conducted by a member of the health physics department. The training was well structured and thorough, specifically in the areas of radiation monitoring devices usage and CHS administrative requirements. The training was immediately followed by a written examination. The licensee's responses to this t violation is considered satisfactory. This item is considered close . $xit Meetings Exit meetings were conducted at the conclusion of each portion of the in-spection. The division manager of nuclear operations was informed of the above finding ,

.

-r - ,e ~-- - -,-r-- - u-. , a- s ,-,--;m,,,,-,,-r- , , - - w y-- --p