IR 05000213/1997001: Difference between revisions

From kanterella
Jump to navigation Jump to search
(StriderTol Bot insert)
 
(StriderTol Bot change)
Line 1: Line 1:
{{Adams
{{Adams
| number = ML20141F830
| number = ML20151L907
| issue date = 05/08/1997
| issue date = 07/31/1997
| title = Insp Rept 50-213/97-01 on 970106-0407.Violations Noted.Major Areas Inspected:Licensee Operations,Engineering,Maint & Plant Support
| title = Ack Receipt of 970606 Ltr Informing NRC of Steps Taken to Correct Violations Noted in Insp Rept 50-213/97-01
| author name =  
| author name = Conte R
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
| addressee name =  
| addressee name = Feigenbaum T, Mellor R
| addressee affiliation =  
| addressee affiliation = CONNECTICUT YANKEE ATOMIC POWER CO., NORTHEAST UTILITIES SERVICE CO.
| docket = 05000213
| docket = 05000213
| license number =  
| license number =  
| contact person =  
| contact person =  
| document report number = 50-213-97-01, 50-213-97-1, NUDOCS 9705220064
| document report number = 50-213-97-01, 50-213-97-1, NUDOCS 9708080013
| package number = ML20141F808
| document type = CORRESPONDENCE-LETTERS, OUTGOING CORRESPONDENCE
| document type = INSPECTION REPORT, NRC-GENERATED, TEXT-INSPECTION & AUDIT & I&E CIRCULARS
| page count = 2
| page count = 101
}}
}}


Line 19: Line 18:


=Text=
=Text=
{{#Wiki_filter:.
{{#Wiki_filter:. - - - .. . _ . - - - . - - . . - - .. . - _ . . _ - - . . -
-
p..
U.S. NUCLEAR REGULATORY COMMISSION
 
==REGION I==
    !
Docket No.: 50-213 License No.: DPR-61 Report No.: 50-213/97-01  ,
    !
Licensee: Connecticut Yankee Atomic Power Company i P. O. Box 270  l Hartford, CT 061410270
 
Facility: Haddam Neck Station Location: Haddam, Connecticut Dates: January 6 - April 7,1997 Inspectors: William J. Raymond, Senior Resident inspector Ronald L. Nimitz, Senior Radiation Specialist Laurie A. Peluso, Radiation Physicist Eben L. Connor, Project Engineer Approved by: John F. Rogge, Chief, Projects Branch 8 Division of Reactor Projects l
9705220064 970508 "
PDR ADOCK 05000213 G  PM j
 
    -,  _ _ _ _ - _ - _ _ -
.
.
.
  #
EXECUTIVE SUMMARY Haddari Neck Station NRC Inspection Rsport No. 50-213/97-01 This integrated inspection included aspects of licensee operations, engineering, maintenance, and plant support. The report covers a three month period of resident inspection; in addition, it includes the results of announced inspections by regional inspectors in the areas of radiological controls, environmental monitoring, and plant procedure Plant Operations:
  -.
The conduct of opereting activities was acceptable, as were the operator actions to maintain stable defueled conditions, and to monitor the status of spent fuel cooling and systems in long term preservation. An inspection item will follow licensee initiatives to categorize and control plant systems for decommissioning, and to reduce the number of illuminated annunciators. The quality of procedures used for shutdown operations was acceptable, and licensee initiatives to improve and maintain procedures were satisfactor However, inadequate calibration procedures resulted in the declaration of all gaseous and liquid effluent radiation monitors required by the technical specifications to be inoperable, and the need to implement compensatory measures to monitor effluents by the associated release pathways. Cold weather preparations were adequate to protect support systems for spent fuel cooling; however, the failure to complete thorough plant walkdowns and ,
July 31, 1997 Mr. Ted . Executive Vice President and Chief Nuclear Officer Northeast Utilities Service Company c/o R. A. Mellor,~ Director Site Operations and Decommissioning Connecticut Yankee Atomic Power Company 362 injun Hollow Road East Hampton, CT 06424-3099 SUBJECT: INSPECTION REPORT NO. 50-213/97-01
poor insulation conditions on some piping resulted in freeze damage to a process lin l As in past inspections, human performance errors by operators and workers detracted from good performance. The procedure violation resulting in the operation of red tagged equipment was a significant and recurrent event. Several other examples of failure to j adequately follow plant procedures were noted. Poor performance was demonstrated in
        '
the administration of the licensed operator training program, and to assure that operators are fully qualified initially and remain qualified through periodic retraining. Weaknesses were noted in the control of overtime to assure that excessive work hours were approved per the administrative guidelines. Although the spent fuel remained adequately cooled at  ;
all times, a discrepancy in the design basis and deficiencies in the material condition of the
        '
service water system created a challtinge to the adequacy of the spent fuel cooling system. Licensee actions to address these deficiencies, as well as the operator actions to  I implement alternate cooling methoris for the spent fuel pool, were acceptabl l Maintenance:
i The maintenance and surveillance activities completed this period were generally  1 acceptable to assure important plant s'/ stems remained operable, to support operability  )
evaluations and design change work, and to address emergent issues that challenged  j adequate cooling of the spent fuel. Exceptions to good performance included a weakness in the process to examine service water (SW) pipe for corrosion, and a faulty test method used to measure leakage through a check valve. Additional discrepancies were noted in
        !
ii i
        !
I1
 
t plant material conditions, in the implementation of the program to identify and correct material discrepancies, and in the lack of progress in addressing deficient condition Poor performance was noted in the failure by station personnel to follow procedur6s to implement the operational surveillance program, and in the repetitive failure to adequately implement technical specification surveillances in a timely manner. The extensive corrosion and general degradation in the service water system resulted in an inoperable condition for the system relied upon to cool the spent fuel. This finding appears as another i example of poor plant material conditions that challenge systems important to plant safet i The recurrence of plant material deficiencies and problems in the area of technical specification surveillance testing revealed ongoing weaknesses in the corrective action proces ;
Ennineerina:
Mixed performance was noted in the engineering support of operations. Engineering performance was good in response to emergent design basis issues and corrosion degradation in the SW system. Engineering evaluations were good to identify the inoperabilities in the SW piping, to assess the interim use of the degraded system, and to support the correction of design and corrosion issues. Engineering support for design i changes was also good to correct the SW problems. Exceptions to good performance
. were noted in the failure to fully integrate site and corporate engineering support, and to assure continued chemical (Bulaab) treatment of the SW system during shutdown conditions. Poor performance was noted in the failures to track design issues to completion (waterhammer), to properly classify design issues (two phase flow) for the shutdown condition of the plant, and to track commitments to the NRC. These issues appear as weaknesses in the corrective action proces Several design basis discrepancies were discovered either by good licensee staff and QAS initiatives to identify and resolve discrepancies, or by the Configuration Management Plan (CMP) group as the reviews to reconstitute the plant design and licensing basis are completed. The identification of additional design basis issues is expected until the completion of the CMP plan for the shutdown and decommissioned plant. However, l exceptions to good performance were noted in the failure to address the design basis issue adequately relative to SW temperature, which appears as an example of a corrective action weakness. Inspection items will track the completion of licensee actions to correct discrepancies in the plant design and material condition Plant wide weaknesses in the corrective action process were noted during this period as events occurred that were a repeat of past problems. While the number of deficiencias identified by licensee has generally increased, a large number of deficiencies were identified either by self disclosing events or by the independent oversight groups. The licensee has yet to complete initiatives to improve the corrective action program to assure consistent quality root cause investigations and to implement effective corrective action iii


I l
==Dear Mr. Feigenbaum:==
This letter refers to your June 6,1997 correspondence, in response to our May 8,1997 letter.


I
Thank you forinforming us of the corrective and preventive actions documented in your letter.
-
l
.
Plant Support:
The licensee continued to implement controls for work within the radiological controlled area in accordance with commitments outlined in its December 9,1996, letter to the NR A number of program improvements, in response to the November 2,1996, fuel transfer canal event, were completed. Weaknesses were identified in the contamination control and ALARA program that should be addressed prior to decommissioning. The capabilities of the quality assurance organization were enhanced through the addition of individuals


with extensive radiological controls experience and experience in decommissionin Because of NRC concerns regarding the adequacy and effectiveness of the radiological )
These actions will be examined during a future inspection of your licensed program.
controls program, a Confirmatory Action Letter (CAL 1-97-007) was issued on March 4,199 ,
      ,
The licensee continued to implement an overall effective radiological environmental 1 monitoring program (REMP) including management controls, quality assuranco audits, radiological environmental monitoring, and meteorological monitoring program. The Radiological Effluent Monitoring and Offsite Dose Calculation Manual (REMODCM) was properly implemented. The 1996 audit report effectively assessed program strengths and weaknesses and had improved from previous audit assessments. No deficiencies in the Updated Final Safety Analysis Report commitments were identified. An inspection follow-up item was identified related to determining the licensee's conformance with 40 CFR 19 Safety Assessment & Quality Verificatioru Several findings by the QA audit and surveillance groups during the period demonstrated good performance by the oversight groups to identify deficiencies in plant operating activities. The failures to complete correc.tive actions, including the failure to complete committed regulatory actions, were a previously identified weakness. The licensee has i implemented plans to address this area and to improve performance. Despite the progress I
      '
in identifying deficiencies, the licensee had demonstrated continued weaknesses in implementing adequate corrective actions. While the number of deficiencies identified by l licensee staff has generally increased, a large number of deficiencies were identified either 1 l
by self disclosing events, by the independent oversight groups (quality organization, NRC).


There have been further examples of events that occurred which were a repeat of past problems. The examples included the deficiencies in the tracking and timely completion of surveillance activities; the recurrence of personnel performance areas in broad areas of plant operations; and, the failure to preclude plant operation outside the design basis (river water temperature, service water two phase flow). The licensee has yet to complete its initiatives to improve the corrective action program to assure consistent quality root cause investigations and to implement effective corrective action I l
Your cooperation with us is appreciated.       .
I i
          !
l iv
      !
l l


l
Sincerely, Original Signed By:
 
Richard J. Conte, Chief Projects Branch 8 Division of Reactor Projects Docket No. 50 213      ')
    - . .- -.  ., . . - -
i cc: w/o cv of Licensee's Response Letter      {
.        .
B. D. Kenyon, President - Nuclear Group D. M. Goebel Vice President - Nuclear Oversight     4 J. K. Thayer, Vice President - Recovery Officer, Nuclear Engineering and Support F. C. Rothen, Vice President - Work Services
        .
     -
.
          '
        (
  ' R. Johannes, Director - Nuclear Training
TABLE OF CONTENTS    i
        !
        '
EXEC UTIV E StJ M M A RY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii TABLE OF CONTENTS . . . . . . . . . . . . . . . . . . . . . . . ...................... V
        ,
R E PO RT D ETAI LS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 Su m m a ry o f Pla nt St a tu s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1. O P ERA TI O N S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
. 01 Conduct of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 01.1 Review of Operating Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
'
        '
01.2 Status and Control of Systems in the Defueled Mode (URI 97-01-01) . . . 2 01.3 Spent Fuel Pool Cooling and Building Ventilation ................. 3 01.4 Inoperable Spent Fuel Cooling System . . . . . . . . . . . . . . . . . . . . . . . . . 4-
. . 01.5 Implementation of Alternate SFP Cooling ...................... 6 01.6 Loss of a 115 kV Distribution Line ........................... 7 01.7 Inoperable Effluent Monitoring System ........................ 8 01.8 Control of System Configuration (VIO 9 7-01 -0 2. a ) . . . . . . . . . . . . . . . . 8 01.9 Inadequate Control of Boiler Operations (VIO 9 7-01 -02.b) . . . . . . . . . . 10 01.10 Conclusions for Conduct of Operations ...................... 11 02 Operational Status of Facilities and Equipment . . . . . . . . . . . . . . . . . . . . . . . 11 O2.1 Cold Weather Preparations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 03 Operations Procedures and Documentation  ......................... 13 03.1 Procedure Quality for Shutdown Operating Activities . . . . . . . . . . . . . 13  ,
05 Operator Training and Qualification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 05.1 Inaccurate Operator Training Records (URI 9 7-01 -0 3 ) . . . . . . . . . . . . . 14  '
06 Operations Organization and Administration . . . . . . . . . . . . . . . . . . . . . . . . . 17 06.1 Staffing and Control of Overtime (VIO 9 7-01 -0 2.c ) . . . . . . . . . . . . . . . 17
'
08 Previous Operations Open issues (92901) . . . . . . . . . . . . . . . . . . . . . . . . . . 19 08.1 (Closed) VIO 94-21-01, inadvertent Boron Dilution . . . . . . . . . . . . . . . 19 08.2 (Closed) URI 94-27-01, Loss of Electrical Separation . . . . . . . . . . . . . . 20 08.3 (Closed) LER 94-011-00, Unplanned Loss of Spent Fuel Cooling ..... 20 08.4 (Closed) LER 94-015-01, Main Steam Vaives Exceed Lift Setpoints . . . 21 08.5 (Closed) IFl 96-08-01, RHR Calibrations and Leakage . . . . . . . . . . . . . 22 08,6 (Closed) LER 95-023-00, Failure to Prepare Special Report . . . . . . . . . 22 08.7 (Closed) URI 96 201-10, Alternate Auxiliary Feedwater Sources ..... 22 v
     %
 
.
.
08.8 (Closed) LER 96-015-00, Containment Air Monitor Trip Valve . . . . . . . 23 II . M AI NT EN A N C E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 M1 Conduct of Maintenance ...................................... 23 M1.1 Maintenance Observations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 M1.2 Surveillance Observations ................................ 28 M2 Maintenance and Material Condition of Facilities and Equipment .......... 31 M 2.1 Material Condition Deficiencies (VIO 97-01-02.d, URI 97-01-04) ..... 31 M3 Maintenance Procedures and Documentation . . . . . . . . . . . . . . . . . . . . . . . . 33 M3.1 TS Surveillances Covered by Procedures (VIO 97-01-02.e) ......... 33 M4 Maintenance Staff Knowledge and Performance . . . . . . . . . . . . . . . . . . . . . . 34 M4.1 Failure to Complete Surveillances (VIO 97-01-05, VIO 97-01-06) . . . . . 34 M8 Area Summary and Status of Regulatory Findings  .................... 37 M8.1 (Closed) IFl 96-01-01, Cable Vault Materials Condition , . . . . . . . . . . . 37 M8.2 (Closed) DEV 96-04-02, Heavy Load Program Commitments . . . . . . . . 37 M8.3 (Closed) IFl 96-08-04, Auxiliary Feed Water Overspeed Trip ........ 37 M8.4 (Closed) IFl 96-08-05, Steam Generator Hold Down Bolts . . . . . . . . . . 38 M8.5 (Closed) IFl 96-08-06, Observations of Procedural Quality.. . . . . . . . . . 38 M8.6 (Open) URI 96-08 15, Start-up issues (7/24/95 NRC Letter) ........ 38 M8.7 (Closed) URI 94 27-04, Surveillance Frequency Exceeded . . . . . . . . . . 39 M8.8 Conclusions for Maintenance .............................. 39 t il . E N G I N E E R I N G . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 El Conduct of Engineering ....................................... 40 E Service Water System Modification - Water Hammer (URI 97-01-07) .. 40 E1.2 Service Water System Evaluations - Corrosion (URI 9 7-01 -0 8 ) . . . . . . . 43 E1.3 Conclusions for Conduct of Engineering ...................... 45 E3 Engineering Documentation - Design Basis Discrepancies (40500) . . . . . . . . . 45 E3.1 Handling Loads Over Stored Fuel ........................... 40 E3.2 Control Room Habitability . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 46 E3.3 Mis sed Commitm e nts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 47 E3.4 Service Water Design Basis issues  .......................... 48 E3.5 Service Water System Water Hammer . . . . . . . . . . . . . . . . . . . . . . . . 49 E3.6 Inoperable Effluent Monitor - Stack Noble Gas . . . . . . . . . . . . . . . . . . 50 E3.7 Spent Fuel Building and Yard Crane Design Basis issues . . . . . . . . . . . 50 E3.8 Conclusions for Engineering Documentation (URI 97-01-09)  ........ 51 E6 Engineering Organization and Administration  ........................ 51 E Corrective Action Program Weaknesses . . . . . . . . . . . . . . . . . . . . . . . 51 E8 Miscellaneous Engineering issues (92902) . . . . . . . . . . . . . . . . . . . . . . . . . . 53 vi
 
.
.
E Spent Fuel Pool Design to Support Full Core Off-load . . . . . . . . . . . . . 53 E8,2 (Closed) IFl 94-09-03, HPSI Relief Valve Setpoint Drift ............ 54 E8.3 (Closed) VIO 96-04-03, inadequate Safety Evaluation . . . . . . . . . . . . . 55 E8.4 (Closed) IFl 96-04-04, Heavy Load Controls . . . . . . . . . . . . . . . . . . . . 55 E8.5 (Update) URI 96-06-06, Battery Oscillations and Ground . . . . . . . . . . . 55 E8.6 (Closed) URI 96-201-12, Analysis Supporting LPSI Pump Shutdown  .. 56 E8.7 (Closed) URI 96-201-31, RWST instrument Calibrations ........... 56 E8.8 (Open) URI 96-02-03, Control Room Habitability ................ 57 IV. PL A NT S U PPO RT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 59
        !
-R1 Radiological Protection and Chemistry (RP&C) Controls . . . . . . . . . . . . . . . . . 59 R1.1 External and Internal Exposure Controls . . . . . . . . . . . . . . . . . . . . . . . 59 R1.2 ALARA Program ....................................... 61 R1.3 Radiological Environmental Monitoring Program (IFl 97-01 -10) . . . . . . . 63 R1.4 Meteorological Monitoring Program (MMP) . . . . . . . . . . . . . . . . . . . . . 66 R2 Status of RP&C Facilities and Equipment ........................... 67 H2.1 Inoperable Ef fluent Monitors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 67 R3 RP&C Procedures and Documentation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 69 R3.1 Whole Body Counting ................................... 69 R3.2 Contamination Controls (URI 9 7 -01 - 1 1 ) . . . . . . . . . . . . . . . . . . . . . . . 70 R5 Staff Training and Qualification in RP&C (URI 9 7-01 - 12 ) . . . . . . . . . . . . . . . . 70 i
R6 Radiological Protection and Chemistry (RP&C) Organization and Administration ............................................. 71 R6.1 Management Controls ................................... 72 R7 Quality Assurance in RP&C Activities ............................. 73 R7.1 Quality Assurance Audit Program . . . . . . . . . . . . . . . . . . . . . . . . . . . 73 R7.2 Quality Assurance of Analytical Measurements  ................. 75 i
l R8 Miscellaneous RP&C issues .................................... 76 !
R8.1 Decommissioning Project Planning .......................... 76 l R8.2 Followup of the November 2,1996, Fuel Transfer Canal Event ...... 76 i R8.3 (Open) URI 96-12-01,02: Exposure Assessment, Dose Calculations .. 79 l R8.4 H ou s e k e e ping . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 83 R8.5 Confirmatory Action Letter . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 83 R8.6 UFSAR .............................................. 83 l
P2 Status of EP Facilities, Equipment, and Resources .................... 84 l P2.1 Emergency Plan Staffing ................................. 84 )
S1 Conduct of Security and Safeguards Activities . . . . . . . . . . . . . . . . . . . . . . . 85 S1.1 Fit n e s s f or D uty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 85 S1.2 Guard Inattentive To Duty ................................ 85 i vii     j i
l l
l
 
  --- . _ _ _ - - _ _ . -
F
   .
   .
I l
L. M. Cuoco, Senior Nuclear Counsel G.= P. van Noordennen, Manager, Nuclear Licensing H. F. Haynes, Director - Nuclear Training      j J. F. Smith, Manager, Operator Training      !
l l  .
l$.dl
I S1.3 Response to Potential Threat .............................. 85 S1.4 Failure to Search Packages (VIO 97-01-02.f) ................... 85 V. MANAGEMENT MEETINGS ....................................... 86 X1 Exit M e eting Summ ary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 86
 
l I
l l
l
 
l l
VIIi    )
i l
 
_  _
.
.
REPORT DETAILS Summarv of Plant Status Haddam Neck remained shutdawn with the reactor in a defueled condition during the inspection period. The licer.see formally ended refueling outage RFO #19 and maintained stable plant conditions while planning the activities needed to prepare the post shutdown decommissioning activit" report and prepare for decommissioning the plant. NRC activities at the site during the peiiod included the reviews by the resident inspector of post operating activities and the plans to commence decommissioning, and the inspections by region based personnel in the areas of effluent monitor calibrations, radiological controls, and the environmental monitoring progra Significant events during this period included the discovery that contaminated materials had been inadvertently released from the site, and that the radiation channels monitoring plant effluents had not been properly calib.ated. Although spent nuclear fuel stored in the spent fuel pool remained adequately cooled, the licensee declared the spent fuel cooling system inoperable due to analyses that showed that service water cooling lines might not remain operable for certain design basis events, and due to corrosion induced defects in the service water piping. Licensee actions were in progress at the conclusion of the period to address concerns in these area Several changes in the licensee management organization occurred during the period, and a major organizational change was announced on March 11,1997. Mr. Russell Mellor was appointed to the new position of Director, Site Operations and Decommissioning, reporting directly to Mr. Ted Feigenbaum, Executive Vice President and Chief Nuclear Officer. M .
Mellor began his new duties in March 24,1997. Mr. Jere LaPlatney resigned from the position of Unit Director effective March 27,1997. Mr. Gary Bouchard was named as the Unit Directoc, with responsibility over the functional areas of operations, maintenance, health physics, security, chemistry and building services. Mr. Gerald Waig assumed the position of Operations Maneger, and Mr. Douglas Heffernan assumed the position of Maintenance Manager. Mr. J. Haseltine remained the Engineering Director with oversight over design engineering, engineering programs, corrective actions and the configuration management plan. Mr. Brain Wood was named the Business Manager, with responsibility for materials management, contracts, administration and finances. Mr. Noah Fetherston assumed the position of Decommissioning Project Manager, with oversight for cost and scheduling. Mr. Richard Sexton assumed the position Health Physics Manager effective on March 31,1997. Lastly, on January 22, the licensee announced that Mr. Bernard Fox would resign from the position of Chairman of the Board for Northeast Utilities effective on August 1,199 NRC activities at the site during the period also included the following visits and tours:
members of the NRR Decommissioning Projects Directorate including Mr. Michael Masnick, ;
Mr. Morton Fairtile, and Mr. Seymour Weiss toured the site on January 9,14 and February l 10,1997. Mr. John Rogge, Chief of Reactor Projects Branch #8 toured the site on ;
Jaruary 14,15, March 12 and 13,199 l i
 
I l
 
J
 
- - - - - .- - - - - - .- -.._ -. - - . - . . . -  _ _ _ . .- =.-
.
W
.        ,
;      2  !
 
l. OPERATIONS
 
j
         '
         '
01 Conduct cf Operations'
    *
t
9709000013 970731 PDR ADOCK 05000213 a  aa  =
'
qui.sppui.ssu..
Using inspection Procedure 71707, the inspector conducted periodic reviews of  3 plant status and ongoing operations. Operator actions were reviewed during
          ,
      . l periodic plant tours to determine whether operating activities were consistent with l
  ~ .  - - , . . . .m.-, ,- - , -..   , , - - - _ . - -
. the procedures in effect and the conditions of the license requirement .
          ,_,l
!        !
The purpose =of this inspection was to review the licensee activities to maintain the l i plant in the defueled condition, and to prepare for decommissioning activitie l 01.1 ' Review of Ooeratino Activities    ,
        !
Operating activities during this period included those operations needed to maintain i
'
stable plant conditions with the reactor defueled, to maintain adequate level in the 'l spent fuel pool, and to assure adequate cooling of the spent fuel. The inspector ;
reviewed the licensee controls over those systems needed to assure the adequate  j
;
  : cooling of spent fuel under design basis conditions, and to monitor the status of - l 4 -
radiological conditions at the plant and in plant effluents to the environment. The '
inspector independently verified that plant systems were properly aligned to satisf .
the license condition l The licensee maintained one service (SW) pump operating (pumps were rotated to  j
. equalize the run times) to support spent fuel pool cooling. One component cooling  j
  .wster pump, and one of two turbine building closed cooling water pumps remained  1
'
in service. . The normal and emergency electric distribution system remained in  ]
service (except for periods of testing and repair).to support spent fuel pool cooling i
.
and plant operations. A number of control board annunciators (approximately 90)  J
'
remained illuminated consistent with the plant shutdown and defueled. The inspector reviewed the status of each annunciator and the reason it existed, and . ,
determined that the conditions were normal for the shutdown and defueled  l
  . condition of the plant. Operator actions in response to off normal conditions were I reviewed and were found to be consistent with the respective annunciator i  procedur .2 Status and Control of Systems in the Defueled Mode (URI 97-01-01)
Following the end of refueling outage #19 with the completion of defueling, and the licensee shutdown plant equipment and placed systems in standby or layup
  . conditions for long term preservation. The licensee continued to implement
'
, procedure NOP 2.0-4, Layup of Systems and Components, which provided the
!  guidance for the general alignment and preservation of systems and components
 
'
 
        '
Topical' headings such as o1, M8, etc., are used in accordance with the NRC_ standardized reactor inspection report outline. Individual reports are not expected to address all outline topic !
  . , . ,. .  .  .


.
.
>
.
.
Mr. Ted cc: w/cv of Licensee's Resoonse Letter R. Bassilakis, Citizens Awareness Network J. M. Block, Attorney for CAN J. P. Brooks, CT Attorney Generals Office M. DeBold, Town of Haddam W. Meinert, Nuclear Engineer State of Connecticut SLO Distribution w/cv of Licensee Response Letter Region i Docket Room (with concurrences)
Nuclear Safety information Center (NSIC)
PUBLIC NRC Resident inspector R. Conte, DRP M. Conner, DRP C. O'Daniell, DRP W. Dean, OEDO S. Weiss, NRR, DRPM, PDND M. Fairtile, PM, NRR M. Callahan, OCA W. Travers, SPO R. Correia, NRR F. Talbot, NRR D. Screnci, PAO, ORA DOCDESK Inspection Program Branch, NRR (IPAS)
DOCUMENT NAME: G:\ BRANCH 7\REPLYLTR\HN97-01.RPY Ts receive a copy of this document, indicate in the boa: "C" = Copy without attachment / enclosure *E" = Copy with attachment / enclosure "N" - No copy OFFICE Rl/DRP    /    l NAME RConte DATE 7 /Jy97  / /97 OFFICIAL RECORD COPY


during extensive plant shutdowns and outages. The inspector independently verified that the reactor, reactor coolant system and the containment were maintained in a satisfactory condition for long term preservatio The Operations Manager issued a directive (memorandum ODM 96-157) allowing the manipulation of systems as specified by existing plant procedures for maintaining shutdown conditions, but restricting the modification of Mant operating configuration until a process was developed to assure the orderly transition from an operating statu Licensee actions continued during this period to develop a process to identify the proposed system changes necessary to support decommissioning, and which would assure the safety evaluations required under 10 CFR 50.59 would be don Additionally, the licensee began preparation of a proposed change to the technical specifications that would address the permanently defueled condition. The licensee planned to devise a means to deactivate or eliminate the annanciators that were not needed for the defueled condition, with a goal of achieving a black board again. In preparations for decommissioning, the licensee developed a procedure (ENG 1.7-6)
that provided a method to classify all plant systems into one of the following four categories: operable, available, lay-up, or abandoned. _Licentee engineering-personnel began a review during this period of all plant syste ms as identified on the list of critical plant drawings to classify plant systems according the criteria in ENG 1.7-6. The licensee recognized the need for a method to provide a clear display of plant system status on the main control boards as plant systems are classified and released from operation The inspector reviewed the licensee plans to provide for a smooth transition from an operating to a decommissionin0 status. NRC review of this area was in progress at the end of this inspection period. This item is considered unresolved pending the completion of licensee actions listed above, and subsequent review by the NR Specifically, this item covers the actions to (1) classify plant systems to identify which need to remain operable to assure maintenance of spent fuel pool cooling, and which can be modified or abandoned in accordance with 10 CFR 50.59 and removed permanently from service; (ii) devise and implement a method to provide a clear display of plant system status on the main control boards as nierit systems are classified and released from operations; and, (iii) to reduce or eliminate unnecessary annunciators to facilitate operator actions te monitor plant conditions and respond to off normal conditions (URI 97-01-01).
014 Scent Fuel Pool Coolina and Buildina Ventilation Inspection Scoce (71707)
The purpose of this inspection was to review the licensee activities to monitor the status of fuel stored in the spent fuel pool and assure the adequate cooling of spent fue _ _
.
.
.
4 Q)servations and Findinas The spent fuel ventilation and cooling systems were maintained functioning per normal operating procedures (NOP) 2.10-1 and 2.15-3 to keep the spent fuel cooled. The "B" spent fuel pool (plate) heat exchanger and at least one of the two spent fuel cooling pumps remained in service to keep fuel pool temperature below 150 degrees Fahrenheit (F). The spent fuel building enclosure and ventilation systern remained operable. The inspector reviewed licensee activitie.s to assure compliance with the following Technical Specifications (TS):
...
TS 3.9.11, SFP Water Level TS 3.9.12, Spent Fuel Building Air Cleanup System TS 3.9.15, SFP Cooling There were no activities during this period involving the movement of fuel or heavy loads over the spent fuel pool. The licensee maintained the boron concentration in the spent fuel pool at greater than 2500 ppm. The licensee conducted routine surveillance of the spent fuel pool and building, which included the tours by the nuclear side operators once each shift per SUR 5.1-0A. No inadequacies were identifie I Spent fuel pool temperature remained in the range of 80 to 85 degrees F, until I March 18,1997, when the pool was cooled down to 75 degrees F in preparation for work on the service water (SW) cooling lines (see Section 01.4 below for further details on the installation of an alternate cooling system). Spent fuel pool temperature increased to a maximum of 100 degrees when the alternate cooling system was installed on March 31. Operator observations noted a pool heatup rate of about 1.3 degrees F/ hour while implementing the bypass, which agreed well with the licensee prediction that pool heat up rate would be less than 1.7 degrees F/ hour. Stable pool temperatures were maintained in the range of 95 to 98 degrees F while on the bypass, with changes attributed to slight variations in nver j temperature and supply flo l 01.4 Inocerable Soent Fuel Coolina Svstem On March 11, the licensee declared the spent fuel uooling system inoperable at 6:01 p.m. based on an engineering evaluatior, that concluded the service water pipes providing cooling water to the SFP hr at exchangers might become inoperable due to postulated waterhammer events fahowing a loss of normal power event (LNP). The NRC review of the waterhamnier issue is provided in Section E1.1 of this report. Although inoperable per the technical specifications, the spent fuel cooling and service water systems remained functional to keep the fuel adequately cooled at all times. The licensee reported this event to the NRC on March 11 per 10 CFR 50.72(b)(2)(l) as plant operation in a degraded and unanalyzed conditio Technical Specification 3.9.15 requires that the spent fuel cooling system be operable during the storage of irradiated fuel assemblies from a full core offload in the pool. The specification requires that both spent fuel cooling pumps be operable
 
  .
  .
.
with at least one pump and the plate heat exchanger in operation. The heat exchanger was deemed inoperable due to the reliance on the service water system for cooling, the operability of the SW system could not be assured for a design basis event. Althouph the operability of the ultimate heat sink (including the service 1 water system) is m <,obed in TS 3.7.12, the requirements of that specification applied to operational modes 1 through 4, and were not applicable for operation in the defueled condition. The action statement for TS 3.9.15 with the SFP cooling system inoperable was to suspend operations to add irradiated fuel to the pool and to initiate corrective actions to restore the SFPCS to an operable status as soon as possibl The inspector reviewed licensee actions to stage equipment and revise procedures for alternate cooling of the SFP heat exchangers using fire hoses. The licensee revised AOP 3.2-59 (TPC 97-59) by adding Attachment 9 to describe the use of the temporary cooling method. The revised AOP was reviewed by the plant operations review committee (PORC) on March 13,1997. Procedure SUR 5.1-0A was also changed (TPC 97-56) to have the NSO stage equipment to implement the alterriate cooling if the pool temperature reached 120 degrees F. The use of fire hoses had been previously demonstrated to be operationally feasible in the Fall of 1996 when e a similar alternate supply had been established (reference Bypass 96-63 and Engineering Memorandum CYDE-96-0564).
The temporary alternate cooling method was technically acceptable in an October 1996 engineering evaluation, which showed that for a spent fuel decay heat load of ;
2 x 10 ** BTU /hr, the pool temperature could be maintained below the license limit j of 150 degrees F by supplying the pool with as little as 100 gpm of cooling flow supplied by 3 inch diameter fire hoses. The 1996 evaluation was updated in Safety Evaluation SY-EV-97 01, which was approved by the PORC on March 14,199 The updated analysis showed that for the current pool heat load following the full i core discharge (4.0 x 10 +' BTU / hour), the fire hosns supplying a minimum of 100 gpm was sufficient to maintain pool temperatures below 150 degrees F, assuming river temperatures at the design basis maximum of 90 F. The licensee installed one section of metal pipe in the temporary return side header to measure the cooling flow with an on line flow instrument (controlatron) while on the bypass. Service l water cooling flow remained in the range of 160 to 180 gpm. The inspector j verified that the fire hoses were consistent with the design specifications assumed in the safety evaluation (reference PONN Supreme fire hose specifications).
On March 26, the licensee declared the SW return pipmg imm the SFP heat exchangers inoperable at 7:15 p.m. as a result of a corrosion defect in the 6 inch carbon steel pipe. The pipe was inoperable based on an engineering evaluation that concluded that the pipe could not support design basis loads barod on an ,
engineering evaluation (see Section M1.1 - authorized work order 97-806) for l further details on this issue). The licensee reported this event to the NRC on March l 26 per 10 CFR 50.72(b)(2)(1) as plant operation in a degraded and unanalyzed l condition. The licensee approved a change to NOP 2.24-3 (TPC 97-60) which added instructions in (Attachment 7) to use the fire hose equipment previously staged for the March 11 event in the event alternate SFP cooling was needed on a
      )
l
.
.
.
CONNECTICUT YANKEE  ATOMIC POWER COMPANY HADDAM NECK PLANT 362 INJUN HOLLOW ROAD . EAST HAMPTON, CT 06424-3099 June 6,1997 Docket No. 50-213 CY-97-058 Re: 10CFR2.201 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Haddam Neck Plant Reply to Notice of Violation NRC Intearated Inspection Report No. 50-213/97-01 In a letter dated May 8,1997,N the NRC transmitted to Connecticut Yankee Atomic Power Company (CYAPCO) an inspection report covering the period from January 6, 1997 - April 7,1997, documenting three (3) Severity Level IV violations for the Haddam Neck Plant (HNP). As discussed in the report, the NRC Staff cited CYAPCO for violation of certain NRC requirements. The violations identified dealt with failure to follow procedures, failure to complete technical specification surveillances, and failure to complete effective corrective actions.


non-exigent basis. The licensee implemented the NOP 2.24-3 on March 31 to install the alternate service water cooling lines while actions were completed to i install the check valve and replace piping as needed due to corrosion induced defect Licensee engineering completed an assessment that was approved by the Plant Operations Review Committee on March 26 to evaluate the acceptability of the continued use of the normal service water supply and return piping. The licensee found it acceptable to continue to use the normal service water system to cool the SFP pending the completion of repairs. The assessment included considerations for the types of deficiencies (physical or analytical) postulated in the supply and return piping, the consequences of pipe failure, including postulated flooding if a line failed, and the additional measure to maintain the SFP area under augmented operational surveillance. The licensee switched to the alternate cooling method for the SFP heat exchanger on March 3 ,5 Imolementation of Alternate SFP Cnolina_
Pursuant to 10CFR2.201, Attachment 1 provides CYAPCO's response to the individual violations. The NRC Inspection Report 97-01 also requested information on how CYAPCO is addressing the issues of performance standards and procedure adherence at the HNP. CYAPCO recognizes that human performance issues and thus, performance standards, will play a key role in the safe and successful storage of spent
in preparation for maintenance work on the service water (SW) system that supplies cooling water for the spent fuel pool (SFP) heat exchanger including the addition of -
      -
a check valve in the supply line and replacement of a section of return piping with !
nuclear fuel and decommissioning at the HNP. The management team has initiated a concerted effort to raise the performance standards for all personnel on site and establish high standards for the decommissioning process. Included in this effort is reinforcing high standards for the core values of honesty, integrity and commitment to the job, including program and procedure ownership and adherence, safe working practices and a questioning attitude. These core values have been established in the management and supervisory team through group meetings and discussions. Group meetings to discuss these core values are also being conducted between senior site management and non-management / supervisory personnel.
reduced wall thickness, operations setup of fire hoses was reviewe I At4:48 a.m. on March 31,1997, operations isolated the SW to the SFP cooling I heat exchanger with the pool temperature at 75 F. This system isolation was l performed using temporary procedure change (TPC) 97-60 to NOP 2.24-3, Filtered 1 Service Water System and Adams Filter Operation. The TPC, PORC approved on March 27,1997, added Attachment 7 for the temporary supply and return of SW to SFP heat exchangers while work was in progress to install a new check valve. The plan was to connect two hoses from the PAB second floor Adams filter drain lines to the SFP heat exchanger inlet manifold check valves and have a single fire hose return the SW to the PAB ground level valve connection. This was the same hookup used earlier for other SW wor Maintenance removed the spool pieces and hooked up the previously used nozzle The OAS inspection plan called for a visual check for debris or foreign materialin the lines prior to operations hookup of the fire hoses. Step 1.8 was to open SW-V-643A and rod out pipe nipple and valve to remove any debris. Maintenance assisted operations in performing Step 1.9, Install 2" globe valve and 2 % or 3 inch l hose connection at open end of piping from SW V-643A. Someone questioned if QAS was to inspect the open pipe and work control was contacted. Work control i contacted QAS who insisted that the installed valve had to be removed prior to his inspection, so that was don The inspector noted that operations / maintenance f ailed to stop after Step 1.8 for the OAS hold point. However, there was no indication'in the procedure where the ;
hold point was and the OAS inspection plan never indicated the step number where '
the inspection was intended. The inspector considered this a poor plant work i
l
 
    .. .  . _ _ .- . _ _ _ _ -___
l
 
.
 
.
practice. In discussions with the inspector, QAS pointed out that communications  )
'
were less than adequate and agreed that indicating the step number on the QAS inspection plan was a good idea. Prior to the end of this inspection, QAS had  .
modified QASI-PS-CY-2.02A, Step 6.3.2 to include detailed instructions to indicate  '
the procedure step at which the [QAS] inspection will be performe '
Although this work was scheduled to be completed before the end of day shift  i March 31,1997, considerable delays, mostly due to personnel availability, were  l encountered and the SW cenectior through the fire hoses for SFP cooling was not restored until 8:04 p.m. The SFP water exiting the heat exchanger was 85 F and the pool temperature had increased to 94 F. With SW being supplied by fire hoses, the SFP bulk temperature decreased about 1 degree F per shif In summary, the licansee operations to supply temporary service water cooling to the spent fuel pool via fire hoses were adequate. QAS was responsive to improve their communications with operations / maintenance for QAS hold points. The spent fuel cooling system remained in a degraded condition with the plant in the action statement for TS 3.9.15 at the end of the inspection period. The licensee proceeded with engineering evaluations, modifications and maintenance activities in ;
n an expeditious manner to restore the spent fuel cooling system to an operable
, status. The modifications included the implementation of a design. change request (DCR) 97-002 to install a check valve in the SW supply line. The new check valve was welded in and acceptance testing was being performed at the end of the inspection perio .6 kgjitpf a 116 kV Distribution Line On March 13,1997, a problem in the offsite electrical distribution network resulted in the loss of redundancy in the 115 kV power supply for the Haddam Neck Site (adverse condition report ACR 97-130). Specifically, an electrical perturbation during the startup of an electrical generation station in Middletown, Connecticut  '
resulted in a plant trip. When the breaker that would normally operate to separate the plant from the 115 kV system failed to open, the offsite electrical protection scheme tripped a secondary breaker that de-energized Line 1572 at 6:00 a.m. Line 1572 is the normal supply to Middletown 1772, which is one of two feeds for the Haddam Neck station. Line 1772 remained energized by a feed from Haddam Line 1206 through OCB 389T399, which is normally closed to cross connect the offsite supplies in the Haddam Neck 115 kV switchyard. Thus, all plant buses remained  '
i energized throughout the electrical transient. The emergency diesel generators were operable at the time, but were not required to operate. There was no impact on plant operations or operating equipmen The Haddam Neck plant operators communicated with the offsite load dispatcher to identify the cause of the problem to the Haddam Neck offsite supply, the actions being taken to correct the problem, and the estimated time to recover Line 177 i Line 1772 was re-energized from its normal source at 9:10 a.m. on March 13. The event occurred as the licensee was taking measures in response to a plant design deficiency and associated analysis, which postulated that the service water system
 
-
.
.
 
providing cooling water to the spent fuel pool cooling system might fail due to water hammer loads associated with a loss of offsite power event. Plant operators responded appropriately to the challenge to the onsite power suppl .7 Inocerable Effluent Monitorina System Technical Specifications (TS) 3.3.3.7 and 3.3.3.8 require that certain radiation monitors be operabia at all times to monitor the status of liquid and gaseous releases from the site. The licensee declared all technical specification monitors inoperable on February 5,1997, as a result of an NRC inspection (reference Inspection 97-02) which found inadequacies in the calibrations program. NRC identified deficiencies in the procedure used to calibrate the gaseous and liquid effluent monitors, such that the accuracy of the monitors could not be assure The radiation monitors affected included liquid (RM 18 and RM 22) and gaseous (RM 14A) effluent monitors. Inspection 97-02 also identified deficiencies in the calibration of the wide range stack monitor, RM 14B, and in the spent fuel pool radiation monitor R 19. The inspector reviewed licensee actions this period in response to this issue. Although the detection channels were TS inoperable, all channels remained on line (except for periods of maintenance or testing) and available for operator use to monitor the associate pathwa The licensee completed compensatory actions to comply with the action statement of TS 3.3.3.7 and 3.3.3.8, which included the evaluation of releases via the associated pathways. The licensee obtained and analyzed periodic grab samples whenever effluents by the associated pathway were in progress. The inspector reviewed licensee compensatory samples and analyses on February 4, for the release of the A waste test tank. The release was processed in accordance with release permit L-10 and Action 46 of Technical Specification Table 3.3-9, which required that two independent samples of the tank be analyzed to assure it is acceptable for release, and that the release calculations and discharge valving be independently verified. No inadequacies were identified. The inspector verified licensee actions to implement the compensatory measures periodically during the
, inspectio All channels remained inoperable at the end of the inspection period pending completion of licensee actions to complete a proper calibration on the channel Inspection 97-02 and Section R2.1 of this report provide further details on this matte .8 Control of System Confiauration (VIO 97-01 -02.a)
The purpose of this inspection was to review the licensee process to control the physical configuration of the plant. This review included the implementation of the tagging process during the conduct of work activities, and the control of systems removed from service due to plans to decommission the plant. Licensee actions to issue and/or remove tags under the following clearances were reviewed:
 
.
 
* Spent Fuel Cooling Pump Preventive Maintenance (96-055)
* North Service Water Header (97-032)
* EG-2B Testing and Troubleshooting (97-043)
* A HPSI Pump Supply Breaker (96-748)
* A Charging Pump Supply Breaker (96-808)
* Service Water Containment Header (96-662,659,846,1097)
* Service Water SFP Header (97-108,130)
Except as noted below, no discrepancies were identifie Violation of Red Tan Controls On February 19,1997, during preparation for "A" prefilter changeout, a contracted health physics worker opened the prefilter access door with a Red Danger tag taped to the front of the door. Once the door was opened, the danger tag was not visible to other workers, so they also entered the prefilter area to continue the work. The danger tag warned of the automatic initiation of the carbon dioxide fire protection system, thus, this was a personnel safety issue. At CY, all tagging is governed by WCM 2.4-1, Equipment Tagging. This procedure clearly prohibits equipment tagged out shall not be manipulated in any way. The individual involved stated that he had received a pre-job briefing, knew that the system had been tagged out of service, and that the tagging walkdown had been performed. He said he saw the danger tag ,
on the door and assumed that it was one of the isolation tags for his wor j i
The licensee issued ACR 97-93 to describe and assess this event and to implement corrective actions. ACR 97-93 suggested that the danger tag should have been removed when the prefilter was isolated, the tag was not hung in a manner that j was clear to the workers, and the worker should not have opened the door. Plant management was active in the requirement that all managers review company policy regarding observance of all tags. The licensee looked at previous correctia actions and found they had been ineffective in that repeated tagging errors have continued over the past several years. They have had Station Staad Downs, where no work in a particular area or in all areas is allowed, provided tagging training, and held discussions with plant personnel. They conclude that personnel appear not to have realized the consequence of operating red-tagged equipmen The licensee's corrective actions for this event included: 1) operations sent a station memo to all station personnel emphasizing the importance of not operating red-tagged equipment; 2) the tagging group was to verify that any conflicting tags are removed before adding a new clearance to a piece of equipment: 3) all department supervisors were to verity that conflicting tags are removed before starting system work; 4) the training department was to change the computer-based training (CBT)
for station tagging to an annual cycle and make it mandatory; and 5) establishing different method of enforcing confined entry controls, through the installation of a sign on the ventilation door. The inspector reviewed the records provided by the licensee that showed that all departments have reviewed the CY Access Training Supplement: Equipment Taggin l .
 
i The inspector noted that the failure to observe red tag controls and the failure by contractor personnel to observe station work practices were recurrent issues at Haddam Neck, in December 1995, a contractor operated a red tagged 480 volt l breaker - reference Inspection 95-27, Section 3.3). The February 19 event showed that past licensee corrective actions were not effective. The failure to heed a posted Red Tag on February 19 was a serious event that could have led to plant or :
personnel damage. This was a violation of a plant procedures, and one of six i examples of a violation of Technical Specification 6.8.1. (VIO 50-213/97-01-02.a)
014 Inadeauate Control of Boiler Operations NIO 97-01 -02.b)
Several events occurred during the period in which licensee management identified l and addressed deficiencies in performance by plant personnel, including operator I Adverse condition report (ACR) 97-19 concerned the discovery on January 5,1997 i that a nuclear side operator had left a key ring unattended for 45 minute i Corrective actions included security checks of vital areas, counseling the individual, i
      '
and adolessing expectations with all plant departments. The licensee noted this was the third event of this type in the last two year Two events occurred which had minimal impact on plant safety, but which  l demonstrated poor operator performance. The events concerned the overflow o !
non-radioactive water from the heating steam condensate receiver tank, which provided feed water to the plant heating boilers, a non-safety related system. ACR l 97-24 concerned the inadvertent overflow of about 100 gallons of water from the !
condensate receiver tank on January 11. The event occurred because the nuclear j side operator (NSO) started a manual fill evolution of the condensate tank, and left the tank unattended while he continued other rounds. The operator chose to manually fill the tank to a higher level because of his concerns on the reliability of the automatic level controls based on past experiences. ACR 97-27 was issued on January 15 when the automatic makeup valve HC-MOV-491 A failed open while the tank was aligned for automatic makeup. The tank overflow to the floor drain was discovered by the NSO during routine rounds. ACR 97-74 concerned the overflow of the condensate tank on February 9 when an NSO (different NSO than the January 11) began a manual makeup and then left the tank unattended as he continued other ror.nd For each event, the licensee took action to counsel the operators involved and to review management expectations on performance with plant operators. The  l licensee also reviewed the spills .o sssure no release limits were exceeded. The boiler activities were generally governed by procedure NOP 2.19-8A, but after initial boiler startup, the condensate tank fill evolutions were generally conducted as a
" skill of the trade." The licensee had recognized the need to improve the controls in NOP 2.19-8A and added requirements in Revision 3 dated January 31,1997 for the I operator to stay in attendance whenever manually filling the tank. However, Step 6.1.12a requires the operator to maintain the condensate receiver tank one-half to two-thirds full when operating the condensate makeup manually. Collectively, the ' _ . .. _ . _ -. _  _ _ _ , . . _ _ _ _ _ _ - . . . _
..
*
          [
'
 
i events demonstrated poor operator attention to duties, poor licensee control of i
processes, and inadequate corrective action to correct deficiencie !
I          ..
The failure to properly fill the condensate tank was contrary to NOP 2.19-8A; and    !
,
was the second of six examples of a violation of Technical Specification 6.8.1 (VIO  l 97-01-02.b).        i
,          .
01.10 Conclusions for Conduct of Operations l
-
The conduct of operating activities was acceptable, as were the operator actions to
'
maintain the plant stable in a defueled condition and to monitor the status of systems in layup of long term preservation. ' An inspection item will follow licensee  1 initiatives to better control the status and categorization of plant systems in post operations mode, and to reduce the number of illuminated annunciator Although the spent fuel remained adequately cooled at all times, a discrepancy in
.
the design basis and deficiencies in the material condition of the service water    i system created a challenge to the adequacy of the spent fuel cooling syste ;
          '
Licensee actions to address these deficiencies, as well as the operator actions to -
; y ' implement alternate cooling methods for the spent fuel pool, were acceptabl [
Jnadequacies in the procedures used to calibrate certain radiation monitors resulted
'
l 1 in the declaration of e!.I gaseous and liquid effluent monitors required by the    ;
          '
technical specifications to be inoperable, and the need to implement compensatory measures to monitor effluents by the associated release pathways.
 
I -
As in past inspections, human performance errors by operators and workers    i detracted from good performance. The violation of procedures resulting in the  -
gperation of red tagged equipment was a significant event that could have resulted    i in unsafe conditions. Several other examples of a failure to adequately follow plant  !
procedures were note Operational Status of Facilities and Equipment
: 02.1 Cold Weather Preparations
. Inspection Scope (71714)
,
The purpose of this inspection was to review licensee actions to assure plant    ,
systems were adequately protected from freezin Observations and Findinas The licensee had a program to assure systems were protected from cold weather    !
conditions. The plant design features and processes remained as noted by past    i NRC inspections in this area (reference inspections 93-22 and 96-01), which  '
included heat traced circuits for process piping and tanks, and administrative controls to assure the freeze protection circuits were maintained and operating. The  ;
administrative controls included a walkdown by maintenance personnel of all    j


          ;
(1) NRC letter from J. F. Rogge to T. C. Feigenbaum, "NRC Integrated inspection Report No. 50-213/97-01" dated May 8,1997.
i l
_ _ _ _ _ _ _ . . _ _ _ . - - _ . _ .~.. _ _ - - _ - - _ , . .--


      - . . _ _ _ _ _ _ _ _ _ _ _ _ _ .
.
.
circuits per procedure PMP 9.9-146 to verify the condition of heat trace circuits, and identify and repair deficiencies; and, the operator checks of heat trace panels and circuits per operations department instruction ODI-146. Additional controls in this area included periodic reviews by the quality assurance group, such as the QAS surveillance (CY-96-090) cornpleted on December 13,199 Problems in adequately implementing the administrative controls were identifie The QAS surveillance noted a weakness in the planning and implementation of the annual preventive maintenance on some freeze protection circuits (adverse condition report ACR 96-1351), and problems establishing the proper temperature setpoint for a circuit on the demineralized water storage tank (ACR 96-1350). These issues were reviewed by plant management and corrective actions were initiated to address the discrepancie Despite the preparations and controls for cold weather, the licensee actions were not sufficient to preclude damage to a plant process line. The licensee identified on January 3,1997 that a 3/4 inch service water line providing filtered water to a steam generator (SG) sample cooler had burst due to freezing. Actions were taken to isolate the leak and to report the deficiency (ACR 97-002). The leak did not
. jeopardize cooling to important plant systems. The SG sample cooler was not required to be operable due to the shutdown status of the plant. The followup investigation determined that the heat trace circuits were energized, but damaged and wet insulation had resulted in a freeze condition in the lin The licensee followup actions were to review the status of insulation on other lines and address other deficiencies in the freeze protection circuits. The action plan and summary of actions taken were provided in memorandum CYDE-97-003 dated January 8,1997, and in a memorandum to the Maintenance Manager dated January 22,1997. Based on previous industry operating experiences, the licensee was sensitive to the protection of the spent fuel pool and provided enhanced visibility and monitoring of building temperature around the boundary. Supplemental heating was used in some plant areas (pipe trenches). A longer term action included the plan to review the basis for plant freeze protection after completing the system categorizations necessary to decommission the plan c. Conclusions The licensee implemented established procedures and controls to assure plant systems were adequately protected from cold weather conditions. Important plant systems, including the spent fuel pool and support systems and the containment, were adequately protected from freezing. However, the failure to complete thorough plant walkdowns and poor insulation conditions in some process piping resulted in freeze damage to a process line. Licensee actions to respond to this  i event were acceptabl i
l
  . __  ._ ___ _ +
l
l
*
        \
,
-      l 13    1
4 03 Operations Procedures and Documentation    l I
l 03.1 Procedure Quality for Shutdown Operatina Activities  ;
        ;
        ' Inspection Scooe (42700)
.
The purpose of this inspection was to review plant procedures governing activities to support testing, maintenance and operation of plant in a shutdown and defueled conditio ; Observatiens and Findinas This inspection was performed in response to NRC concerns raised by an operational event (reference NRC Inspection 96-80), in which some procedures for l'
"
shutdown operations were inadequate. In response to that finding, the licensee initiated actions to rewrite procedures and improve procedure quality. The licensee ,
completed a major effort to upgrade shutdown and refueling procedures prior to the {
resumption of activities to offload the core from the reactor coolant system to the l spent fuel pool. The core offload was completed on November 15,1996. NRC  l j , review of the initial procedure upgrade effort was described in inspection 96-11,-
which found that, generally, procedure quality had improved because of the licensee initiatives completed prior to the core offloa The list of procedures reviewed during this period is provided in Attachment 1, and included those procedures already in effect for plant operation, test and maintenance activities, along with new procedures or procedure revisions that the licensee found to be necessary to address deficiencies. The inspector found that
< the procedures reviewed conformed with the requirements of the technical specifications and met the guidance of Regulatory Guide 1.33. The procedures were technically adequate and useable. The inspector noted minor discrepancies (such as typographical errors, or enhancements in cross referencing or wording) that could be attributable to the lack of attention to detail in the preparation and approval of procedures. No procedure deficiencies were identified that would have resulted in an unsafe condition for the plant or plant personnel.
The overall controlling document for procedure preparation and approval remains Administrative Control Procedure (ACP) 1.2-6.5A, Station Procedures. Temporary Procedure Change (TPC) 96-802 added a new Section 1.9.3 that instituted a "Do Not Use" process to remove unneeded procedures when the related systems are
,
permanently removed from service. This TPC was incorporated into ACP 1.2-6.5A, Revision 1, effective March 7,1997. Instead of a blanket cancellation of unneeded procedures, CY's plan was to replace the procedure in the controlled manuals (control room, work planning and control, administration, and applicable departments) with a signed and approved Do Not Use (DNU) notice sheet as each procedure comes up to its biennial review date. Of the approximate 1100 operations procedures, about 200 have been so marked at the end of the inspectio j
  .  -    .
    .__  _- . . .__ _
.
l
.
k
For the Emergency Operating Procedures (EOPs), all emergency responso procedures (Reactor Trip, Safety injection, Steam Generator Tube Rupture, etc.)
were labeled DNU when EOP 3.1-0, the controlling procedure, was revised. In addition, EOP 3.1-12, Emergency Boration, and EOP 3.1-21, Pressurizer Spray Valve Malfunction, have been labeled DNU. EOP 3.1-48, Loss of Refueling Cavity :
       '
       '
Inventory resulted from an NRC commitment and will be handled separetely late All other EOPs are being converted to Abnormal Operating Procedures (AOP) as shown belo EOP 3.1-10 Partial Loss of AC  AOP 3.2-68 EOP 3.134 Complete Loss of Control Air  AOP 3.2-66  ,
_
EOP 3.1-46 Total Loss of Semi-Vital Power  AOP 3.2-65  l EOP 3.1-49 Partial Loss of DC  AOP 3.2-67 EOP 3.1-50 Loss of MCC-5  AOP 3.2-64 The AOPs not needed any more are being labeled DNU again as each procedure comes up to its biennial review date. At the time of the inspection,12 of the 41, -
D4Mr4J2/LL app -   .
not including the new ones listed above, had been so labeled. The inspector l reviewed the AOPs labeled DNU and found no improperly labeled procedure The other procedures under operation's control, ACPs, Normal Operating Procedures (NOPs), Preventive Maintenance Procedures (PMPs), Surveillance (SUR),
Annunciator (ANN), Work Control Manual (WCM), and some Administrative (ADM) l
      '
procedures, will be classified DNU, if no longer needed, at its biennial review date.


' For a listing of procedures reviewed by the inspectors, see Attachment Conclusion This review found that procedure quality was acceptable, and that licensee initiatives to improve, control and maintain procedures in accordance with the ;
. _ _ _   .
technical specifications were satisfactory. The licensee's approach to remove l I
  -     t
unneeded procedures from controlled manuals as each procedure approached its required review date was acceptable. A notable example of poor procedure quality were the findings relative to the proper calibration of the radioactive effluents instrumentation channels, as described in another NRC inspection that occurred during this period (reference inspection 97-02). The inadequate calibration procedures resulted from licensee activities prior to the plant shutdown in July, 1996, and prior to the subsequent licensee initiatives to improve plant procedure Operator Training and Qualification 05.1 Inaccurate Ooerator Trainino Records (URI 97-01-03) Insoection Scoog
.      i The purpose of this inspection was to review operator training records and the l management response to the discovery of deficiencies in the program to maintain !
qualification records for personnel who applied for an NRC license l l
 
_ - .
-      ,
l l
.
l 15    l b. Observations and Findinas
 
On February 7,1997, the licensee identified deficiencies in the program to maintain I qublification records for personnel who applied for an NRC license to operate l Haddam Neck. Following the failure by severallicensee personnel to pass an NRC licensed exam on Millstone 1 in December 1996, Northeast Utilities initiated an Independent Review Team (IRT) to investigate the cause for those failures. The IRT identified deficiencies in the process to train and qualify the candidates for the ,
exam. Based on these findings, NU expanded the scope of the IRT review to j
      '
include a review of the most recent operator licensed program for Millstone 2, Millstone 3, and Haddam Neck. The last licensed operator program for Haddam l Neck was completed in August 1996, in which all 12 of the candidates who  I completed the examination process received a license. The IRT began a review of the training records for the 1996 class in January 1997, and reported its preliminary findings to the licensee (CYAPCo) on February 6,1997 in Adverse Condition Report (ACR) 97-6 The IRT found that, based on a review of the required training for the Haddam Neck 1996 Licensed Operator Upgrade Training (LOUT) and Licensed Operator Initial ec Training (LOIT) programs, the licensee submitted inaccurate information to the NRC on Form 398 in support of an application for a license under 10 CFR 50.55. The form is required to be submitted by the licensee prior to the candidate being examined by the NRC. One purpose of the form is to certify that the individual has i successfully completed the facility licensee's requirements to be licensed as a reactor operator or senior reactor operator. The IRT found that three of the LOUT students did not meet the program requirements for standing watch while under instruction for the required number of hours (at least 320). Other deficiencies were identified, including the failure by some candidates to complete the required number of reactivity manipulations, and the failure of one candidate to complete program prerequisites for on the job training. 10 CFR 55.9 requires that information provided to the NRC by an applicant for a license shall be complete and accurate in all material respect There were 12 candidates who completed the 1996 licensing class. As of February 7,1997, nine of these 12 operators were not using their licenses at the facility because of transfers to other facilities as part of the licensee's staff reductions for ,
the plant decommissioning. Two licensed operators were on the operations  l department watch bill, as was a licensed senior reactor operator. The SRO subsequently left Haddam Neck for a position at another facility. The Operations !
manager issued a directive on February 13,1997 (memorandum ODM 97-018)
stating that none of the 12 operators from the 1996 class could perform licensed duties until the training program discrepancies were resolved. The two operators who remained at Haddam Neck were allowed to remain on the watch bill and allowed to perform non-licensed dutie The licensee formally notified the NRC of these deficiencies in a letter dated March 3,1997, which also described the actions taken to investigate this matter, identify the causes, and to correct the deficiencies in the operator training program and its
 
  .
.
.
 
administration. The licensee classified the deficiencies in ACR 97-63 as a severity level "B" issue that required a root cause investigation to determine why the deficiencies occurred. Additional corrective actions included plans to rectify the deficient information submitted to the NRC, review the matter for reportability, expand the IRT reviews to include the training for previous licensed operator classes, and complete a review to assess whether similar problems existed in the operator requalification program. The licensee committed to conduct additional reviews of the licensed operator requalification program by April 4,1997. The NRC formally recognized the licensee's corrective action plan in a confirmatory action letter issued on March 7,199 On March 27,1997, after review of the training records from the 1995-1996 licensed operator initial training , program for three individuals, the Haddam Neck Operations Manager concluded that, although there were several administrative errors, all three individuals successfully completed all requirements of the training program. Thus, the Operations Manager removed the restrictions from the conduct of licensed duties for all three operators (ODM 9'7-054, 055, 056).
 
l On April 3, the licensee completed the review of the licensed operator 4- requalification training (LORT) program and identified deficiencies similar to those in the LOIT (ACR 97-166). The problems included deficiencies in administering the LORT, and the failure of candidates to attend all LORT training sessions. The :
Operations Manager reviewed the audit results against the operators still holding i active licenses at Haddam Neck, and identified on April 9 one operator who might not have completed all the LORT program requirements. This individual was not l presently assigned to the control room watch bill, (ACR 97-180), but was removed j from licensed duties pending a complete review of actions to meet the LORT programs requirement !
Finally, a licensed operator identified to his supervision that he may not have met I the reactivity manipulation requirements of the LOIT (ACR 97-190). The Operations i Manager issued a memorandum on April 10 to remove this individual from licensed duty pending a full review of the actions to meet the LOIT program requirement The individual involved noted that although he was in the control room and in l attendance for reactivity manipulations, he participated in some exercises as an observer and did not actually perform the reactivity manipulation. The deficiency in ACR 97-190 was significant in that it was not identified by the IRT review of the individual's participation in the 1996 LOIT. When determining whether licensed operator candidates had completed the number of reactivity manipulations required by the training prograra, the IRT had relied on a review of training documentation, but had not interviewed the candidates. The licensee was reviewing this finding for generic implications for both the CY and Millstone training program NRC review of this area was in progress at the completion of this inspection perio This matter is unresolved pending the completion of licensee actions as required by the March 7 Confirmatory Action Letter, and the completion of the NRC reviews of this matter (URI 97 01-03).
 
.
.
17 Conclusions Poor performance was demonstrated in the administration of the licensed operator training program, and to assure that operators a fully qualified initially and remain qualified through periodic retraining. The NRC and licensee review of Haddam Neck licensed operator training program was in progress at the end of this inspection perio Operations Organization and Administration 06.1 Staffina and Control of Overtime (VIO 97-01 -02.c) Insoection Scoce (71707)
The purpose of this inspection was to review the licensee's actions to organize and reduce operations staffing to support the decommissioning mode, and to control work hours but critical plant staf l Observations and Findinas    l l
The operations department staffing was reduced by about half during this inspection !
"iod as a result of the decision to decommission the plant. The licensee also (
.iounced the selection a new Operations Manager effective on February 24, I a 997. Starting in January 1997, the operations department was comprised of i about 30 personnel, with the shift operators organized into 5 four-person crews on a rotating shift schedule. Each shift (labeled Crew B through F) had 1 senior reactor operator / shift manager (SM),1 reactor operator / control operator (CO), and 2 non-licensed operators / nuclear side operators (NSO). Additional personnel were assigned to the work crew (tagging), a procedures and training group, and for administrative support. By letter dated February 19,1997 (B16196), the licensee notified the NRC of the nineteen licensed operators who would discontinue licensed l duties at Haddam Neck due to transfers to jobs outside of the statio Technical Specification 6.2.2 establishes the facility staffing requirements and establishes the minimum shift crew composition for operational modes 1 through The minimum staffing for a defueled condition was not specified in the current license conditions. The licensee maintained minimum shift staffing in accordance with the requirements for Mode 6 - refueling. The licensee maintained a five member fire brigade, as required by Section II.1.6 of the Technical Requirements Manual. The fire brigade was comprised of the control operator as fire brigade leader, the two NSOs and the health physics and chemistry representatives assigned to each shift. Further, as a result of the deficiencies identified in the licensed operator training program (discussed in Section O.5 above), the Operations Manager issued a memorandum on February 13,1997 limiting two licensed operators to non licensed duties. The inspector verified that the licensee met the intended shift manning requirements despite the above constraints. The licensee intends to address minimum shift staffing requirements for the defueled condition in a formal revision to the f acility technical specifications; the licensee plans submit
 
.
.
 
the proposed TS revision in April 1997. The present shift staffing levels would meet the proposed TS requirement Technical Specifications 6.8.1 and 6.2.2.f require that administrative procedures be developed and implemented to limit the working hours of facility staff who perform safety related functions. These procedures should follow the general guidance of the NRC Policy Statement on working hours (Generic Letter No. 82-12). The licensee implemented procedure NGP 1.09 to meet the requirements of TS 6.2.2.f, which provides limits on the use of overtime that meets the guidance in the generic letter. During this period, the licensee determined that the NGP 1.09 requirements for the use and approval of overtime were not met in two occasions. The first instance was documented in adverse condition report (ACR) 97-64, and concerns the discovery that a licensed operator exceeded the guidelines during work shifts on February 5-6,1997. The operator worked two shifts with less than eight hours off in-between shifts. NGP 1.09 states that personnel must have prior approval to work with less than 8 continuous hours off between scheduled work periods (including shift turnover time). The operator was not approved to work in excess of the guidelines because the operations supervisor who scheduled the work was not aware of the procedure limits requiring that shift turnover time be included in the 8 ;
. hours of time off from werk. The procedure violation was discovered (after the l fact) by the operator assigned to do the work. Licensee corrective actions were l described in memorandum ODM 97-20, which included counseling the supervisor j who scheduled the work, and reiterating the NGP 1.09 requirements to all shift ;
managers. The inspector reviewed the use of overtime by operations (union)
- personnel for the week ending February 1,1997, which included a summary of all overtime worked in 1997. Although overtime was scheduled on a regular basis, the amount of overtime worked either weekly or for the year-to-date (YTD) was not excessive. The overtime worked ranged from none to a high of 13.5 hours for the weekly summary, and ranged from none to a high of 26.5 hours for the YTD category. There were no instances (except as described in ACR 97-64) where Gie overtime worked required approval per NGP 1.0 The second instance of overtime limit violations was described in ACR 97-33, and concerns the findings by the nuclear safety and oversight (NS&O) organization. The NS&O group conducted an investigation in response to a concern transmitted to the NRC regarding the control of work inside the containment at Haddam Neck during a forced shutdown in July 1994 following a fire in a reactor coolant pump (reference NRC inspections 94-17 and 94-18). The concern provided to the NRC indicated that a Health Physics Technician worked eight straight days without a day off, and on the eighth day after ten hours of work, the technician was required to work additional excessive hours. The licensee summarized the results of the investigation of this matter in a letter to the NRC dated January 17,1997 (B16099). The licensee concluded that there were six instances for which overtime authorization records could not be found for some of the Health Physics and Instrumentation &
Controis personnel involved in the 1994 event, and that the TS limits for work in excess of 72 hours, or for working 24 hours in a 48 hour period, had been exceeded. The following specific violation of the NGP 1.09 and TS 6.2.2.f limits were noted:
 
-  .. . - , - . -_- . .
.
19   '
* For the week ending 7/30/94, Individual A exceeded the limits for 16 hour of work in a 24 hour period, and for working 24 hours in a 48 hour period;
* For the week ending 8/13/94, Individual A and Individual B exceeded the limits for working 72 hours in a 7 day period; and,
* For the week ending 8/20/94, Individual C and Individual D exceeded the limit for working 72 hours in a 7 day perio i i
The licensee's investigation found that many other personnel worked overtime in ;
-
excess of the guidelines following that event, but this work was authorized as l required by NGP 1.09. In response to this matter, and as described in the followup !
to ACR 97-33, the licensee's corrective action included plans to address this issue
      '
,
generically by reiterating the importance of observing overtime limits in  ]
supervisory / manager safety meetings with station employees. The failure to control ,
and approve the use of overtime in 1994 and in 1997 in accordance with the  i administrative requirements of NGP 1.09 was a failure to follow procedure NGP 1.09, and was the third of six example of a violation of Technical Specification 6.8.1, and a violation of TS 6.2.2.f (VIO 97-01-02.c). Conclusions The licensee completed a reduction in the operations staffing in response to the decision to decommission the plant. The licensee maintained shift staffing sufficient to mrt minimum staffing levels for the number of licensed operators, and to meet the iire brigade requirements. NRC review of the adequecy of the reduced resources to administer the operations procedures work load was in progress at the conclusion of the inspection. Weaknesses were noted in the licensee controls to-assure that work hours for operators and other personnel who perform safety related work functions were limited in accordance with guidelines on the use and
<
approval of overtim Previous Operations Open issues (92901)
08.1 (Closed) VIO 94-21-01. Inadvertent Boron Dilution During inspection 94-21, the inspector reviewed a reactivity anomaly that resulted from operators failure to obtain the required chemistry samples of demineralized effluent following a 17 minute flush to ensure equilibrium. The boron concentration was lower then the primary and about 270 percent millitho (pcm) of reactivity was added as a result of this operation. This was a violation of Technical Specification 6.8.1. By letter dated November 15,1994, CYAPCO responded to the violation delineating the corrective actions taken. Due to the December 5,1996 notification to permanently cease operation of the Haddam Neck facility pursuant to 10 CFR 50.82(a)(1)(1), the possibility for reactor reactivity anomalies no longer exists. This issue is considered close l
 
.
.
 
00.2 (Closed) URI 94-27-01. Loss of Electrical Seoaration During inspection 50-213/94-27, the inspector questioned the practice of cross-connecting motor control centers (MCCs) between redundant safety trains. MCC-3 and 4 received power from redundant 480 volt buses 4 and 7; MCC-8 received power from redundant buses 5 and 6. Buses 4 and 5 are powered from emergency diesel generator EG-2A; buses 6 and 7 are powered from EG-2B. Abnormal operating procedure (AOP) 3.2-28, " Locating 480 Volt AC Grounds," allowed shutting a cross-tie breaker on MCC-3, MCC-4, or MCC-8. Although the AOP required operators to enter technical specification (TS) action 3.8.3.1.1.a (with an allowed outage time of 8 hours) when the cross-tie breakers were shut on MCC-3, 4 or 8, shutting these tie breakers for ground search investigations cross-connected redundant safety trains for several minutes. The licensee revised AOP 3.2-28 on i December 29,1994 to preclude closure of the MCC 3,4 and 8 tie breakers during I ground isolation. Operator logs for 1994 showed five instsnces where grounds were identified on 480 volt buses. In one case, Ju.i 10,1994, the operators shut ;
the manual cross-tie breaker for MCC-3 for four minutes. For the other cases, the l licensee did not close the cross-tie breakers on the specified MCCs. This issue was l left unresolved pending NRC review of the proposed LER, and verification of i additional corrective action By letter of January 25,1995, CYAPCO forwarded LER 50-213/94-29-00, Motor Control Center Tie Breakers Prohibited by TS. This LER says the root cause was a l procedural deficiency due to an incorrect interpretation of the action statement for TS 3.8.3.1. Corrective actions taken include the above-mentioned procedure revision and a change to Administrative Control Procedure (ACP) 1.2-6.5A, Station Procedures, to require that, as part of the biennial review, all TS Action Statements
  . are clearly identified and applicable to the Limited Condition for Operation. The -
l inspector found these corrective actions acceptable, therefore, this unresolved issue is close ,3 (Closed) LER 94-011-00. Unolanned Loss of Soent Fuel Coolina During inspection 50-213/94-14, LER 94-011-00 was reviewed. The event related to failure of the "B" spent fuel pool (SFP) cooling pump seal. Inspector review of the LER noted that the licensee did not document the root cause for the misalignment and seal f ailure of the "B" spent fuel pool cooling pump. Additionally, the inspector did not identify any maintenance controls for the scheduled preventive maintenance duration to assure that the bulk pool temperature would not exceed 166 F. Therefore, the LER remained open for inspection of these two issue In Final Safety Analysis Report Change Request (FSARCR) 96-CY-27, approved November 7,1996, CYAPCO proposed updates to Sections 7.6, All Other Instrumentation Systems Required for Safety, and Section 9.1.2, Spent Fuel Storage. This FSAR update greatly improves the description and analyses for the spent fuel pool syste __ - _ _ _ _ _ _ _ _ _ _ _ _ _ __
  .
  .
  -
  -
o    ,
in response to inspector questions, the licensee provided plant information report (PIR) 94-079, Spent Fuel Pool Pump Seal, and a September 22,1994 PIR 94-079 Followup Memo This followup memo indicates that the components of the failed SFP pump revealed severely overheated and damaged bearings, shaft, shaft seals, and mechanical seal. The amount of oil in the reservoir was found to be lower than recommended and it showed signs of being burnt. The historical data section states that, both the "A" and "B" SFP cooling pumps are of similar design (however, supplied by different vendors and of different sizes] and have had numerous failures over the last few year In discussions with the past SFP system engineer, the inspector learned that the
"A" pump old shaft seal was replaced with a new labyrinth type shaft sealin 1994, that a "B" pump breaker tripping problem was resolved in 1995, and that a failed
"B" pump discharge check valve was relocated downstream to reduce vibration and replaced with a new check valve in 1996. Review of maintenance records indicate repetitive problems with bearing tube oil adequacy and leakage have occurred over the years. The engineer explained that this was due to the critical " bubbler" level that controls the amount of oil around the shaft. Operations and maintenance have learned to keep the bubbler set correctly. It was also stated that the unbalanced SFP pumps supplied by different vendors and of different size did not perform well  ;
s together. An Engineering Work Request (EWR) had been submitted to System Engineering to purchase replacement pump or upgrade existing pump to increase capacity on May 25,1995. This was being tracked as 95-SE099. However, this EWR was not funded and has been dropped. The engineering management
. committed to re-review the EWR recommendations based on the critical need for the SFP system even when the Nuclear Island concept is adopte '
Since the Maintenance Rule Functional Failures (MRFF) criterion allows three failures in a rolling 24 month period and the check valve failure was found during normal preventive maintenance, the licensee counted only one MRFF of the SFP cooling system (the 1994 failure of the "B" SFP cooling pump seal). Thus, the SFP system has been categorized as a 10 CFR 50.65(a)(2), not requiring special monitoring system. The inspector indicated that further review of MRFF performance would be performed at a later time during the NRC Maintenance Rule Inspection required at all plant .4 (Closed) LER 94-015-01. Main Steam Valves Exceed Lift Setooints On June 16,1994 during setpoint verification testing, the licensee determined that four main steam (MS) safety valve (SV) pilot valves were outside the TS 3.7. allowed setpoints. The two pilot valves for MS-SV-24 lifted at a higher value then allowed and were declared out-of-service. The licensee's root cause was seat adhesion resulting from similar materials being used for both the pilot disc and nozzle. A similar problem with five active pilot valves associated with the four main steam valves was reported in LER 94-006, dated March 23,1994. The immediato corrective action was to readjust the setpoints. Long term corrective action was to evaluate possible replacement of the pilot valves with new type _-. --  . . .. . .  . -- - . . .-
-        >
        :
P
.
22  ,
The LER 94-015-01 supplement was submitted on March 2,1995, to provide the
'
results of corrective actions to date. CYAPCO, working with the oilot valve vender, found that dissimilar materials (such as inconel 718 and Cobalt MP-35N) performed acceptably on bench tests. The licensee replaced the eight pilot valves with the improved design in July 1994. This LER is close .5 (Closed) IFl 96-08-01. RHR Calibrations and Leakaae
        !
During inspection 50-213/96-08, operators experienced a delay in reaching cold
,
shutdown due to an initial inability to meet a limit specified in procedure NOP 2.9- '
This NOP was revised (by temporary procedure change 96-387) to increase the heat exchanger bypass flow (from 500-1000 gpm to 1000-1500 gpm), and to allow the use of hand held temperature gauges to compensate for a suspected calibration problem with the permanently installed instruments. The cooldown to Mode 5 was completed within the time limits specified by TS 3.0.3. The licensee initiated adverse condition report (ACR) 96-790 to address questions on the leakage of FCV-796 and instrument calibrations. Due to the decision to permanently cease-operation of the Haddam Neck facility, the leakage of FCV-796 and instrument
'  calibrations is no longer important. This issue is considered close ,6 LClosed) LER 95-023-00. Failure to Prepare Special Report On December 29,1995, the licensee determined that a Special Report should hava been issued within 90 days, in accordance with TS 3.5.1, for charging injection
  : flow into the reactor after manual initiation of safety injection during the manual l  reactor trip on July 27,1995. The manual trip of the reactor and turbine was initiated following steam flow / feed flow mismatch alarms after the "B" main feed pump motor breaker trip. -The initiating event, according to LER 95-016-00, dated August 22,1995, was a ground on 4160 volt bus "A."
,
,
The licensee provided a copy of the required Special Report letter, dated March 5, 1996. This letter recaps the event and states that the cause of the failure to report
U. S. Nucisar R::gulatory Commission
. was misinterpretation of what constitutes " injection flow." Originally, the plant staff concluded that the safety injection system did not actually inject water because the reactor pressure remained above the shutoff head of the high and low pressure safety injection pumps. However, the charging pumps, running during power operation for normal makeup and RCP seal water injection, did switch suction from the volume control tank to the refueling water storage tank (RWST).
: ,
CY-97-058/Paga 2    ,


Since the RCS pressure dropped to 1740 psig, the charging pumps with a nominal 2700 psig discharge pressure, did inject RWST water into the reactor. The licensee analysis for this LER was acceptable. This LER is close ,7 (Closed) URI 96-201-10. Alternate Auxiliarv Feedwater Sources During inspection 96 201, concerns regarding demineralized water storage tank makeup flow path configurations, transfer flow rate capabilities, as well as the
-
,  operator's ability to establish emergency procedure directed makeup flow paths before the onset of turbine-driven auxiliary feedwater pump cavitation were
The management team has worked with their individual department staff to establish department level standards that reflect these higher core values. All departments have
      . .
, completed draft standards and expectation documents that have been reviewed by l senior site management to assure consistency with the core values. Each of the :
.
department standards and expectations addresses the expectation of adherence to site !'
 
procedures. Site-wide and department level standards and expectations will be
expressed. However, due to the decision to permanently cease operation of the Haddr.m Neck facility, these concerns are no longer important. This issue is considered close .8 (Closed) LER 96-015-00. Containment Air Monitor Trio Valve On July 27,1996, during local leek rate testing (LLRT) in cold shutdown, the 3/4 inch outboard containment air monitor isolation valve, VS-SOV-12-1, failed its Type C test. The actualleak rate could not be quantified and was assumed to be greater than allowed because of the inability to pressurize the test boundary. The in-series inboard containment isolation valve, VS-TV-1848, was operable and passed its LLRT. After removal, it was discovered that the rubber valve disc was separated from the stem. The cause of the separation was incorrect assembly of the valve's internals during manuf acturing or installation. The licensee was to replace this valve with a different design during the refueling outage. However, due to the decision to permanently cease operation of the Haddam Neck facility, this was never don The inspector determined that the requirements of TS 3.9.4, that each penetration
; published by June 30,1997. This process of ensuring consideration of the input from all site staff will help to ensure site-wide ownership of these core values.
. providing direct access from the containment to the outside shall be closed off, was met during the final core offload by having the VS-TV-1848 isolation valve close Thus, this issue is closed, ll. MAINTENANCE M1 Conduct of Maintenance Using Inspection Procedure 71707,61726 and 62703, the inspectors conducted periodic reviews of plant status and ongoing maintenanc M 1.1 Maintenance Observations Insoection Scope (62707)
The inspectors observed all or portions of the following work activities:
* PMP 9.1-31, Diesel In-Leakage and Fuel Oil Transfer Pump Availability
* PMP 9.2-20, Calibration of IST Gages for SW system (AWO 96-1699)
* PMP 9.2-19, Calibration of IST Pressure Gagas, F1-1438A (AWO 96-1699)
* AWO 96-8735, SW Return Line 6-WS-151-250, Lower Level SFB
* AWO 97-1469, SW Supply Line 6-WS-151-126, All Levels SFB e AWO 97-1468, SW Return Line 6-WS-151-250, All Levels SFB
* AWO 97-806, SW Return Line 6-WS-151-250, Lower Level SFB e AWO 96-3991, Spent Fuel Cooling Pump P21 1 A Lubrication
* PMP 9.9 5, Spent Fuel Cooling Pump Coupling Inspection (Section 0.7)
* AWO 96-9836, North Service Water Header NDE Exams
* AWO 97-1451, SW-CV-963 Installation (DCR97-002)
* AWO 97-1455, SW-V-964 Installation (DCR97-002)
 
_ - _ _ _ _. . . . . _ _ .. _ _ ___. _ _  __ _ ._ _ _ ___ _ . __ __
'
i
-.
            ;
            !
.          i
;            !
4      24      l
-            )
  :o AWO 97-1518, Repair SW Return Line 6-SW-WS-151-250
:


Except as described below, thu inspector had no further comments in this are .
; Observations and Findinas
:            .
i  AWO 96-3991 The inspector observed licensee personnel complete preventive    i maintenance inspections and lubrication of the "A" SFP cooling pump per AWO 96-
,
3991 and procedure PMP 9.9 5. The work was completed in accordance with the    .
l  work controls and procedure. instructions. Maintenance personnel were    !
i  knowledgeable of the equipment.- No discrepancies were identified.-    -
AWO 97-806 On March 21,1997, the licensee performed ultrasonic inspection of a 3 inch segment of line 6 WS-151-250, which is part of the service water return
*
            :
i  header from the spent fuel heat exchangers located on the lower level of the SF ;
'
The inspection was performed as a part of the monitoring program to follow general corrosion noted in the line during an inspection in October,1996 (AWO 96-8735).    >
When replacing a pipe "T" in the return line, the licensee identified general corrosion  :
over a 3 inch section of pipe just down stream of the butt weld for the "T". The    l u  : degraded pipe in October.1996 had a minimum wall thickness of 0.100 inches, as F  -
compared to the nominal O.280 inch wall for the schedule 40 carbon steel pip ,
            ;
The October defects in were dispositioned as acceptable per NCR 96-267    '
-            1 i
The March 21 inspections noted continued corrosion with general wallloss over the    ;
{
  . . -same areas identified in October 1996, but with a 0.030 inch wall loss noted in one    2 4  -  localized area (grid section 10-11); the minimum wall thickness at that location was
;  O.067 inches. This defect was evaluated under nonconformance report (NCR) 97-
,
  -  003 and was found unacceptable in that a structural analysis showed that the pipe    !
could not withstand all design basis loads. The line was considered inoperable, as was the SW cooling lines for the SFP cooling system. Based on these results, the licensee began a program of expanded UT examinations of other locations in the    i SW system supply and return piping to establish the extent of the general corrosion
            *
,
,
            '
Other actions that have been taken to improve performance standards include holding
;
problem. The licensee also took actions to repair degraded service water piping in
 
accordance with the ASME Code. This matter is discussed further in Section E No discrepancies were identified in the troubleshooting efforts or operability l  evaluation l AWO 97-1468 and 14769 Ultrasonic (UT) examinations were completed on March l  27, March 31 and April 1 as a result of the defect identified on March 21 (AWO 97-    l 806) that rendered the SW system return header inoperable. The licensee followed    i the guidance of Generic Letter 90-05 to select a sample of 5 locations for UT    1'
examination that would be in piping similar to the original defect, and deemed susceptible to corrosion. The objective of the sample selection was also to e
determine the extent of the corrosion problem.
 
'
For the first sample set of 5 locations, the licensee performed ultrasonic inspection  l (UT) at two locations on line 6-WS 151-126 (AWO 97-1469), which is part of the
            ~
'
service water supply header to the spent fuel heat exchangers located on the lower    i
.
y ,wy ,,
y p w--,o-- ,
e-c-- r< ,- g a- 4 - v.,, ,,,m_..,a e ee... .. _,, , __.,__. . --s .-_ s m ,- --- .. -
 
.
.
 
and mid levels of the SFB The licensee also performed ultrasonic inspection at three locations line 6-WS-151-250 (AWO 97-1468), which is part of the service water return header from the spent fuel heat exchangers located on the lower, mid and upper levels of the SFB, Allinspections examined the general wall conditions around the pipe circumference over a two inch long segment using a one inch square inspection grid, inspection areas were expanded longitudinally as needed to characterize any defect The UT results showed evidence of general wallloss and corrosion, but to a much lesser extent than the initial defect. The pipe wall thickness was above 0.150 inches in all areas inspected except one (UT Location #5 - see the table below). l The pipe wall thicknesses remained generally above 0.200 inches, with a few locations showing pitting with remaining wall thickness below 0.200 inches. The degraded pipe sections were evaluated on April 1,1997 (reference Calculation N SDS-1556MY, Change #2); all defects were found acceptable in that a structural analysis per ASME Section XI Code Case N-480 showed that the pipe could withstand all design basis loads. A new Tmin for the pipe was established at 0.142 inches by this calculation, with a minimum allowable local wall thickness of 0.064 {
inches for the defect (Taloc). Although the defect did not create an inoperable )
g condition, the licensee took actions to repair degraded service water pipe segment-in accordance with the ASME Cod The licensee's initial response was to stop the piping examinations at this point, l based on the erroneous assumption that no further expansions were required based l
+n all defects showing wall thicknesses greater than the Taloc value of 0.064 !
inches. The inspector questioned this approach with licensee engineering, and j stated that if any defects were found with wall thickness below the Tmin value of 0.142 inches, then the guidance of GL 90-05 would require that additional ;
examinations be performed. After further consultation between the site and design !
engineering groups, the licensee expanded the UT examinations of the SW piping to an additional 5 locations. The second sample of UT examinations (UT Locations #6 through #10) were selected in both the supply and return headers at locations that were deemed susceptible to the corrosion mechanis The minimum wall thickness (Tmin) observed at each of the ten areas inspected was as follows:
UT Minimum Wall Measurements on Service Water Pipe Expansion Sample N UT # UT LOC #1 UT LOC #2 UT LOC #3 UT LOC #4 UT LOC #5 ID/elve Supply (30") Return (30)' Return (40') Supply (40') Return (50')
T min 0.178 in 0.152 in 0.154 in 0.182 in 0.116 in
 
    . - -  . .
i
.
 
  '
  '
Expansion Sample No. 2  l UT # UT LOC #6 UT LOC #7 UT LOC #8 UT LOC #9 UT LOC #10 PAB *
individuals accountable for unacceptable performance through appropriate counseling i and/or disciplinary actions. Additionally, a mission statement has been developed and l Issued reflecting the high standards necessary for the safe storage of spent nuclear i fuel and conducting the decommissioning process. The work observation program at the HNP will be upgraded by June 30,1997 to provide increased management and l
ID/elev Supply (50') Supply (63') Return (63') Return (63') Return (32')
supervisory observations of work activities and opportunities for coaching and positive
T min 0.196 inch 0.148 inch 0.150 inch .160 inch 0.100 in (0.040 in pit)
< reinforcement The improved Corrective Action Program has implemented a routine i trending report that provides management with timely data regarding performance
i Based on the data in the second expansion sample, the UT data at Location #10 !
!
showed an apparently deep pit with a general wall thinning down to 0.100 inches within one examination grid, and a very localized but very deep pit within that area showing a remaining wall of 0.040 inches. This defect was found in a section of the return piping located just above the east door of the auxiliary building and just ;
lssues.
upstream of the tie in to the main SW return header (ACR 97-179). The licensee
' initiated actions to complete calculations to support an operability determination for '
this portion of the SW header, and to obtain additional nondestructive examination data that would allow better characterization of the defect. The UT results at Location #10 showed substantial wall thickness remained over most of the circumference of the pipe, with wall thicknesses in the range of 0.200 - 0.256 inch in the upper (unwetted) hemisphere of the pipe, and wall thicknesses in the range of ;
" O.143 to 0.178 inches in the 60wer (wetted) hemisphere of the pipe. The general !
wall thickness within the 4 square inch UT exam locations around the 0.040 inch !
defect site was in the range of 0.174 to 0.198 inches. Licensee actions in this area .
were in progress at the conclusion of the inspectio ,
Licensee engineering used the above data to evaluate the conditions in the service ,
water system. Similar to the first set of data, the second expansion sample showed pipe wall thicknesses generally above 0.200 inches with extensive but randomly distributed pitting, and some pits that resulted in remaining wall thicknesses below 0.200 inches, as shown in the table. A general pattern emerged in which corrosion resulted in pitting that was generally worse in the return header that in the supply, which was attributed to the fact that the return header was not always full (the SW ,
system that was open to the atmosphere). The pattern shown by the UT data was also supported by visual exam on the two pipe spools that were cut out of the return header as part of the repair of the defects that were unacceptable. The inspector observed the cut out cross sections of piping and noted that substantial pipe wall thickness was evident over a significant portion of the spool, even in the areas of generalized pitting corrosion. The deep pits were highly localize This matter is discussed further in Section E1.2. No discrepancies were identified in the conduct of piping inspections and the actions to characterize degraded wall thicknes AWO 971518 The inspector reviewed this work activity on April 5 - 7, which involved the repair of a section of the SW return piping. The repair location was in Line 6-SW-WS-151250 just upstream of support WS-RH-6A on the 30 ft elevation of the spent fuel building. The repair involved replacement of a segment of piping that included the pipe wall defect that rendered the SW system inoperable (AWO
  .


    .. __ ~- -- .-.
In summary, CYAPCO takes these violations very seriously and is committed to implement and complete the broad scope corrective actions to improve station i performance. We will continue to keep the NRC Staff informed on our progress in these areas and are committed to demonstrating improving trends in areas of past
'
; weakness prior to proceeding with major decommissioning activities.
i r
' '


97-806: UT exams completed on 3/21/97), as well as the segment of pipe with general wallloss down to 0.152 inches (UT Loc #2). The repair replaced the pipe with the original defects plus at least 1 inch of pipe beyond the defect. UT exams t were used to assure the adjacent pipe had sufficient wall thickness that was not ;
The following are CYAPCO's commitments made within this letter. Other statements l within this letter are provided for information only.
affected by the generalized corrosion. The inspector observed the QA inspector complete a liquid penetrant examination (LPE) on April 7 of the root pass for the
:
. two butt welds per procedures NU-VE-2 and NU-LP-1. The quality of the welds was generally good, as indicated by the appearance and the results of the LPE. The !
!
workers were familiar with the welding and LPE processes and procedure Licensee activities were in progress at the end of the inspection period to complete and accept final welds. No inadequacies were identifie AWO 97-1519 The inspector reviewed this work activity on April 7-8, which involved the repair of a section of the SW return piping. The repair location was in ,
. CY-97-058-1 Publish site-wide and department level standards and expectations by l June 30,1997.
        '
Line 6-SW-WS 151-250 about 4 feet above the floor on the 47 ft elevation of the spent fuel building. The repair involved replacement of a segment of piping that included the pipe wall defect showing a minimum wall thickness of 0.116 inches-(AWO 97-1468: UT Loc #5). Additional supports were added to support the line in l the vertical direction while the affected spool piece was replaced. The inspector ,
I m . noted the licensee used additional UT inspections (UT Loc #5A) to assure the .I adjacent pipe had sufficient wall thickness that was not affected by the generalized l corrosion. Licensee activities were in progress at the end of the inspection period-to prepare the site to cut the affected line and install the new spool piece. No inadequacies were identifie l AWO 97-1451 The purpose of this work activity was to install check valve SW-  4 CV-963 in the service water supply header 6-WS-151-126 as part of Design  I Chango Request DCR 97-002. The inspector reviewed the activities in progress  j during this period to complete the installation, which was completed satisfactoril I The welding appeared to be of good quality, as evidenced by the appearance of the root and final welds, and by the satisfactory completion of non destructive !
examinations.
 
.
AWO 97-1455 The purpose of this work activity was to install test valve SW-CV-964 in the service water supply header 6-WS-151-126 as part of Design Change Request DCR 97-002. The inspector reviewed the activities in progress during this period to complete the installation, which was completed satisfactorily.


. Conclusions The maintenance activities observed this period were acceptable. The work was generally of good quality and was performed by knowledgeable personnel. The control of work was good, with evidence of good planning and coordination of the personnel involved in the work. Licensee actions were appropriate and conservative to complete SW piping inspections, characterize defects and to repair sections
,
,
'showi6g degraded wall thickness. The extensive corrosion and general degradation in the service water system resulted in an inoperable condition for the system relied
l CY-97-058-2 Upgrade the work observation program by June 30,1997.
  ~
        !
 
-, ,
 
    . __
.        .
        . i 28    l
;
'  upon to cool the spent fuel. This finding ~ appears as another example of poor plant material conditions that challenge systems important to plant safet ML2 Surveillance Observations-Using Inspection Procedure 71707,61726 and 62703, the inspectors conducted periodic j . reviews of plant status and ongoing surveillanc a. - Insoection Scope (61726)
l  The inspectors observed portions of the following surveillance activities:
e Emergency Diesel EG-2B Manual Starting and Loading (SUR 5.1-1578)
e Emergency Diesel EG-2A Manual Starting and Loading (SUR 5.1-157A)  )
J e Radiation Monitoring System Calibration (SUR 5.2-81.6, Rev 18)  a e All Modes Locked Valve Checklist (SUR 5.1-126, Rev 24)
e- SW Check Valve SW-CV-963 Bench Test (AWO 97-1448)
e SW Hydrostatic Pressure Test (ENG 1.7-65, AWO 97-1451)
e . Functional Testing of SW-CV-963, (ST .11.7-201)
w e SFP Heat Exchanger Temporary SW Supply Flow Test (ST 11.7-203)
t e- . Inservice Testing of SW Supply.to SFP Cooling Check Valve (SUR 5.7-217) .
  -e . inservice Testing of SW Pump Discharge Check Valve (SUR 5.7-89) Observations and Findinas SUR 5.1-17B The purpose of this test on January 22 was to demonstrate the 4  standby readiness of EG-28. The test was performed by starting the diesel manually from the excitation control panel. When the operator started the engine,
,  rather than run initially at 450 rpm speed as expected, the engine went to full speed at 900 rpm and automatically flashed the field. An engine start failure alarm  l annunciated. The operators shutdown the diesel pending further troubleshooting
        '
,
and repair. The deficiency was described in ACR 97-39. Licensee investigations on January 23 and 24 included troubleshooting the engine start circuitry, running the
;
diesel with test instrumentation to monitor the performance of the speed sensing i  relays in the starting circuit, and a satisfactory test of the diesel at rated load. The i
.
failure could not be repeated and no defective component was identifie The licensee concluded the deficiency was probably caused by one of two relays in
!-  the manuel start circuit, whose failure could not prevent the diesel from performing C
  .its design basis function. An engineering evaluation completed on January 27, 1997 concluded that EG-2B remained operable (reference CY-TS-97-033). The licensee continued to test EG-2B weekly until February 19 while instrumented to monitor the start circuits; however, the deficiency did not reoccur. The licensee resumed a normal test interval for EG-2B of once per month. The inspector
,
reviewed the troubleshooting plan and activities, reviewed the engine start circuits
!-  and logic (reference drawing 16103 31099 sheet 3), the engineering operability
*
  . evaluation, and the subsequent augmented testing on the engine. No discrepancies.


i CY-97-058-3 Complete the evaluatlon to replace the make-up valve to the
'
'
  ' were identified.
condensate receiver tank for the auxilisry boiler by July 31,1997.


-. _ , . . , - . ~ - _ ._ ,,,, _ -
CY-97-058-4 Complete implementation of WCM 3.3-1 by September 1,1997.
 
- . _ _ _ .
    - _ . .
.
 
SUR 5.1-17A The purpose of this test on February 12 was to demonstrate the operational readiness of EG 2A. The test was completed satisfactorily in that the diesel started, loaded and energized the associated 480 volt emergency bus as required. However, when the engine was shutdown, the licenses noted that the engine remained at 900 rpm instead of 450 rpm as expected for a 11 minute cooldown period during the shutdown sequence. The licensee evaluated the possible causes for the problem and concluded the safety function of the diesel was not affected. Licensee troubleshooting in March identified a faulty relay, and plans were made to replace the relay during the April test outage of the diesel. The problem recurred during a EG-2A monthly operability tests in March (ACR 97-127)
and April (ACR 97-178), but the diesel performed all safety functions correctly to start and carry associated loads. The planned relay repairs were not completed in April due to operational restraints in removing EG-2A from service with degradations in the service water system. The repairs were deferred until the May test perio The engine remained operable and in standby pending further relay repairs and investigation of the proble SUR 5.1-178 On February 5,1997, the inspector observed a special performance (extra test) of surveillance (SUR) 5.1-17B, Emergency Diesel Generator (EG)-2B su Manual Starting and Londing Test. EG-2A had twice experiencing.an abnormal . ;
- shutdown where the cooldown sequence (operation at half-speed for about 12 minutes) had not automatically occurred. This condition was documented in two ACRs and AWO 97-0873 was prepared for troubleshooting the problem with EG-2A during the next system outage (scheduled for May 12,1997).
 
The extra surveillance of EG-2B was to ensure its operability since EG-2A was not shutting down correctly. To allow for this special test, starting the EG but not loading the generator and tying to the emergency bus, Temporary Procedure Change (TPC) 97-17 to SUR 5.1-17B, authorizing EG startup to full speed without l loading, was PORC approved on January 30,1997. For this observed test, EDG-2B ,
automatically slowed down to half speed for unit cooldown, as designe l
 
Prior to the EDG-2B surveillance run, a chemical technician collected a sample of the fuel oil from the 550 gallon engine mounted day tank. The inspector noted that the technician had no paper work with him but was knowledgeable about collecting the '
i sample. The oil collection was in accordance with SUR 5.4-41, Diesel Oil Surveillance, and had been scheduled by Work Order CY-96-07919. Nothing in CY-96-07919 or SUR 5.4-41 required data entry (tne procedure had no sign-off blanks).
 
This was considered acceptable because it was within the skill of the technician AWO 97-1448 The purpose of this test on March 26 was to conduct a pre-installation leakage test of the 6 inch swing check valve that would be installed in the service water supply line (6-WS-151-126) to the SFP heat exchangers per DCR 97-02. The Anchor Darling swing check valve (serial # E6318-54-5) was an ASME lil, Class 2 component rated for 100 degrees F at 275 pounds. The purpose of the plant modification was to assure that the SW supply lines remained filled with water following a postulated loss of power. The engineering evaluation for DCR 97-02 assumed that this function would be assured so long as back leakage past the


,
,
CY-97-058-5 Convert applicable procedures to surveillance procedures by
;  September 1,1997, i
! If you should have any questions, please contact Mr. G. P. van Noordennen at j (860) 267-3938.


check valve seat was limited to 2 gallons per minute for the approximate 48 second interval between the loss of power end the start and load of the emergency diesel generator and associated SW pump during an LNP. The licensee established a leakage limit of the check valve of 2 gpm and 0.5 gpm, for insitu and bench testing, respectivel The inspector observed the test setup and results of the bench testing completed on March 26. The check valve was pressurized in the reverse flow direction at a pressure of 8 psig, as leakage was measured with a graduated cylinder. The inspector verified that the test pressure corresponded to the maximum pressure for the elevation head of water from the high point in the service water system to the elevation the location of the check valve. The licensee measured 20 ml/ min during the test, which was less than 0.01 gpm and much less than the 0.5 gpm acceptance limit. Test personnel were familiar with the test equipment and AWO controls. Engineering personnel provided good oversight of the test activitie ENG 1.7-65 The purpose of this test on April 4 was to verify the integrity of the safety class 3 service water supply header following the installation of check valve SW-CV-963. The hydrostatic pressure test was conducted in accordance with
Very truly yours,
- ASME Code Section XI,1983 Edition. The 110 psig class system was tested to-121 psig (110%) with a 10 minute hold period, followed by an inspection for leakage at the new welded and mechanical joints. The results were satisfactor No inadequacies were identifie ST 11.7-201 This functional test of SW-CV-963 was performed on April 4 to assure adequate flow to pravent flutter. The test was completed successfully to demonstrate that SW flow to the heat exchangers was not degraded by the additional restrictions introduced by the new check valve in the SFP supply lin The measured flow was satisfactory at 920 gpm, which was above the minimum valve assumed in the design basis of 855 gpm. Further, the test identified the minimum flow the licensee needed to maintain through the supply line to avoid valve flutter under low flow conditions. No inadequacies were identifie SUR 5.7-217 The purpose of this inservice test of SW-CV-963 was to assure it opened to allow adequate cooling of the SFP and, to verify that it closed and was leak tight to assure the back leakage met the safety function assumed in DCR 97-002. The acceptance criteria was that measured back leakage was less than 7570 cc/ min, met the assumptions in the safety evaluation for DCR 97-002. The licensee's engineering evaluations (CREARE letter dated March 17,1997) showed that if back leakage from the SW supply header was limited to 2 gpm, the postulated waterhammer event on during a LNP condition would be prevente The test was completed satisfactorily to show that the check valve opened and provided adequate flow to the SFP heat exchangers. The licensee also measured zero back leakage from the check valve during the performance of the test on April 4 (test method for SUR 5.7-217 Revision 0). However, test personnel noted the need to improve the test method to isolate the test boundary, and that a partial vacuum was drawn in the piping upstream of the check valve pr;or to the
 
measurement of back leakage. The inspector reviewed the test method and results, and determined that the SW piping was partially drained during the test, and that the test method had failed to adequately measure actual back leakage. The inspector's concerns were discussed with the operator in charge of the test on April 4. The licensee prepared ACR 97-173 to describe the deficiencies in the test method. Licensee actions were in progress at the conclusion of the inspection period to revise the test method and measure back leakage from SW-CV-96 ST 11.7-203 The purpose of this test was to measure flow to the SFP heat exchangers with the temporary hoses installed per NOP 2.24-3 for various supply and return configurations. This test was successfully completed to provide the data for an engineering evaluation of the adequacy of service water cooling flow to the SFP heat exchangers when using the fire hoses, Conclusions The surveillance activities observed this period were acceptable. The preparations for and conduct of testing was generally good. Pre-job briefs were very good and provided for good planning and coordination of the personnel involved in the test, a: and good control of test activities. The surveillances were performed by-knowledgeable personnel. Although evaluations to address degraded condition on the emergency diesels were acceptable, two diesel problems remained not fully resolved and licensee actions to address deficiencies on the emergency diesel i generators were not as aggressive as in the past. The tests conducted were acceptable to demonstrate the operability and readiness of plant equipment. An exception to good performance in testing was the faulty test method used on ST j 11.7-201 to verify acceptable back leakages from service water valve SW-CV-96 )
l M2 Maintenance and Material Condition of Facilities and Equipment  i M 2.1 Material Condition Deficiencies (VIO 97-01-02.d. URI 97-01-04)
1 Inspection Scope    j l
The purpose of this inspection was to assess the status of the plant material I conditions, and the performance of licensee processes to address discrepant conditions, Qbservation and Findinas in the work controls area, the licensee documented a finding (ACR 97-90) on February 18,1997, that trouble report tags were not removed as required following maintenance activities. Based on a review of 275 trouble report tags hanging throughout the plant, operators found that 113 (41%) were associated with trouble reports and work orders that had been completed. Trouble reports (TRs) that remained after the work was complete made it hard for operators to perform their job properly to identify discrepant plant conditions, since some items that are not ,
operating properly are not being reported under the trouble report system because it l l
l l
 
  .
  .
a
_
 
m-
is assumed that there is an active trouble report and repair plan in existence. The licensee determined that this problem occurred because workers were either not familiar with or were not following the requirements of procedures WCM 2.1-1, Work Control Process, and WCM 2.12, Trouble Report / Job Scoping. TRs that are deleted because the work is performed under minor maintenance WOs are difficult to track and the workers may not be aware of the existence of a TR ta ;
The ACR was discussed at the February 20,1997 meeting of the management review team (MRT) and assignments were made to correct this problem. Licensee actions to address this matter were in progress at the end of the inspection perio However, Work Control Manual WCM 2.1-2, Trouble Reporting / Job Scoping, Revision 3, requires in Steps 1.2.3 and 1.3.3 that trouble report tags be removed and closed epon development of work packages, or for work that will be completed under blanket authorized work orders. The failure to remove TRs was a failure to implement WCM 2.1-2, and was an example (the fourth of six) of a violation of Technical Specification 6.8.1 (VIO 97-01-02.d).
 
Other discrepancies in plant material conditions were noted during the period, and included the corrosion degradation of the service water piping (described in Section e M1.1 above), the generally poor conditions around the service water Adams filter area in the upper level of the auxiliary building, and the poor lighting conditions in the containment caused by a number of lights that have burned out. Other material discrepancies included the recurring problems with the radiation monitoring system, including an unresolved design defect in the SCANRAD compute The licensee tracks key performance indicators (KPis), including the backlogs in l total trouble reports and work orders, and the trends in work deferrals and the ;
amount of rework. The KPis for the above parameters show a large amount of outstanding work, with little progress over the first quarter of 1997 to improve performance. The lack of progress was attributable to several reasons, including a decision to not conduct work on systems that were no loager operable or required i for decommissioning; a decision to defer outage related testing on systems that I were no longer operable or required for decommissioning; the management decision )
      '
to restrain the conduct of work in the radiologically controlled area, pending the development and implementation of plans to improve radiological controls; and, the reduction in available workers as the staff was reduced to the decommissioning organization. The adequacy of staffing in the maintenance area to address plant material conditions and to preserve the systems that remain important to the safe storage of fuel remains an area that requires further evaluation by the NR This item is unresolved pending further NRC review of licensee actions to address the deficiencies in ACR 97-90, and licensee performance to address the backlog of
 
station work (URI 97-01-04).
 
_ . .
.
.
33 Conclusions Past NRC inspections have identified concerns with discrepancies in plant material conditions and the performance of licensee programs to preserve systems important to safety (reference Inspections 96-80, 96-10, 96-11 and 96-12). The above findings appear as recurrent problems in this area.


_ . _ - - - ._-  . _ . . . . . . _ . . . _ _. . . _ . _ . _ _ . - - _ . _ . _ _ _ _ . _ . . _ _ . . . ~ -
- ' -
.,        .
. U. S. Nuclear Regulatory Commission  -
_,
CY-97-058/Pago 3
.
.
M3 Maintenance Procedures and Documentation M3,1 TS Surveillances Covered by Procedures (VIO 97-01 -02.e)
:    CONNECTICUT YANKEE ATOMIC POWER COMPANY i
      ,
i fMAW T. C. Feigenba6m
      ' Insoection Scooe (61726)
:    Executive Vice President and
This inspection was performed to review the licensee actions to periodically review station procedures. During inspector review of a change to Administrative Control l Procedure (ACP) 1.2 6.5A, Station Procedures, a concern was raise Observations and Findmag    ,
!
l m ACP 1.2-6.5A, Station Procedures, Revision 0 dated November 13,1996, states in Section 1.3.6 that procedures in the Surveillance Procedure category are reserved !
Chief Nuclear Officer i
for surveillance inspections or tests which are required by the Technical Specifications (TS). ACP 1.2-6.5A also requ;res, as part of the biennial review, that plant personnei verify all TS Action Statements are clearly identified and applicable -
cc: H. J. Miller, NRC Region i Administrator M. B. Fairtile, NRC Project Manager, Haddam Neck Plant W. J. Raymond, NRC Senior Resident inspector, Haddam Neck Plant
l to the Limited Condition for Operation. The licensee implemented this requirement I through a revision to Work Control Manual (WCM) 3.3-1, Technical Specification Surveillance (TSS) Tracking, Revision 2, issued July 18,1995. WCM 3.3-1 4 established a master database for all TSS with an administrator, required an annual audit of the master list database to ensure each TSS was on the database and tied to a specific surveillance procedure, specified a January action request for review of all TSS procedures, and provided critical function review question ,
,
l The inspector requested the licensee to provide the resu!ts of the annual audit of surveillance procedures required to be done per ACP 1.2-6.5 and WCM 3.3-1. In response to the inspector's February request regarding reviews to assure all TS realized requirements are covered by surveillance procedures, the licensee reported !
that a review of the master database had not been completed per WCM 3.3-1 and initiated the required review. Although the assignment to complete the review had been made (via February Action Requests), the reviews were stillin progress as of April 4,1997. Further, in response to the inspector's request, the licensee provided a OAS audit showing numerous problems in implementing the operational -
surveillance program. Two of the problems identified were:il one surveillance (SUR 5.2-69) where a second TSS requirement was not reference and the first TSS was I no longer required; and, ii) nor surveillance procedures PMP 9.1-31 and ESP 14.1-4 j were used to meet TSS requnments. The deficiencies identified by the NRC and l the QAS audit were conditions contrary to procedures ACP 1.2-6.5A and WCM 3.3-
,
,
1. The failures to implement the surveillance program per these administrative
      !
I
      <


.
4
        .
      -


procedures were the fifth of six examples of a violation of TS 6.8.1 (VIO 97-01-02.e). Conclusions The deficiencies described above showed poor performance by plant personnel in following station procedures and in implementing the operational surveillance progra M4 Maintenance Staff Knowledge and Performance M4.1 Failure to Comolete Surveillances (VIO 97-01-05. VIO 97-01-06) Inspection Scope (61726)
  - - -
The scope of this inspection was to review the licensee performance for completing surveillance tests in accordance with the technical specification requirement Observations and Findinas Several events occurred during this period in which the licensee's staff failed to complete required surveillances as required. Most incidents were identified by licensee personnel and were entered in the ACR program for management review and followup. The issues noted during this period are summarized below:
,   .
ACR 97-57 wd, issued due to the discovery on February 4,1997 of the failure to complete a fire system surveillance (18 month fire system inspection) as required by the technical requirements manual TRM 16.1-3. The surveillance was due to be performed in September 1996. The test was not done due to a communications error among operations personnel during the scheduling of surveillance tests. The error occurred because another test completed per TRM 16.1-13 on 9/30/96 was >
  ,
mistakenly listed as the completion of TRM 16.1-3. The fire system inspection was subsequently completed satisfactorily, t
  .
ACR 97-66 was issued due to the discovery on February 6,1997 that the reactor coolant chemistry had not been verified as required by the technical specification TS 4.7.7 requires that the RCS be sampled every 72 hours to assure that chloride, fluoride and oxygen levels are below certain limits to minimize the potential for stress corrosion cracking of reactor vessel and RCS components. The last RCS taken and analyzed was on November 15,1996 when the RHR system was shutdown following completion the core offload into the spent fuel pool. The licensee stopped taking the samples because chemistry personnel believed the TS requirements were no applicable in the defueled condition. TS 3/4.4.7 states that the chemistry limits and surveillance requirements are applicable at all time Immediate corrective actions were to recommence RCS sampling in accordance with the TS requirements. The RCS water was well within the requirements for chloride and fluoride contamination (by an order of magnitude). The licensee had
 
35 continued to sample reactor cavity water to assure chloride and fluoride levels were satisf actory, but these samples were obtained from the cavity purification system and could not be assured to be representative of the water in the reactor vesse ,
Thus, there is some uncertainty as to what the chloride and fluoride levels inside the )
reactor vessel were during the period from November 1996 until February 199 l Licensee investigations for ACR 97-66 determined that RCS sampling was also stopped in the past when the reactor was defueled for extended periods (e.g.,
during the work on the core shroud in the late 1980's).
 
l This event was reported as licensee event report (LER 97-02). The safety  I significance of this surveillance and technical specification violation was low relative to future Haddam Neck operations since the chloride stress corrosion cracking i mechanism is a concern for subsequent operation of the RCS at normal operating l temperatures and pressures. Due to the decision to permanently cease plant operations, the reactor fuel can no longer be loaded into the reactor, and the RCS will no longer be operated at the normal operating conditions of 550 degrees F and i 2000 psig. However, the failure to sample and analyze RCS Chemistry from )
Nove " >r 15,1996 to February 6,1997 was a violation of TS 4.7.7 (VIO 97-01- 1 05).
 
ACR 97-77 was issued on February 11,1997 concerning to failure to complete two surveillances as required by the technical specifications on the B main station l battery, BT-18. The following tests were scheduled to be performed in October l 1996 to meet the requirements of the technical specifications: SUR 5.7-37, Station Battery Cellinspection, Cell Resistance and Rack inspection, which is required to be performed once per 18 mc..ths per TS 4.8.2.2. and 4.8.2.1.c; and, SUR 5.7-38, BT 1B Battery Service Test, which is required to be performed once per 18 months per TS 4.8.2.2 and 4.8.2.1.d. The tests were deferred to December 1996, which was acceptable since the surveillance would have been within the 125% windo However, the scheduled conduct of the tests was deferred due to the management !
work stoppage placed in effect in November 1996. Plant personnel failed to l adequately track completion of the tests against the required surveillance due dat The test were preformed satisfactorily un February 18-21,1997. Although TS 3.8.2.2 required only one DC distribution system be operable for Mode 5 & 6 operation (none is required with the reactor defueled), the licensee policy was to maintain both DC distribution trains operable and tested per the TS ACR 97-81 was issued for the discovery on 2/12/97 that the service water pumps were not tested as required per TS 4.7.3.b.2. The TS require that the pumps be l demonstrated operable at least once per 18 months by starting on a condition involving a loss of normal power. The TS surveillance was due on 9/26/96; the 25% extension interval expired on 2/10/97. This portion of the SW pump starting logic is normally tested each refueling outage during the conduct of SUR 5.1-18 and 19, the Test of the Train A and B Safety injection Actuation System with Loss of Normal Power. This test was not performed during RFO#19 because of the decision to permanently shutdown the f acility. The licensee initiated actions to
, write a new procedure to test the LNP feature in the SW pump start circuiL l
i l
l
 
__
.
 
ACR 97-142 was issued following the discovery on March 21,1997 that the surveillance of locked valves had not been completed per the schedule required by the technical specifications. TS 4.5.7.C requires that all accessible valves be verified to be locked in the correct position at least once per 18 months. The tests in performed per SUR 5.1-126 which was started and in progress during this inspection period with an assumed required completion date of March 1997. This date was based on the assumption that the surveillance had last been performed in September,1995. On March 21, the duty shift manager noted that SUR 5.1-126 had last been completed in March 1995, and was therefore due to be completed by September 1996. The scheduling discrepancy occurred due to an operator error in recording the date when SUR 5.1-126 was last done, which was taken from the shift logs and was assumed to be in September 1995. The licensee determined that September 1995 was the date when the shift logs were updated when revision 23 of SUR 5.1-126 was replaced with revision 24. The licensee continued the verification of the locked valve checklist, which was stillin progress at the end of the inspection perio ACR 97-126 was issued on March 12,1997 as a r=uit <>f an audit by the QAS group that concluded that the station tracking systems for ensuring the completion j of TS surveillances while in a defueled condition were susceptible to missing a surveillance. The OAS audit found that surveillances were being scheduled to start after the required frequency (but within the 25% grace period); surveillances were l not always scheduled within the production maintenance management system l (PMMS); surveillance requirements are not always tracked by plant departments; the surveillance procedures for calibrating effluent monitors do not reference the i technical specifications, which creates the potential for a missed surveillance; some TS surveillances are performed by non surveillance procedures (i.e., procedures other than SURs); and; surveillances are being completed without the use of automated work orders (which assures tracking and scheduling within the PMMS). !
Licensee action in response to the QAS findings were in progress at the conclusion of the inspection perio The inspector noted that although the above discrepancies were identified as a result of initiatives by either the line organization or the oversight groups, there was a demonstrated weakness in completing surveillance in accordance with the technical specification requirements. Further, this has been a weakness in the past at Haddam Neck, and a concern previously described by the NRC (reference Inspection item 94-27-01 and LERs 96-22, 96-17, 96-04 and 95-12). Tne f ailures to complete the surveillances as described in the above ACRs were a violation of the associated technical specifications. The failure of past licensee actions to correct this condition adverse to quality was a violation 10 CFR 50, Appendix B, Criterion XVI (VIO 97-01-06).
 
c. . Conclusions Poor performance was noted in the repetitive failure by licensec personnel to adequately implement the technical specification operational surveillance progra .
.
.
Docket Nos. 50-213 CY-97-058 Attachment 1
      '
Haddam Neck Plant Reply to Notice of Violations NRC Intearated Inspection Report No. 50-213/97-01
      .
  .
June 1997 m o


M8 Area Summary and Status of Regulatory Findings M8.1 (Closed) IFl 96-01-01, Cable Vault Materials Condition During inspection 50-213/96-01, the licensee determined that the material condition of housekeeping in the cable vault had degraded because the building ventilation system and the two sump pumps had become inoperable. This resulted in excessive moistwe in the room along with a buildup of accumulated water on the lower level of the vault. The inspector toured the cable vault on several occasions during the period and verified that the moisture and water did not cause an immediate impact on plant equipment. This matter was left open pending the completion of licensee actions to address the poor housekeeping and material conditions in the cable vault, and subsequent review by the NRC. The inspector visited this area to confirm corrective actions. The floor has been recoated, new grating and non-slip pads were being installed, and the sump pumps were operable although a small amount of water was still on the floor. This inspection follow-up item is close M8.2 (Closed) DEV 96-04-02. Heavy Load Proaram Commitments During inspection 50-213/96-04, deviations of CYAPCO's commitments of July 20, 1981 and April 16,1982, to satisfy NUREG 0612 (Phase 1) expectations on crane operator qualifications, were reviewed. In addition, a licensee letter dated June 29, j 1984, stated that nondestructive testing will be done on the identified lift devices '
. on a ten-year cycle. One identified lift device was the PAB floor block lift devic Plant procedure PMP 9.5-131, "Special Lifting Device Inspection and/or Load Testing," did not have this lift device included in the procedure. The NRC's safety evaluation on the " Control of Heavy Loads (Phase 1)," dated August 20,1984 noted that Phase I was acceptable at Haddam Nec The licensee response to these deviations, dated August 12,1996, included j commitments to revise procedure ACM 2.2-9, Control of Crane Operations, and On- 1 The-Job Training Guide to include and verify physical qualifications, in additicn, WCM 2.2-7, PAB/ Pipe Trench Floor Block Lifting Procedure, was revised to designate the use of the lift rig when lifting floor blocks. The inspector confirmed these changes were made. This deviation is close M8.3 (Closed) IFl 96-08-04, Auxiliary Feed Water Overspeed Trio During a quarterly test of the "B" auxiliary feedwater pump mechanical overspeed trip device, the trip valve did not fully close. The proposed corrective actions include developing a preventive maintenance procedure for the mechanical overspeed trip device, revising the acceptance criteria to define acceptable mechanical overspeed trip function, and replacing the trip linkages during the upcoming refueling outage. The issues were lef t for an inspector follow-up in IR 50-213/96-08. Due to the decision to permanently cease operation of the Haddam Neck facility pursuant to 10 CFR 50.82(a)(1)(l), the auxiliary feedwater system is no longer of safety importance. This issue is considered close .
.
M8.4 (Closed) IFl 96-08-05. Steam Generator Hold Down BQ1ts During a containment inspection, the licensee discovered that one of the #2 steam generator hold down bolts was broken. The 3 inch diameter by about 32 inch long bolt was found lying on the floor near its sliding / support block located ori the lower skirt assembly. It was presumed that the bolt " popped" out of its hole in the support block. This matter was left for inspection followup in IR 50-213/96-0 Due to the decision to permanently cease operation of the Haddam Neck facility, the steam generators are no longer of safety importance. This issue is considered close MQJ (Closed) IFl 96-08-06. Observations of Procedural Quality During inspection 50-213/96-08, the NRC identified that no periodic or preventive maintenance program exists for (Westinghouse] safety-related with lockout (WL)
relays. Inspector review of the maintenance history of the WL relays identified one corrective maintenance activity in 1990 to replace switch contacts on the "B" high containment pressure relay. It was also noted that the WL relays are tested and provided successful results in the past three refueling outages during the
+ performance of SURs 5.1-18 and 5.1-19. This issue was discussed with the licensee during the course of the inspection. The licensee initiated an assignment to consider the development of a periodic maintenance program for these safety related relay The inspector was provided a licensee engineering and electrical maintenance i analysis of the classified of these relays, within the scope of 10 CFR 50.65, dated l March 20,1997. This analysis concluded that, based on the limited wear the relays see, the limited benefits of an "out of the board" PM, and the fact that there is a 4
      '
large potential for human error involved in removal and re-installation of the relays, the outage-based surveillance procedure tests are felt to be the best and most informative preventive measure available to ensure fulfillment of the Maintenance Rule, it was also pointed out that due to the permanent shutdown of Haddam Neck that virtually all safety related WL relays, outside of electrical distribution, are now no longer required. The inspector confirmed that there were no Westinghouse safety related WL relays in the electrical distribution system. This inspection followup issue is close M8.6 (Onen) URI 96-08-15. Start-un issues (7/24/95 NRC Letter)
In a July 24,1996 letter from the NRC to the licensee, seven technical issues, some involving possible amendment requests were identified that need to be resolved for a plant restart. The major modifications and technical issues to be addressed during the outage include:
a) CAR f ans alternate service water modifications and license amendment; this item includes the performance of an integrated safety assessment to address the known related service water system issues (operating pressure, CAR fan performance and containment integrity).
.
  .
  .
39    l l
~ . -. _ -. .- -- .. .- . -. .- - -_ -
b) modifications to mitigate a main steam line break event - the main feedwater !
-
pump trip and discharge valve closure on high containment pressure, and j license amendmen I c) the conduct of the station battery test using the profile with the AFW ;
., .
hydraulic pumps running,    j I
  .
d) the completion of a new station battery calculation to demonstrate battery j operability with the AFW pumps shutdow !
l e) actions to address other discrepancies in material conditions, such as steam I I
generator hold down bolts and main steam bridge structural stee f) modifications to address RHR NPSH - eliminate the reliance on containment back pressure, and to upgrade the containment sum ,
g) Type "B" leak rate of containment penetration P-5 j
.       .
Due to the decision to permanently cease operation of the Haddam Neck facility, a issues a) CAR fans performance, b) main steam line break, part of e) dealing with steam generator hold down bolts, f) RHR containment back pressure NPSH, and g)
containment penetration P-50 leak rate test, are no longer important to the ;
permanently shutdown, defueled, and later decommissioned status. These issues ;
are considered closed. Likewise, issues, c) conduct of station battery profile test, !
d) new station battery calculations to demonstrate battery operability, and the part !
- of e) dealing with main steam bridge structural steel, remain to be corrected by the '
licenseo and inspected by the NR i M8.7 (Closed) URI 94-27-04. Surveillance Frecuency Exceeded  )
l This item was created in response to a failure to adequately complete surveillance requirements. The initial corrective actions were reviewed and NRC followup review (Inspection 95-14) left the matter open pending further review of the i
      '
conduct of surveillances for adverse trends. Licensee performance in the conduct of operational surveillances has declined, as summarized above in Inspection items 97-01-04 and 97-01-05. Licensee responses to NRC concerns in this area will be tracked under Inspection item VIO 97-01-05 & 06. Inspection item 94-27-01 is close l M8.8 Conclusions for Maintenance The maintenance and surva%nce activities completed this period were generally acceptable to assure important plant systems remained operable, to support operability evaluations and design change work, and to address emergent issues that challenged adequate cooling of the spent fuel. Exceptions to good performance included a weakness in the process to examine SW pipe for corrosion, and a faulty test method used to measure back leakage from a check valve. Additional discrepancies were noted in plant material conditions, in the implementation of the
 
_ _  _ ._ _ _ _ __
*
*
      ,
U. S. Nuclear Rcgulatory Commission CY-97-058/ Attachment 1/Page 1
.
*
 
        .
program to identify and correct material discrepancies, and in the lack of progress in
h
  ~
, Restatement of Violation 1 i
addressing deficient condition Poor perforrnance was noted in the failure by station personnel to follow procedures to implement the operational surveillance program, and in the repetitive failure to adequately implement technical specification surveillances in a timely manner. The extensive corrosion and general degradation in the service water system resulted in *
; Technical Specifications 6.8.1 requires that written procedures and/or administrative  ;
an inoperable condition for the system relied upon to cool the spent fuel. This finding appears as another example of poor plant material cond;tions that challenge systems important to plant safety. The recurrence of plant material deficiencies and problems in the area of technical specification surveillance testing revealed ongoing weaknesses the corrective action proces Ill. ENGINEERING  !
policies be established, implemented and maintained covering the activities as  ;
      !
recommended in Appendix A of Regulatory Guide 1.33. Regulatory Guide 1.33 3 requires that procedures be established governing plant operations and administration, f work controls, and security.
E1 Conduct of Engineering E1.1 Service Water System Modification - Water Hammer (URI 97-01-07) Inspection Scoce (37551)
The purpose of this inspection was to review the licensee evaluations of the potential for two phase flow in the service water system, and to complete modifications to preclude postulated waterhammer events, Observations and Findinos The inspector reviewed the licensee's actions in response to a concern raised by a previous NRC inspection in which the potential for two phase flow and water hammer loads in the service water system were identified. The NRC concern was ;
identified during the Engineering and Licensing inspection conducted in 1996, and -
was identified as inspection item 96-201-2 Following the inspector's request this period for information of the status of the engineering review of this issue, the licensee found that the potential for two phase i flow in the service water system had been identified, but not completely resolve Specifically, following the inspections in 96-201, the licensee contracted the services of a vendor to analyze the potential for two phase flow in the SW syste The initial engineering evaluations were reported to the licensee in vendor report TM 1788 dated July 9,1996, which concluded that there was no significant potential for column separation in the SWS following a loss of normal power (LNP)
event. However, following additional analyses, the vendor submitted a revised technical eva!uation in report TM-1788a dated August 14,1997, in which there were two locations in the SWS where column separation could occur. Specifically, the locations were in the higher elevations of the SW supply and return lines to the spent fuel pool cooling system (SFPCS) heat exchangers (63 foot elevations on the 6-WS-151-126 supply piping, and 63 foot elevation of the 6-WS-151-250 return piping).


.
.
.
,
 
General Response i in addition to the station-wide efforts to raise performance standards and actions to
The vendor's analysis for two phase flow conditions was postulated to occur as a result of the draining of water from the high points in the SW headers to the I SFPCS, as would occur when the SW pumps shut down following a loss of electrical power. As the water drained from the high points in the system, a partial vacuum condition inside the piping would result in separation of the water column in l to a vapor and liquid phase (column separation). The separation occurs at the high l point elevations because the water pressure at the top of the column falls below the vapor pressure of the liquid while the bottom of the column is open to atmospher On a LNP, the SW pumps would trip and the piping system would depressurize l within about 2 seconds. On restoration of power following the start of the emergency diesel generators, the SW pumps would be started in about 45 seconds from the start of the LNP event. The sudden repressurization of the SW lines and the collapse of the water column would result in waterhammer in the piping syste l
      '
The vendor analysis postulated that the associated water hammer loads could be excessive, and result in as much as a three inch displacement of the piping system when the collapsed water column reached sharp bends in the piping syste j On receipt of the vendor's report in August 1996, the licensee completed a preliminary assessment of the SFPCS supply lines, which concluded that the SW a- cooling lines had not been affected adversely affected by the design deficiency...
This conclusion was based on the observation that there had been LNP events in the past 28 years of the plant (either as a result of planned testing, or inadvertent power outages), and that a postulated three inch displacement of the piping system !
would have been either observed or left evidence in the piping system. A licensee design engineer walked down the SW headers and identified .no evidence in the piping or attached insulation indicative of a waterhammer event in the line Although the resolution of the analysis results remained unresolved in August, 1996, the licensee classified the matter as an issue that had to be resolved prior to plant restart, and assigned the project to a plant engineer to develop a modification that would resolve the issue (modifications involving vacuum breakers and check valves were under consideration). However, the project was not completed for various reasons, and no action was taken to complete the modifications or to otherwise resolve the analytical problem through the end of 1997, in response to the NRC inquiries this period, licensee engineering reviewed the status of the vendor's analyses. The vendor evaluations in August assumed that the SW return side isolation valve SW-AOV-9 would fail open on a LNP, since that was a conservative assumption in the analysis of the SW system for plant operations at full power. In practice, SW-AOV-9 would fait closed on a LNP. The vendor evaluated the potential for column sepatation assuming AOV-9 operates as designed, and concluded that the SW return lines would remain operable under these conditions. However, the supply side piping could continue to drain into other headers (e.g., the diesel generator supply piping) in the 45 seconds the SW pumps were not operating, and result in column separation. Engineering evaluations in March 1997 of the effects of the postulated water hammer concluded that the pipe stress levels would be unacceptable for the postulated transient. This conclusion wes based on the postulated piping stresses and support loads at the locations of the highest waterhammer loads, and included the conservative assumptix that the
 
.
.
 
service water pipe wall thickness was at the lowest value seen as a result of erosion and corrosion inspection. The assumed minimum wall thickness was that identified and repaired at a "T" fitting in the SW return line from the "A" and "B" SFP heat exchangers in the Fall of 1996. Under these conditions, the licensee concluded that the SW pipe pressure boundary would fail under a design basis LNP even Based on the above, the licensee declared the spent fuel cooling system inoperable on March 11,1997, and made a report to the NRC per 10 CFR 50.72(b)(2)(l)
regarding plant operation in a degraded and unanalyzed condition. The licensee initiated actions this period to address the design deficiency. Possible fixes considered included additional analyses of the potential for two phase flow, and design changes to install either vacuum breakers of a check valve in the affected piping. The option selected was the development of design change request (DCR 97-002) to install a check valve (SW-CV-963) and associated test valve (SW-CV-964) in the SW supply line to the SFP heat exchangers. The licensee installed a 6 inch Anchor Darling swing check valve in the SW supply header located in the southeast corner of the auxiliary building near the Adams filters. The inspector verified that the check valve rated for 150 psi service conditions and was procured 2.- as a safety related component. The licensee also contacted the original vendor >
whose engineering evaluations assisted in the identification of the design deficiency to provide assistance in developing the check valve performance criteria needed to eliminate the potential for excessive water hammer loads. Licensee actions were in progress at the end of this inspection period to install and test the check valve, and to complete and approve the safety evaluation for DCR 97-00 Several discrepancies in licensee performance were noted in development and
< resolution of this item. Once the licensee's contractor had identified the potential-two phase flow as a design discrepancy, the licensee inappropriately classified the issue as a plant " restart" item in August 1996. The design discrepancy was a condition adverse to quality that was relevant to plant operations in the shutdown condition, which should have been resolved in the Fall of 1996 prior to the offload of the core into the spent fuel pool. The design engineering group used reasonable engineering judgements based on a walk down of the SW lines in August 1996 to assess the actual potential for excessive water hammer loads, but left final resolution of the matter inconclusive. Further, the design engineering group showed poor performance to track the issue internally to assure the timely completion of further analyses and the development of modifications to address the issue. This was an example of weaknesses in the licensee's corrective action process to assure that conditions adverse to quality are promptly addressed to preclude recurrenc Finally, licensee actions to track and resolve NRC inspection item 96-201-24 were poor. The design vulnerabilny in the SFP support systems would have remained had it not been for the NRC followup of this matte _ ___
_ ._ . _ _ . .
, .-       !
l
!
l 43  '
      ' Conclusions These issues are similar to deficiencies in licensee performance highlighted by past NRC inspections and addressed at the escalated enforcement conference held on December 5,1996. As followup to this issue under ACR 97-119, the licensee planned to perform a root cause investigation to determine how this issue was tracked by engineering, and to verify that other issues were property resolved. The licensee also planned to assess this discrepancy and report his evaluation per 10 CFR 50.73. This item is unresolved pending the completion of licensee actions to address the SW two phase design deficiency, and subsequent review by the NRC (URI 97-01-07).
 
EL2 Service Water System Evaluations - Corrosion (URI 97-01-08)  >
On March 26, the licensee declared the SW return piping from the SFP heat ,
      '
exchangers inoperable at 7:15 p.m. as a result of corrosion induced excessive wall thinning in a 6 inch carbon steel pipe 6-WS-151-250 (ACR 97-154). The defect was located in a three inch area of heat exchanger return piping about one foot downstream of the common heat exchanger return tee ('T'). See Section M ., (authorized work order 97-806) for further details on this issu ;
The corrosion defect was identified during a planned ultrasonic (UT) inspection o the line in an area of known wall thinning. The pipe wall thinning was previously identified during inspections on 1996 for microbiologically influenced corrosion and was found acceptable at that time (reference Calculation 96-SDS-1556MY, Revision
- O dated 10/27/96). Using the data available in October 1996, the licensee assumed that the pipe would remain operable for up to 1 year based on the predicted 7 corrosion rates, and recommended that re-inspection be performed in six months.-
The pipe was determined to be inoperable in March 1997 as a result of additional wall thinning at the site of the original defect, based on an engineering evaluation by the Structural and Design support Group that concluded that the pipe could not withstand design basis loads - a combination of deadweight, pressure and seismic (reference Calculation 96-SDS-1556MY, Revision 1, dated 3/27/97).
 
The analyses were performed in accordance with the methods specified in Geric !
Letter 90-05, and ASME Section XI Code Case N-480, as appropriate. The !
licensee's calculations predicting inoperability were conservative due to the use of worst case moments when evaluating the loads in the piping system. This assumption was necessary because the licensee does not have detailed pipe 1 stresses at all points in the SW system, but does have moments at certain points in the system from the design basis calculation of record (Stone and Webster Letter dated May 23,1966, and NUS Corporation Report TR-76-27 dated January 27, 1977). The inspector independently evaluated the structuralintegrity of the SW i piping using the licensee data and the GL 90-05 and ASME calculational methods to j verify the licensee's conclusion !
I l
l l
 
      :
-
-
improve procedural adherence discussed in the cover letter, additional information is  '
        ,
provided in the individual responses for each of the six examples noted.


l
      )
-
I
      ;
44    l The licensee implemented plans to replace the defective pipe section and to conduct a repair in accordance with the ASME code (details are provided in Section M1.1).
An engineering evaluation was completed on March 26,1997 to assess whether :
the normal supply lines to the spent fuel cooling system could remain in servic l The licensee concluded that continued use of the SW lines was justified until a j bypass was piaced in service as the repairs began during the week of March 3 l The engineering assessment considered the conservatism of the calculations; the l'
ruggedness and ductility of the piping, such that catastrophic failure was unlikely-the heat up rate of the pool and the availability of alternate cooling supplies for the )
SFP heat exchangers; the location of the degraded sections such that the safety '
function would be performed if the further degradation occurred; and, the consequences of flooding from postulated leakage in the SW li" should I degradation occur. As part of the corrective actions, the licen. oegan a program to examine more locations in the supply and return headers to identify the extent of piping degradation. The licensee adopted the approach of Generic Letter 90-05, which required that pipe exams be extended to 5 similar locations. The sample expansions would continue until no further unacceptable defects were found. No inadequacies were identified in the engineering evaluation, or in the initial plan to examine more SW piping. NRC inspector input was required to assure the e. appropriate second expansion sample was selected, due in part to the failure to fully integrate the technical support from the site and design engineering group Past NRC inspections reviewed licensee actions to disposition degraded conditions '
in a timely manner (reference lospection 94-05, 94-07, 94-14 and item 96-06-05).
As a result of the decision to permanently shutdown the plant, the licensee identified that all areas of the service water system was considered to be at risk to MIC degradation. This included all main supply header piping, all branch  1 connections, the emergency diesel generator supply and return piping, and the i spent fuel pool cooling supply and return piping. The licensee's program for MlC prevention continued to use Bulaab 8007 to combat potential MIC. However, Bulaab injection was terminated af ter the initial startup of the system in mid-1996 due to mechanical difficulties and problems with the injection skid. Following the preoperational test phase of the new system, the licensee had discontinued Bulaab injection and had not resumed the syste.m operation as of March 1997. This was caused in part by uncertainties (and erroneous assumptions) regarding the status of the SW system following the decision to decommission the plant. The licensee has since recognized the need to continue Bulaab injection as a M!C mitigation measure for as long as the SW system has a safety function in the plant design and licensing basis. Licensee actions were in progress at the end of the inspection period to commence Bulaab treatmen The short term actions included repair of known SW corrosion defects and examinations of the SW system to evaluate the overall status of the pipin However, the licensee recognized the need to provide an alternative to SW for long term cooling of the spent fuel pool. The licensee had developed plans as part of the Decommissioning project to create a " nuclear island" around the spent fuel pool, which would eliminate the reliance on the SW system. The licensee decided this period to accelerate the development of the nuclear island that would allow the
  .. -.  - . - _ . - -
.
: .      ,
partialimplementation by the end of 1997 of the mechanical portion of the alternate !
SFP cooling system. The licensee briefed the inspector on the conceptual desig The new system included the use of an intermediate cooling loop that used demineralized water and which would tie into the SW side of the SFP heat  ,
exchangers. The intermediate cooling loop would include the e7 of spray coolere, which would be supplied make up water from an another new ' stem. Neitho of the new systems would be susceptible to MIC corrosio The licensee addressed these issues and his plans to address them in ietters to the NRC dated March 31, and April 2,1997. Licensee activities were in progress at the conclusion of the inspection to complete the actions described above. This item is ;
open pending further NRC review of the licensee actions to assure adequate systems for the spent fuel pool cooling. Specifically, the item is open pending (l)
further NRC review of licensee implementation of the program to prevent and mitigate MIC corrosion in the SW service water system; (ii) the completion of s licensee actions address degraded welds and piping; (ii) the completion of licensee '
actions and evaluations to assure that the existing SW system will remain acceptable source for SFP cooling for as long as the SW system is needed to 1 provide that function; and (iv) the completion of license long term actions to install
; an alternative cooling system for the SFP (URI 97-01-08).
E1,3 _Qonclusions for Conduct of Enaineerina Mixed performance was noted in the engineering support of operations. Engineering i
      '
performance was good in response to emergent design basis issues and corrosion degradation in the SW system. Engineering evaluations were good to identify the inoperabilities in the SW piping, to assess the acceptability of interim use of the degraded system, and to support the corrective actions to address the design and material discrepancy issues. Engineering support for design changes was good to provide timely corrective actions to mitigate the design and corrosion induced )
problems. Exceptions to good performance were noted in the failure to fully integrate site and corporate engineering support, and to assure continued chemical (Bulaab) treatment of the service water system during shutdown conditions. Poor performance was noted in the f ailures to track design issues to completion (waterhammer), to properly classify design issues (two phase flow) for the  !
shutdown condition of the plant, and to track commitments to the NRC. These j issues appear as additional weaknesses in the corrective action process to assure j that conditions adverse to quality are promptly addressed to preclude recurrenc !
E3 Engineering Documentation - Design Basis Discrepancies (40500)
Severalissues were identified during the period which appeared as additional examples of problems in defining or implementing the plant design basis, in meeting licensing commitments, and in completing effective corrective actions. The examples are summarized belo .
.
.
j Restatement of Violation 1.a.
:
i Work Control Manual WCM 2.4-1, Equipment Tagging, Revision 7, was written  ,


E _H_gndlina Loads Over Stored Fuel On February 19, the licensee notified the NRC per 10 CFR 50.72 (b)(1)(ii)(B) of a condition constituting plant operation outside the design basis. The condition also was a past violation of Technical Specification (TS) 3.9.7. TS 3.9.7 was implemented in 1989 and states that loads in excess of 1650 pounds shall be prohibited from travel over fuel assemblies in the storage pool. The limit was chosen to ensure plant practices were consistent with assumptions used in the accident analyses. The TS bases states that the restriction of loads in excess of the nominal weight of a fuel and control assembly and associated handling tool ensures that, in the event the load is dropped, (i) the radioactivity released will be limited to that contained in a single fuel assembly, and (ii) any possible fuel distortion will not result in a critical array. The weight of 1650 pounds was considered to be the nominal weight of the fuel, control rod and handling tool combination. During reviews on February 4,1997 (ACR 97-56), the licensee discovered there was no documented basis nor calculation justifying the load limit of 1650 pound The nominal weights of fuel and associated components varied for different fuel cycles. The nominal combined weights for the components used between 1989 c and 1994 was 1690 pounds [? 187 pounds (fuel assembly) + 153 pounds (control ,
pursuant to the above and states in Step 1.2.2 that any equipment tagged with a l red tag shall not be operated by anyone. Contrary to the above, on February 19, 1997, a contractor health physics technician opened the red tagged prefilter
      ;
'
rod) + 350 pounds (handling tool)]. From 1994 to 1996, the nominal weight was
access door for auxiliary building filter FL-70-1 A.
      '
: Reason for the Violation i
approximately 1725 pounds [1250 pounds (fuel assembly) + 175 pounds (control  1 rod) + 300 pounds (handling tool)]. The licensee was in conformance with the design basis when the load drop analysis was revised in 19S iur loads up to 2300 ,
(
- pounds (vendor report HI 941225). However, the, TS limit was not revised and !
l This violation was caused by inadequate human performance. l
remained at 1650 pound !


- The licensee identified this matter during reviews in response to an NRC request for information contained in a letter dated January 9,1997. The licensee responded to l the NRC in a letter dated February 19,1997 (B16185), and committed to assess  i the need for a TS revision to remove any ambiguity. The licensee stated that no fuel would be moved in the spent fuel pool until this issue was resolved. The completion of licensee actions to assure the movement of any load over in the spent fuel pool is in conformance with the limits in TS 3.9.7 is considered an open item, and is Part A of unresolved item 97-01-0 E;L2 Control Room Habitability
An apparent cause determination was performed which concemed opening the
; door to a filter housing without having the red tag cleared. The ventilation door
;
had a red tag on it that was not a part of the work order under which the work was :
: being performed and which should have been removed prior to allowing work to  ;
l be performed on the filter unit. The technician opening the filter door assumed the l l red tag was part of the work order he was working on and did not check the red  ;
tag or his supervisor to determ' if clearance was required. The person showed a lack of understanding of the i ] process and a lack of a questioning attitude. i The technician had received trah...ig on tags as part of their initial site access and !
qualification training. l Corrective Steps that have been taken and the results achieved immediate corrective action was to counsel the technician and his supervisor on the importance of never operating red tagged equipment. Guidance was provided to station personnel, including contractors, stressing the importance of complying with the tagging program and never operating equipment with red tags. Each department head verified that a meeting with their employees had been held to discuss the event and reinforce the importance of following tagging procedures.


On February 7,1997, the licensee notified the NRC per 10 CFR 50.72 (b)(2)(iii)(D)
_
of a conditica that could have prevented the fulfillment of a safety function of systems needed to mitigate the consequences of an accident, in particular, to meet the NRC requirements for TMI Action Plan item til.D.3.4, the licensee credited the l use of self contained breathing apparatus to protect the operators in the control <
room as actions were taken to mitigate poctulated design basis accidents. The assessments of control room habitability did not use formal calculations subject to quality assurance checks, and was lacking in that an important input assumption, the amount of control room inteakage, was not verified. The licensee concluded that this condition was a violation of 10 CFR 50 Appendix B, Design Control, and


  .- - - .. - - - - -_ - . . . . _ . - - -
. .
,
,
 
  '
47 was also reportable per 10 CFR 50.73 (a)(2)(v). The licensee planned to submit a licensee event report to document this issue and describe the followup actions needed to resolve the discrepancy. The potential radioactive source terms from the postulated design basis accidents for an operating plant were no longer a concern for a permanently shutdown reactor. The licensee has yet to complete the accident analyses for the decommissioning phase, and intends to address the future design requirements for control room habitability when that work is complete. NRC concerns for this area are already tracked by inspection item 96-02-0 fyL2 Missed Commitments By letter dated January 6,1997 (B16047), the licensee notified the NRC of commitments that had not been met. In a previous response to a Notice of Violation dated August 21,1996 (reference inspection 96-04), the licensee committed to take several actions to correct a violation and a deviation from NRC .
U. S. Nucinar R:gulatory Commission
requirernents, and to take actions to avoid recurrence. The violation concerned an inadequate safety evaluation for moving fuelin the spent fuel pool (VIO 96-04-03),
and the deviation involved a lifting device that had not been inspected per the heavy loads program (DEV 96-04-02). During this inspection, the inspector reviewed the L following completed actions to assure all commitments were met:
  (i) UFSAR change request 06-CY-28 was completed on December 19,1996 to state in Section 15.5.2.2 that a minimum of 7 feet of water submergence will be maintained for spent fuel in transit in the spent fuel pool. The change request will be included in the 1997 UFSAR update submitta (ii) The Reactor Engineering Manager completed a review of the safety evaluation to justify use of a sling with the spent fuel handling tool. The review was completed on November 8,1996, and identified the weaknesses in the original evaluation that caused the violatio (iii) The licensee revised the training guides for crane operators (OJT Guides MM-510-025, MM-510-024 and MM-510-014) to require the verification that workors are physically qualified to the requirements of standard ANSI B3 The guides were updated by November 9,199 (iv) Work control procedure WCM 2.2-7 was revised in Step 1.4.3 to delete reference to a speciallifting rig, and to refer to rigging equipment when lifting the floor blocks in the primary auxiliary building. The procedure was rcvised on October 22,1996, and, (v) The licensee completed a review of other commitments to assure no others had been missed. The review was described in memorandum NL\CY 97-08 dated January 27,1997, which also identified the reasons why the commitments made in the August 1996 submittal were missed along with actions to assure future commitments would be met. The licensee attributed the past failures to new personnel assigned to the Nuclear Licensing
 
.
.
 
department, who were inadequately trained on the process for tracking commitments and assuring assignments were completed in a timely manne Further NRC review of other licensee actions in response to inspection 96-04 is described in Section E8.4 of this report. Except as summarized below, the inspector ha o further comments at this time on the actions by the licensee to address the vution identified in Inspection 96-0 E14 Service Water Desian Basis issues Inspection item 96-01-02 was open due to the identification of a number of discrepancies between the plant licensing and design bases, and the plant operating procedures and practices. One example of the concern was the licensee's discovery in February 1996 that the plant was operating outside the design basis due to river water temperatures that were colder than assumed in the design basis and as listed in the Updated Final Safety Analysis Report (UFSAR). Inspection 96-01 describes the licensee actions to address the technical concerns associated with this matter (Section 4.2). The issue was reported to the NRC on 2/7/96 per 10 CFR 50,72 (b)(1)(ii)(B) and as licensee event report (LER) 96-02 as operation outside the
-
design basis. Corrective actions at that time included the completion of an engineering evaluation to show that plant operation with temperatures as low as 28 degrees was acceotable. Planned corrective action included revising the licencing and design basis to reflect the new lower temperature limi On January 8,1997, the licensee identified additional discrepancies between the licensing basis and operating practices, as described in adverse condition report ACR 97-22. The licensee noted that the river water temperature was 34.6 degrees F, which was less that the design basis temperature of 35 degrees F. The licensee determined that, contrary to the intentions following the February 1996 temperature discrepancy, no action had been taken to update the UFSAR or to revise the design basis regarding the lower limit for SW temperatures. Thus, the SW system licensing and design basis remained 35 degrees F. The licensee completed a reportability evaluation for this matter (ACR 97-26) and determined on January 14, 1997 that the issue was reportable per 10 CFR 60.72 (b)(1)(ii)(B) as operation outside the design basis. LER 97-01 was submitted to the NRC on February 11, 1997. This LDB discrepancy was caused by the failure to complete the previous corrective actions to revise the design basis calculations and to update the UFSA The licensee had intended to revise the lower SW temperature Smit as part of the UFSAR revision planned in 1996, which was not completed. As further corrective actions, the licensee planned to develop a new tracking system to identify this type of commitment to ensure they would be completed. The licensee initiated a review of the remaining backlog of proposed UFSAR changes to assure there are no other similar discrepancies. As part of the configuration management plan, the licensee plans to formally identify and correct UFSAR inaccuracies. During its review of this issue, the licensee identified an additional programmatic deficiency in the treatment of the licensing basis,in that the update to the plant licensing basis was taken to coincide with the annual revisions of the UFSAR made in accordance with 10 CFR 50.71(e). The licensing basis needs to be a "living document" that is revised and
 
.
 
made accurate on an ongoing basis to assure the adequacy of safety evaluations completed per 10 CFR 50.59. This would require a system to maintain a site working copy of the licensing basis, inclusive of the UFSAR, which could be revised as design changes and safety evaluations were completed. The UFSAR update filed with the NRC could then be completed after the fact per 10 CFR 50.71(e).
 
A second licensing design basis (LDB) discrepancy described in ACR 97-22 concerned the process to chlorinate the service water. While reviewing practices for freeze protection of systems at the intake structure, plant operators noted that the injection of sodium hypochlorite had been suspended due to the plant operational status with no circulating water pumps in operation. This was contrary to UFSAR Section 9.2.1.1 which states that the service water system is continuously injected with sodium -hypochlorite at the pump suction. At the time of discovery in January 1997, only one of 4 service water pumps was required to be in operation to support the cooling of plant equipment with the reactor in the defueled condition. Further, all four circulating water pumps were secured and ;
were not required to operate in the shutdown condition (a circulating water pump ;
was started occasionally as needed to provide dilution water for liquid discharges).
 
Further, as a condition of the plant Nuclear Pollution Discharge Elimination System t (NPDES) permit, the hypochlorite must be isolated with no circulating water pum .
in operation. Licensee engineering concluded in an operability evaluation dated I January 14,1997 (CY-TS-0011) that the function of the service water system was not impaired because the intention was to chlorinate the SW system when the potential for macro and micro fouling existed, and the potential for this corrosion mechanism was low with cold river temperatures. The licensee addressed this area in a temporary procedure change (TPC 97-10) to NOP 2.20-4 which clarified that chlorination of the SW system is not required when river water temperatures are below 35 degrees F, and to assure that chlorination is not aligned for injection when intake freeze protection is provided by a slip stream from the SW pumps (which assured compliance with the NPDES requirements). The inspector observed operator actions to comply with the revised procedure instructions as river temperature varied around 35 degrees, and identified no inadequacie This item is unresolved pending the completion of licensee actions to (l) develop a process to provide a site working copy of the licensing basis that is updated continually to assure the adequate completion evaluations per 10 CFR 50.59; (ii)
complete a review of the backlog of proposed UFSAR changes for similar operability concerns; and, (iii) complete the 1997 UFSAR revision to address the design basis discrepancies discussed above (SW system normal operating temperature, SW chlorination, and submergence of fuel on the spent fuel handling tool). This item is unresolved pending the completion of the above items, and subsequent review by the NRC. This is Part B of unresolved item 97-01-0 IL3 2 5 Service Water System Water Hammer This issue, discussed in Section E1.1 above, concerns the discovery in March 1997 of an original design deficiency in the SW supply and return lines to the SFP cooling system. Licensee engineering evaluations found that two phase flow in the SW .
-
 
piping could result in water hammer and excessive piping and support loads in the
- event of a loss of normal power. The issue was reported to the NRC per 10 CFR ,
50.72(b)(2)(i) on March 11 as plant operation in a degraded and unanalyzed '
conditio ;
The issue was another example of inadequate past engineering evaluation of plant ;
systems and performance, which failed to assure the plant could operate  )
< satisfactorily for events considered in the design and licensing basis. The development and progress of this issue within the licensee's organization in the August 1996 to March 1997 time period also showed weaknesses in the processes i
      '
to classify, track and correct conditions adverse to qualit E3.6 Inocerable Effluent Monitor - Stack Noble Gas NRC and licensee QAS review of stack low range noble gas monitor, radiation !
monitor channel RM-14A, questioned the adequacy of the sample' probe geometry i
,
,
(see Inspection 97 02, Section R2.1). The sample probe is located in a ventilation t housing in the effluent pathway to the main stack. The sample probe consists of a single fueled head monitored in the middle of the ventilation duct. Technical e Specification 3.3.3.8 requires that the RM 14A be operable.at all times to provide
CY-97-058/ Attachment 1/Page 2 A Temporary Procedure Change (TPC) was written to WCM 2.1-2, Trouble Report (TR)/ Job Scoping. TPC 97-97 added a statement to the Tagging Required portion of the Evaluator's Checklist under TR/ Job Scoping Guidelines stating " perform walkdown of system / component to verify there are no conflicting tags on system / component that would affect work". This procedure change will identify tags from other clearances during the job scoping process.
'
indication of gaseous effluents via the stack, and be capable of providing alarm and
;
autvisatic isolation of the waste gas effluent On April 4, the licensee issued ACR 97-110 to document the results of an
, engineering evaluation-which concluded that the sample probe was not representative of the stack effluents, due to the potential for nonuniform mixing of the effluent flow steam as it passes the probe. The licensee reported this !
,
- discrepancy to the NRC at 11:15 a.m. on April 4 per 10 CFR 50.72(b(1)(ii)(B) as an l event or condition outside the design basis of the plant. The licensee initiated plans I to route the sample probe from radiation monitor channel RM148, which uses an isokinetic probe mounted in the main stack, to the input of channel RM14A. The resolution of this and other RMS issues (see also Section R2.1) will be tracked as followup to the NRC concerns identified in Inspection 97-0 i El Spent Fuel Buildina and Yard Crane Desian Basis issues ACR 97-34 was issued on January 20,1997 based on the findings by the  l configuration management plan working group that the calculations for the spent j fuel building and yard cranes do not support the conclusions that the cranes are acceptable for handling design basis loads. Discrepancies were identified in calculations CY 524990-178-GC, Revision 0, for the Yard Crane, and 94-NS-02-1050 CY, Revision 0, for the Fuel Building Crane. The issues included: the objectives stated in the safety evaluation for the calculation were not supported in the body of c61culations; the adequacy cf modeling; the adequacy of the acceptance criteria with the exclusion of safe shutdown earthquake loads, the adequacy with which wind and tomado loads were addressed; the adequacy of the l assumptions used to address critical connections; the adequacy of the analysis of i
      !
      ;
    -


_  . ~ _ _ _ _  _ .
A second TPC 97-98 was written to WCM 2.1-1, Work Control Process. This TPC added a statement to the step on verifying tagging under the Work Performance section for the First Line Supervisor / Designee ". . . to verify there are no conflicting tags on system / component that would affect work".
.


members critical to the lateral strength of the structure; and, the adequacy of the ;
Corrective Steps that have been taken to avoid further violations Disciplinary action was taken against the employee. The TPCs have been  l attached to the latest revision of the applicable procedures. A sign has been  l permanently installed on the ventilation door to eliminate the need for future red tagging.      .
documentation for the calculation This matter is unresolved pending the completion of licensee actions to resolve the issues identified in ACR 97-34, and subsequent review by the NRC. This is Part C ,
Date when full compliance will be achieved The above corrective action is complete and full compliance has been achieved. l l
,
Restatement of Violation 1.b.
of unresolved item 97-01-0 E33 Conclusions for Enaineerina Documentation (URI 97-01-09)
The design basis discrepancies discussed above were discovered either by good licensee staff and QAS initiatives to identify and resolve discrepancies, or by the Configuration Management Plan group as the reviews to reconstitute the plant design and licensing basis are completed. Licensee immediate corrective actions in response to the individual issues were appropriate. The existence of design and licensing basis discrepancies was a concern previously addressed by the NRC in past inspections and a pending enforcement action for inspections 96-06,96-08, 96-80 and 96-201. The identification of additional design basis issues is expected t until the completion of the CMP plan for the shutdown and deconmissioned plan ,
However, exceptions to good performance were noted in the failure to adequately )
--g address the design basis issue relative to SW temperature, which appears as an _ ;
example of a corrective action weaknes The completion of licensee actions to address issue described above that are important to plant safety during decommissioning will be reviewed in subsequent
. NRC inspections. Specifically, this item is open pending the completion of licensee actions to resolve (i) the handling of loads over the SFP; (ii) service water system design basis issues; and, (iii) the analytical issues for the spent fuel cranes (URI 97-01-09).


E6 Engineering Organization and Administration EQl Corrective Action Proaram Weaknesses QAS Surveillances and Audits Several findings by the QA audit and surveillance groups during the period demonstrated good performance by the overs ~ight groups to identify deficiencies in .
Procedure NOP 2.19-8A, Auxiliary Boiler Operation, Revision 2 dated January 30, 1997, requires in Step 6.1.12a that the operator maintain the condensate receiver tank one-half to two-thirds full when operating the condensate rnakeup pump manually. Contrary to the above, on January 5 and 11,'1997, an operator began a manual fill of the auxiliary boiler condensate receiver tank, and then left the area. i The condensate receiver tank overflowed on both occasions, resulting in an  j ur. planned discharge to the environment.     !
plant operating activities. Some examples included the following deficiencies: l
Reason for the Violation:
, inadequacies in E-Plan remedial training (ACR 97-122); the failure to properly control software in the radiological effluents program (ACR 97-32); adverse trends in the reliability of the radiation monitoring system, which indicated the RMS was I not meeting the design basis (ACR 97-73); SIP CY-P-97-008: the failure to  l implement the radwaste program in a manner to ensure compliance with 10 CFR 61 (ACR 97 76); the lack of an isokinetic probe for RMS-14A, and a design deficiency in which the sample nozzle was not rated for all flow conditions (ACR 97-110); the failure to adequately implement the radiological environmental monitoring program !
        <
        '
      .
, (ACR 97-139) and, Audit A24057/A25119: the failure to perform the 1996 annual update of the Fire Hazards Analysis as required by procedure NGP 2.1 !
The reason for this violation was a failure to follow procedures.
        ,
 
.
.
 
Effectiveness of Corrective Actions The failures to complete corrective actions, including the failure to complete committed actions, were examples of ongoing weaknesses in the licensee's program tc meet 10 CFR 50 Appendix B, Criterion XVI, which requires that the licensee identify and correct conditions adverse to quality. The licensee has plans to address this area and to improve performance as part of the actions taken and in progress in response to inspections 96-06, 96-08, 96-80 and 96-201. The plan to improve the corrective action program at Haddam Neck prior to decommissioning the plant is described in Sections 12 and 19 of the Decommissioning Project Manual. A dedicated manager has been assigned to administer the ACR program, who is also responsible for enhancing the process. Actions in progress to improve the corrective action program include developing: a new simpler ACR process to assure problems are identified at a low significance level; a simpler action tracking system to replace the existing system (action item trending and tracking system -
AITTS); new procedures to standardize the causal factors used to evaluate problems to assure consistent evaluations of causes station wide; new procedures for tracking and trending problems to assure programmatic and reoccurring causes are easily recognized and effectively corrected; and, initiatives to develop and !
- promulgate management standards at the department level to ensure participation and open communications at alllevels within the new organization. The actions to improve the corrective action program at Haddam Neck is expected to be completed during the second quarter of 199 Despite the progress in identifying deficiencies, the licensee had demonstrated continued weaknesses in developing and implementing adequate corrective action Some examples of the deficiencies in this area included: ACR 97-122: the failure to develop implement and conduct remedial training to assure effective corrective action in response to the August 1996 Emergency Plan Exercise (reference
      '
Inspection 96-07); ACR 97-124: management failure to develop, implement and monitor corrective action commitments made in response to the November 2,1996 contamination event in a manner that will prevent recurrence; ACR 97-31 the failure to take or provide adequate documentation of corrective actions for problems identified in Level D ACRs; ACR 97-59: issued on February 6,1997 due to the OAS finding that due dates for ACR action items were being inappropriately extended; and, ACR 97-134: the failure to investigate and take adequate corrective actions in response to a previous deficiency with the auxiliary building filters (showed high differential pressure - contributed to failures this period).
 
The licensee recognized the need to improve the tracking and trending of action items placed into the corrective action system as a result of adverse condition The licensee issued memorandum DCY 961004 on March 26,1997 to address weaknesses in the process defined by ACP 1.2-16.5. The licensee issued guidance that would improve the traceability of ACR assignments by issuing a single AR number for a given ACR, and then making sub-ascignments as necessary for all associated actions. The guidance also clarified the instructions on making assignments and clarified the relationship between the ACR and the AR status (i.e.,
 
  .. . -- . . . - . - - .-- . -- .  .-
.
        !
i
*
        '
.


        .
the ACR remained open until all evaluations are completed and tracked until all  !
,
,
associated ARs are complete >
in these two events the operators did not follow instructions to keep the tank at the prescribed level. The operators had chosen to manually fill the tank in each case ;
'
because of reliability concems of the  automatic makeup valve.
Further examples of weaknesses in the corrective action program were noted during  ;
this period as events occurred that were a repeat of past problems. The examples included the deficiencies in the tracking and timely completion of surveillance  i activities (see Section M4.1); the recurrence of personnel performance areas in
,
broad areas of plant operations (see Section 01.8. 01.9, 06.1, M2.1, M3.1 and i S1.4); the failure to preclude plant operation outside the design basis (river water temperature); and, the failure to assure quality in operating and test procedures (reference irs 96-80,96-11 and 97-02). Additional examples of weaknesses in corrective action in the engineering area are described in Section E1.3 and E Further examples were noted of repetitive pr recurrent deficient personnel and
        !
program performance, such as: ACR 97-063: failure to verify completeness of
-
training program requiremer ts for the 1995-1996 licensed operator initial training program and failure to assure complete and accurate information to the NRC on Form 39 Licensee actions to address these concerns will be tracked as part of the  !
*- enforcement action for Inspections 96-06, 96-08, 96-80 and 96-201, and in  *
, particular Inspection item 96-201-19. In summary, while the number of deficiencies  ,
identified by licensee staff has generally increased under the ACR program, a large
; number of deficiencies were identified either by self disclosing events, by the
!
independent oversight groups (quality organization, NRC). The licensee has yet to complete its initiatives to improve the corrective action program to assure consistent quality root cause investigations and to implement effective corrective
+
action !
E8 Miscellaneous Engineering lasues (92902)
EHd Soent Fuel Pool Desion to Suocort Full Core Off-load Y
a. Inspection Scoce
'
in response to a finding at another facility that " full core off-loads" had been performed when prohibited by the licensing basis (UFSAR), the regions were instructed to review all facilities for similar problems, b. Observations and Findinos By letters dated August 21 and December 19,1975 (as supplemented by three other letters), the licensee provided a preliminary design concept and an application for a license amendment to increase spent fuel pool (SFP) storage capacity from 366 to 1172, respectively. These letters show the licensee's intent to always provide for full core off-load. Amendment 7, authorizing this change, was issued June 8,1976.' The original FSAR stated that the SFP cooling system was capable of removing the residual heat produced by one and one-third cores while
, _  -. _ _ _
      - _ _ - _. _ _ _ - -
 
l l
 
l i    54 maintaining the pool water temperature below 170 F. No specifics of the heat balance were provide ]
The question of SFP cooling and storage was addressed during NRC inspection 50- l 213/77-02, when an unresolved issue was identified. The licensee was increasing ,
SFP cooling and storage capacities in accordance with Amendment 7, however, the j new fuel racks would be installed before the additional SFP heat exchanger and a j second pump were scheduled to be installed. During inspection 50-213/77-11, the ,
      '
unresolved issue was closed bases on limiting the SFP storage capacity to the original 336 spent fuel assemblies until the parallel heat exchanger and pump were l operabl l By application dated March 31,1995, as supplemented by letter of November 14, 1995, the licensee again requested an amendment to increase the SFP storage capacity to 1480 fuel assemblies. This was to allow full core off-load through the ;
end of the operating license,2007. The heat balance for this change was for a ]
river inlet of 90 F, a SFP maximum of 150 F, and a heat load of 22.4E6 Btu /hr. The j request was approved (with special agreed upon surveillance, analysis, and !
restrictions) by Amendment No.188, issued January 22,1996. This amendment l also added a new LCO which requires that a delay of fuel movement to the SFP
      '
"
was dependent upon the river water temperature. The current UFSAR continues to l allow full core off-load Conclusions
 
The inspector concluded that the Haddam Neck license including the TS and the UFSAR have never prohibited full core off-loads. The issue of cooling system ;
design for full core off-loads was also addressed in IR 50-213/95-20, Section 4.3.- !
This issue remains resolve E82 (Closed) IFl 94-09-03. HPSI Relief Valve Setooint Drift During inspection 50-213/94-09, the root cause for the premature lifting of the high pressure safety injection (HPSI) system discharge relief valve had not been determined although CYAPCO assessment of the short-term operability was appropriate. During inspection 50-213/95-19, CYAPCO implemented short-term actions to replace the valve spring and process a set-point change request to allow ;
a revised setpoint. In discussions with the valve vendor, the issue of horizontal mounting the valve providing uneven seat loading was identified and the licensee initiated plans to reorient the relief valve. Due to the December 5,1996 notification to permanently cease operation of the Haddam Neck facility pursuant to 10 CFR 50.82(a)(1)(l), the HPSI discharge relief valves are no longer of safety importance. This issue is considered close ,
f W
 
__
 
MJ (Closed) VIO 96-04-03. Inadeauate Safety Evaluation  j During the review of fuel handling activities documented in IR 50-213/96-04, two i cases were identified where no written or inadequately written safety evaluations existed. First, for refueling activities conduct prior to May 1,1996, the fuel handling tools providing less than 8 feet of fuel submergence contrary to the design documents, and a defacto change to the facility as described in UFSAR Section 15.5.2.2 and 9.1.4.2. Second, a modification.made to the fuel handling tool for the North Spent Fuel Building crane on May 2, adding a sling to increase the length of the hoist-tool configuration by one foot, was made with an inadequate safety evaluation. The SE did not address the effect of the change on the normal operation of the fuel handling equipment for the safe handling and storage of spent fuel. The modified fuel handling tool, in part, caused an irradiated fuel assembly suspended from the spent fuel building hoist to be incapable of safe storage for 25
  '
hour in the August 21,1996 response to these violations, CYAPCO stated that an approved safety evaluation and a change to the UFSAR to make the licensing basis documents consktent at 7 feet minimum submergence was made. In addition, a
,- comprehensive review of the safety evaluation prepared to initiate the fuel transfer evo;ution was performed. The inspector reviewed the Safety Evaluation, dated August 13,1996, the UFSAR proposed update, and an internal Memo, Minimum i We.ter Depth for Fuel Movement at Haddam Neck, dated May 28,1996. This  !
review indicated that appropriate corrective actions were taken. This issue is ;
consiclered close @4 (Closed) Ifl 96-04-04. Heavv Load Controls During inspection 96-04, the inspector questioned why the mor.cra!? above the emergency diesel generators (EDG) did not have specific procedural heavy load program controls. The licensee said that controls would be during trouble report job scoping and informal involvement of the system engineer when questions are raised by the workers. The licensee believes tnat most heavy load lifts for the diesel generator occur during outages when the equipment is considered inoperable. The issue was left open for review after licensee corrective actions were complete The licensee modified Work Control Manual (WCM) 2.2-8, Control of Heavy Loads, Revision 4, effective October 22,1996 to include a constraint that no loads will be lifted with the monorails over the EDG unless they are out of service. The inspector found the procedure change acceptable and, therefore, this inspector followup item is close BJ (Undate) URI 96-06-06. Batterv Oscillations and Ground During inspection 50-213/96-06, an issue related to ongoing current oscillations and noise problem existing in the dc battery charger "A" and in the de system ra.ses potential for a long term impact on the battery. This highlighted the need for monitoring the battery conditions until the battery is replaced and these issues are
 
_ _  _ _ . _ _ _ _ _ _  __
.
        !
        ~
 
satisfactorily resolved. This issue was left unresolved pending the licensee's  -
satisfactory resolution of these issues and further review by the NRC. The licensee i stated that they expect to have the response completed by May 20,1997. This issue remains unresolved until then .
  ~
ERJ (Closed) URI 96-201-12. Analysis Sucoort,inLPSI Pumo Shutdown During inspection 50-213/96-201, the team reviewed UFSAR Chapter 15.3. regarding accident sequence timo lines, as well as the licensee's integrated safety evaluations (ISE)/CY-93-005, dated June 28,1993, regarding changes to the CAR fan logic and CY-95-013, dated March 15,1995, regarding stopping of the LPSI pump (s). The team found that the licensee's technical basis for securing LPSI  t pump (s) in certain EOP scenarios did not comprehensively address the broad spectrum of design-basis LOCA break sizes, as well as credible single 4ailure  1 scenarios as required by 10 CFR 50.4'5 and Appendix K. This issue was left  !
unresolved pending additional NRC review of the alytical bases for the
. interruption of ECCS flow during transition to sump recirculation. Due to the  ,
decision to permanently cease operation of the Haddam Neck facility, the CAR fan logic and stopping of the LPSI pump issues are no longer important to the safety of  i
  ;- the shutdown, defueled, and later decommissioned reactor. These issues are-  ,
i considered close ERJ. (Closed) URI 96-201-31. RWST Instrument Calibrations
- During inspection 50-213/96-201, the team found that reactor water storage tank (RWST) Level Instrumentation Surveillance Procedures SUR 5.2.10, RWST Analog Channel Operational Test, (Revision 12), and SUR 5.2.68, RWST Level Calibration, s (Revision 11), did not correlate the transmitter output for the calibration points, or the actuation of alarm trip and reset points, to tank water volume on the basis of the physical configuration of the installation. In addition, the team found that there was no basis to correlate the actual height of water above the instrument elevation >
  (hydrostatic head) to the level accuracy measurement indicated in SUR 5.2.10. In response to the team concerns in this area, the licensee stated that a calculation  '
was not available which directly provided the basis for the RWST level alarm setpoints and the calibration procedures. Lack of a design calculation for the RWST level alarm setpoints and indication is contrary to the guidelines presented in procedure SP-EE-320, " Guidelines for Calculating Instrumentation Setpoints for Safety Systems," Revision 1, dated May 23,1994; Americt n National Standards Institute (ANSI)/ISA-S67.04-1988, "Setpoints for Nuclear Safety-Related Instrumentation," and RG 1.105, " Instrument Setpoints for Safety-Related System,"
Revision 2, dated February 1986, which are referer>ced by procedure SP-EE-32 This issue remained unresolved pending completion of NRC review of the licensee actions to ensure all calculation factors are properly considered in the RWST level instrumentation calibration procedure. The licensee placed SUR 5.210 and SUR 5.2-68 in the "Do Not Use" category on December 13,1996. Due to the decision to permanently cease operation of the Haddam Neck facility, the reaci.or water storage tank levelissues are no longer important to the safety function of the
_ ______ _ _ _ - _ _ _ _ _ _ _


.
.
.


shutdown, defueled, and later decommissioned reactor status. These issues are considered close fjLQ IOpen) URI 96-02-03. Control Room Habitability This item was last reviewed in inspection 96-11 and was open pending NRC review of licensee actions to assure the control room environment remained acceptable for the operators followirig design basis events. The radioactive source terms from postulated design basis accidents for an operating plant were no longer a concern for a permanently shutdown reactor. Licensee reviews were in progress during this inspection as part of the configuration management plan (CMP) to recreate the plant design basis. The licensee intends to complete the accident analyses for the decommissioning phase, and intends to address the future design requirements for control room habitability when that work is complete. The CMP group will also review the commitments made under the integrated safety assessment program (ISAP) to determine which items need to be maintained for decommissioning in general, and control room habitability in particular. This item remains open pending the completion the licensee actions under the CMP to disposition control room habitability issues, and the subsequent review by the NR (Closed) URI 96-08-12: Containment Isolation Valves This issue concerned the circuits that controlled certain valves in the letdown and ,
  -
the feedwater systems in response to a signal to isolate the primary containmen I Due to the licensee's decision to cease plant operations and place the reactor in a permanently defueled condition, the containment isolation function for the valves is no longer required. The licensee plans no further action to address this issue (memorandum CY-TS-97-114). This item is close (Closed) URI 96-06-0_5: Timelv Evaluation of MIC Corrosion This item was open, in part, pending further reviews of licensee actions to disposition degraded conditions in a timely manner. The licensee addressed this concern in a revision to procedure ENG 1.7-131, CY Microbiologically influenced Corrosion (MIC) Prevention, Monitoring and Mitigation Program, Revision Specifically, in Step 6.1.1.g, the licensee required that, following piping examinations for MIC corrosion, any reports of nonconforming conditions would be immediately forwarded to design engineering for a preliminary operability evaluation (to be completed within 72 hours), with a final disposition within 14 days. The inspector noted that nonconforming conditions (pitting defects) identified during examinations of service water piping for MIC corrosion in March 1997 were promptly reviewed by design engineering for operabilit Section E1.2 describes additional SW system degradation caused by further corrosion. Item 97-01-08 will track further NRC review of the licensee's program to prevent and mitigate MIC corrosion in the service water system. This item is close ,
  . .
_ _ _ . _ _ _ _. _ . _ . , . _ _ _ _ _ _ . _ ___ _ . _ _ _ _ _ . _. _
        .
.
.
        ,
        :
L j- "
58    J l
7  LER 97-01: River Tamoerature Below UFSAR Limit    l
        !
  'This LER concerned the recurrence of the a'dverse condition in which the plant  !
i  operated outside the UFSAR design basis for minimum service water temperatur ;
        >
;  - This matter was also reported in LER 96-02. This matter is described in Section
      .
4  E3.4 of this report. This LER is close j
;  .      J LER 9617: Main Stack Samole Performed Late    1 4        ;
This LER concerned the untimely collection (27 minutes late) of a main stack sample  l following the plant shutdown on July 22,1996. This event was contrary to  ,
 
Technical Specification 4.11.2.1.3, which requires that stack samples taken within  -
;  8 hours following a plant shutdown. The licensee determined the event was caused  l by a programmatic failure when an excessive number of unprioritized sample  j
}!  requests contributed to the chemist's failure to take the sample as required. The  4
        ~I sample taken at 3:10 a.m. on July 23 showed that no elevated releases were in
  . progress. Corrective actions to preclude recurrence included a review and revision .E of procedures CHM 7.6-32 and CHDP 6.1-4 to improve the guidance for stack  ,
sampling, eliminate unnecessary samples during transients, and providing a "
a -m . prioritized checklist for collecting sample; The safety significance for this event was t-- , .i
- : low. However, the event was an example of an NRC concern in the timely .  !
"
completion of required surveillances. Past similar events were reported in LERs 94-  ,
,
,
23 and 91-27. NRC concerns for the proper completion of surveillances are  '
.
discussed in Section M4.1 above. This LER is close LER 97-02: Reactor Coolant Samole Not Taken
  ' This LER concerned the failure to collect and analyze reactor coolant system (RCS) .  ;
samples for chlorides and fluorides since November 15,1996. Chemistry personnel stopped the chemistry analyses at that time because it was believed it was no longer required with the plant in a~defueled condition. The deficiency was identified 1 on February 6,1997; the licensee reinstituted RCS sampling at that time. This  !
event was contrary to Technical Specification 4.4.7, which requires that RCS samples taken every 72 hours, with a mode of applicability of "at all times."
t The licensee reviewed past practices and determined that RCS chemistry sampling had also been suspended during past period when the reactor was defueled. The  ;
licensee determined the event was caused by the incorrect interpretation of the  I technical specification requirements. The samples taken on February 6 and  r thereafter showed that the TS limits for chlorides and fluorides were me >
Corrective actions to preclude recurrence included counseling technicians on the proper interpretation of the TSs, and reviewing all other surveillance requirements to
'
assure no other surveillances were inappropriately stopped. No others were  I identifie The safety significance for this event was low since the limits chemistry limits established by TS 4.4.7 were established to protect the reactor vessel against
,. degradation due to chloride stress corrosion cracking with the reactor operating at  ;
        ,
h
        .
  -
      -
  - __ . ._ _ _ . . _ _ _
.
      -
1 normal operating temperatures and pressures. Following the decision to cease plant
      '
operations and permanently defuel the reactor, the RCS will no longer be operated at 550 degrees F and 2000 psig. However, the event was an example of an NRC concern in the timely completion of required surveillances. Past similar events wers reported in LERs 94-23,9617 and 96-22. NRC concerns for the proper completion of surveillances are discussed in Section M4.1 above. This LER is close IV. PLANT SUPPORT R1 Radiological Protection and Chemistry (RP&C) Controls BL1 External and Internal Exoosure Controls Insoection Scope (83750)
      :
The inspector toured the radiological controls area and reviewed radiological controls including posting, barricading, and access control, as appropriate, to radiation and high radiation areas and contaminated areas. The inspector also reviewed radioactive contamination controls, in addition, the inspector selectively l
0 reviewed planned radiologicai controls for diving operations within the spent fuel l
transfer cana Observations and Findinas As a result of the November 2,1996, fuel transfer canal airborne radioactivity event (see NRC Inspection Report 50-213/96-12, dated December 19,1996), the licensee suspended all radioactive work that could si0nificantly challenge the radiological controls program pending completion of appropriate corrective actions. The licensee was implementing the corrective actions outlined in its December 9,1996,
' letter provided to the NRC. The principal corrective action was the required review
'
of any significant radiological work by the acting RPM and the work services director. A significant amount of planned program upgrades remain to be completed. A number of significant program upgrades were tentatively scheduled for completion on or about the end of February 1997 (see Section R8.1 of this report for a discussion of corrective actions).
Access to the radiological controlled area continued to be controlled through the use of general and specific radiation work permits. Personnel signed in on the permits electronically and obtained an electronic dosimeter. As part of the corrective actions, all radiation work permits were required to be reviewed and approved by either the radiation protection manager or the radiation protection supervisor. The inspector identified no on-going radiological work activities. Radiation and high radiation areas were selectively reviewed and noted to be properly posted and controlled. Contaminated areas were properly identified with some minor exceptions that were brought to the licensee's attention. The licensee appeared to be providing adequate planning and preparation for diving operations. At the tune of the inspection, diving operations, to complete securing of the blank flange in the fuel transfer canal was on hold due to licensee questions on work procedure ~ - - .- . -  .--  - - - - - - - - . . ..- --  .- . . . .
        . - . ,
. .
          $
'
:
60 Regarding radiological control of work activities, the inspector discussed the following observations with licensee representatives:    4
  - The licensee's work control memorandum dated December 12,1996, did not l
specifically address work activities in potential high risk alpha areas not
  ,
  ,
necessarily located in high radiation areas. In response, the licensee  -
U. S. Nucl::ar Regulatory Commission
indicated that all work activities would be reviewed for radiological controls concerns and all work with potential radiological concerns (e.g., primary  :
F  system piping work, sump work), would be handled in accordance with  !
~
controls outlined in the memorandu ,
,
  - The inspector questioned the level and consistency of radiological oversight of work performed as minor maintenance since minor maintenance work may not necessarily be fermally scheduled via the licensee's weekly work  ,
planning program, in response, the licensee indicsted that the radiological  {
conditions for minor maintenance work would be reviewed and the work would be formally scheduled and planned if it was deemed to be high risk  i
[  work. On January 16,1997, the radiation protection supervisor issued  j guidance regarding prohibition of performance of minor maintenance based  1
'
  -
on radiological conditions,      j l
,
,
The following additional observations were brought to the licensee's attention:  I
CY-97-058/ Attachment 1/Page 3 Corrective Steps that have been taken and the results achieved The involved NSOs were counseled, and guidance was issued to all Operations Department personnel on the subject of heating steam condensate receiver tank overflow. NOP 2.19-8A, "A" Auxiliary Boiler Operation, has been revised to include enhanced instructions for manual filling of the Condensate Receiver Tank. l A trouble report had been issued on this valve on February 11,1997. The Maintenance Department investigated the reliability concems related to the valve j and concluded no repair would be effective. l l
;  ,- The' licensee designated outdoor yard areas as portions of the radiological
Corrective Steps that have been taken to avoid further violations l The specific events and management expectations regarding compliance with operating procedures have been reinforced. The corrective actions have been incorpcrated into the latest revision of NOP 2.19-8A, "A" Auxiliary Boiler Operation ;
  - controlled area (RCA). Personnel egress main plant buildings (e.g., auxiliary j  building) to access the outdoor areas. The licensee did not require personnel to frisk for contamination prior to exiting the plant buildings. This was
l and NOP 2.19-88,"B" Auxiliary Boiler Operation.
"
considered a program weakness in that personnel may inadvertently track -
contamination outdoors from the main plant buildings. In addition, the
>  inspector noted that as the station entered an active decommissioning phase, c  the likelihood of inadvertent tracking of contamination outdoors would increase. The licensee indicated background radiation levels precluded installation of friskers. Further, the licensee indicated radiological  ;
characterization surveys in the outdoor RCA areas,in support of  !
          '
decommissioning activities, were to be performed and that the chemistry group performs sampling of storm drains within the yard areas for radioactivity. The licensee initiated a review of this matte The inspector identified one individual who appeared to be exiting an  j alarming portal monitor at the protected area egress point. The individual  '
was called back and re-monitored with no contamination identifie Subsequent inspector review identified that the personnel contamination  !
monitors at the protected area egress were subject to periodic false alarms due to electrical shorting conditions. The monitors were located next to doors which allowed rain and snow to be blown in or be tracked to the
    ~
          :
monitors. Although security personnel would reset the monitors and require  {
personnel to re enter the monitors, the inspector questioned the adequacy i and effectiveness of the monitors. Further, at the time of the inspection, e        i i
, ')


  .%.. ,  5  _._, .- - _ ,__ _ _ _ _ . . . _ _ _ _
l An evaluation to replace the valve is ongoing.


  . . - . . . .. _ -_ _ - --.
Date when full compliance will be achieved CYAPCO will complete the evaluation on replacing the makeup valve by July 31, 1997. There have been no new events for the auxiliary boiler operation related to this violation.
.
.


personnel appeared to tolerate the false alarms and no action had been taken to initiate repair actions or take the affected monitor out-of-service. T%
Restatement of Violation 1.c.
licensee subsequently took the affected monitor out-of-service and provided guidance to security personnel as to when appropriate personnel should be
,
notified to initiate repair Personnel who had received medicalisotopes for diagnostic or treatment purposes were permitted to exit the protected area and alarm the portal monitors without a secondary check by radiation protection personnel. The -
inspector noted that although security personnel maintained a list of such :
personnel and reset the monitors once the person alarmed it, no secondary i check, by radiation protection personnel, was performed to ensure radioactivity, attributable to licensee operations, was not inadvertently or purposely removed from the station. The licensee indicated this matter would be reviewe '
- As discussed below, the licensee was performing a feasibility study relative l to decontamination of primary systems piping and components to support xposure reduction during decommissioning. The inspector noted that the ;
      '
-
ystem decontamination may change normal radionuclide -  ,
,  :oncentrations/ ratios within systems and/or the radiological controlled are !
,
Consequently, the change may require revisions to instrument calibration i orograms, radiological survey programs, the 10 CFR Part 61 waste stream j enalysis, and the bioassay programs among others. The licensee indicated {
  :at these areas will be reviewe Conclusions Work restrictions were implemented in accordance with licensee commitments to the NRC. The radiological controls program was not challenged due to a conscious


decision to suspend all unnecessary significant radiological work. Personnel contamination controls for egress of station buildings to outdoor radiological controlled areas was considered weak. There was a need to review various radiological controls programs in anticipation of potential changes in the composition of radioactive contamination following chemical decontaminatio R1,2 ALARA Proaram Insoection Scope (83750)    l The inspector selectively reviewed the program to maintain occupational radiation ,
Technical Specifications 6.2.2.f requires that administrative procedures be developed and implemented to limit working hours of facility staff. Procedure NGP 1.09.was developed pursuant to the above, and requires in Step 4.6 that workers have no less than 8 continuous hours off between scheduled work periods and that workers do not work more than 16 hours in a 24 hour period,24 hours in a 48 hour period, and 72 hours in a 7 day period. Contrary to the above, on February .
exposure to as-low-as-is-reasonably-achievable (ALARA). In particular, the l inspector reviewed the licensee's evaluations relative to performance of pnmary j
6-6,1997, an operator worked two shifts without 8 hours off between scheduled work periods. In addition, the results of a licensee investigation on January 17, 1997, identified 6 instances over a four week period in July and August 1994 in which work hours for plant staff exceeded the overtime limits.
        '
system decontamination in order to reduce personnel radiation exposures to ALARA. The inspector also reviewed the licensee's ALARA program relative to support of decommissioning activitie l l
_


_
Reason for the Violation An operator worked two shifts with less than eight hours off between the shifts without the prescribed time off because the individual scheduling the operator was unaware that tumover time is included in the total time worked.
_  _ _ _ _
.
      >
.
;


b. Observations and Findinas At the time of the inspection, the licensee was evaluating the decontamination of the primary system to reduce radiation dose rates on primary piping, to support '
decommissioning. The licensee was in Phase 1 of the five-phase decontamination process. Phase 1 involved a feasibility study of among other matters, cost / benefit and decontamination methods including flow paths and solvent, and expected
;
person-rem savings. Phase 2 of the program invoNes engineering evaluation while phases 3,4, and 5 involve equipment setup, operations, and demobilization, respectively. The decontamination process is described in Electric Power Research Institute (EPRI) Document EPRI-TR-106386, Decontamination for Decommissioning, dated May,199 The inspector noted that the licensee established a 1996 refueling outage ALARA ;
goal of 420.7 person-rem and a 1996 yearly ALARA goal of 455.4 person-re !'
Due to the licensee's decision to cease operationc and decommission the station, the licensee sustained 178.6 person-rem for 1996 of which about 147.8 person- i rem was accrued during limited outage activities, including reactor defuelin Because of the cessation of work activities, the inspector was not able to critically t a evaluate ALARA program perforrnance during the outage, on a per-job basi However, the licensee was tracking exposure daily and publishing a daily exposure ALARA report. Performance was being tracked against an expected daily accrued exposure for all work activities. The licensee has established 1997 ALARA Goals and Targets by department. The 1997 annual goal was 17.5 person-rem with a target of 9 person-rem. The licensee was continuing to evaluate these goal The inspector made the following observations relative to the ALARA program and its capabilities to effectively support decommissioning activities:
-
The ALARA program was not structured to provide for a real-time systematic evaluation of accrued radiation exposure for work activities as the activities progressed. Specifically, there was no clear guidance or expectation regarding the level of review required to ensure that final ALARA goals / targets would be met. The inspector noted that subjective reviews were being performed. Consequently, it was not apparent that the ALARA program, as currently described / implemented, would be capable of providing effective evaluation of ALARA program effectiveness in support of decommissioning activitie The ALARA program, for job pre-planning purposes, permitted subjective evaluation, by job supervisors, of expected worker accrued radiation
<
exposur The ALARA program reviews did not provide for evaluation and control of committed dose equivalent in consideration of the licensee's identified concern with transuranic contamination (i.e., alpha contamination) in selected areas of the statio ~- _
.
ihe inspector noted that the licensee was aware of weaknesses in ALARA planning and had recently tasked ALARA personnel with performing ALARA pre-job reviews, in lieu of job supervisors. As of the end of this inspection period, the licensee had l revised ALARA program procedures to provide enhanced ALARA control Conclusions The licensee was performing a feasibility study of a decontamination process to reduce radiation exposures during decommissioning to ALARA. This was  j considered a good initiative. Weaknesses in the ALARA program, relative to ALARA l
      '
planning for initial and on-going work activities, were identifie B1.3 Radioloaical Environmental Monitorina Proaram (IFl 97-01-10)
1 Insoection Scoce (84750)    l The radiological environmental monitoring progra.m (REMP) was inspected against ,
Sections E.1 and E.2 of the Radiological Effluent Monitoring and Offsite Dose !
Calculation Manual (REMODCM), Regulatory Guide 4.1, " Programs for Monitoring -
I
-
Radioactivity in the Environs of Nuclear Power Plants," and the Updated Final Safety Analysis Report (UFSAR). The following activities were conducted to assess the licensee's capability to implement the progra Examination of the air samplers and automatic water compositors; determination of operability and calibration statu Observation of personnel collecting samples from selected sampling location Discussion of any modifications to the REMP regarding decommissionin Review of any REMP procedure and ODCM change Review of the results of the Land Use Censu Review of control locations to wind direction and D/ Review of sample results and assessment of licensee's evaluation method Discussion of tritium in the wells adjacent to cana Discussion of with licensee compliance with EPA limits 40 CFR 19 Observation and walkdown of the protected area fence to review the location of TLD l l
l .
b. Observations and Findinal The Radiological Assessment Branch (RAB) continued to maintain oversight for the implementation of the REMP, including overall responsibility for quality assurance oversight within the program and most meteorological monitoring program responsibilities. Members of the Production Operations Services Laboratory (POSL)
had the responsibility to implement collection of samples of erwironmental media such as water, soil, sediment, fish and airborne particulates. The environmental samples were prepared and sent to the contractor, Yankee Atomic Environmental Laboratory, for routine analyses. Other responsibilities of the POSL included exchanging and reading environmental thermoluminescent dosimeters (TLDs) and calibrating and maintaining the air sampler The inspector visited the POSL and selected sampling stations to determine whether samples were being obtained from the locations designated in the REMODCM and whether air samplers were operable and calibrated. The sampling stations included air samplers for particulate and airborne iodine, milk sampling stations, a sediment sample location, the composite water sampler, and a number of thermoluminescent dosimetry (TLD) stations for direct ambient radiation measurements. All air e samplers and the water compositor at the selected locations were operational since the previous inspection. The observed air sampling equipment was well maintained, and the associated air volume measurement equipment was calibrated. The air samplers were collecting the amount of air (1-1.5 cfm) as required in the procedur Milk samples were available and the TLDs were placed at locations designated in the REMODCM. Sediment, water, and air filter collections were observed. The sampling techniques were good and performed according to the appropriate procedure The licensee discontinued sampling for air iodines as a result of the decision to decommission site. The requirements were changed in the ODCM, effective February 28,1997. Air iodines were sampled and analyzed through March 3,1997, as required in the previous revision of the REMODCM. The licensee expects this change to be the only change to the REMP as a result of decommissioning. All other aspects of the REMP will remain the sam The air sampling and sediment sampling procedures were reviewed and determined to be acceptable with one exception in the air sampling procedure. The procedure had not been revised to reflect the discontinuation of sampling and analyzing for air iodines at CY. This procedure was used for both the CY and Millstone site Millstone continues to sample and analyze for air iodines. The licensee stated that this procedure will be reviewed and changes will be made where appropriat The inspector reviewed the licensee's TLD program conducted at POSL. Overall, the program was acceptable. The ionization chamber (condenser R-meter) used to verify operability of the Shepherd Panoramic irradiator and ensure thermoluminescent dosimeters are accurately irradiated, was calibrated in January 1997 (see Section R1.1 of the Integrated Inspection Report 50-245; 50-336;50-423/96-09 for details). The results of the calibration showed an
      - - _ _ _ . _ _ _ - - - _ _ _
.
.
m


insignificant difference between the "as found" and the "as left." From these results, it can be concluded that the R-meter verified operability of the Shepherd Panoramic irradiator and ensured the thermoluminescent dosimeters were accurately irradiated prior to January 199 The Land Use Census required by the REMODCM was conducted using Procedure RAB B-7, " Environmental Sample Location Census." The census, performed in 1995, detailed garden and milk locations within 5 miles around the site. Changes as a result of the census will be documented in the 1996 Annual Radiological Environmental Operating Report (REOR). Also, the inspector reviewed the locations of control stations for milk and air sampling. Control stations appear to be in the least prevalent wind directions and lowest D/Q, as require Analytical data from certain environmental samples from January 1996 to March 1997 were reviewed by the inspector. In general, no significant anomalies or increased concentrations as a result of plant effluents were noted. The inspector noted that the required environmental LLDs were met. The inspector also reviewed the tritium (H-3) levels in the onsite well (#15) located adjacent to the discharge canal. The levels of H-3 in the well have historically been detectable and are e attributable to discharge of H-3 from normal plant effluents. Groundwater samples-were collected quarterly from the site welllocated one half mile ESE from the site, and from the control station located about 2.8 miles SE from site. The highest concentration detected at well #15 was 2.0 i O.2E3 pCi/l collected in the first quarter of 1995. Subsequent concentrations were lower, in fact, tritium was not-detectable during the fourth quarter of 1995. In 1996, the highest concentration detected was 1.5 i O.5 E3 pCi/l in third quarter and in 1997, the concentration for the first quarter was 9.91 i O.5 E2 pCill. The reporting level for H-3 is 2.0 E4 pCi/l, as required by the REMODCM, Section E.1, Table E-2. Since shutdown in September 1996, these levels have gradually decreased. The annual dose, reported in the 1995 annual Radioactive Effluent Release Report, from liquid effluents for an individual in an unrestricted area from all pathways of exposure was 0.176 millirem, which was less than the annual limit of 3 millirem to the total body, as required by T.S. 3.11. The inspector discussed compliance with the Environmental Protection Agency (EPA) regulation 40 CFR 190 with the licensee. The inspector walked along the protected area fence with a member from POSL and noted the location of the extra environmental TLDs (documented in the REOR). A set of TLDs deployed by Health Physics (HP) was also on the fence. The inspector determined that the environmental TLDs posted on and near the protected area fence were considered as extra environmental TLDs and were not intended to measure direct radiation from the site to demonstrate compliance with 40 CFR 190. The TLDs for the HP proram were only considered for on site measurements to ensure compliance with cer%in 10 CFR 20 requirements. This inspection area was incomplete and further dit.ussion and information regarding how the licensee ensure compliance with 40 CFR 190 remains to be reviewed. (IFl 97-01-10).
  - -
 
. _ _ - _ _ _ _ _ _ _ _ _
  .
66 Conclusions Overall, Haddam Neck continued to maintain an effective REMP. Environmental sample media were collected and analyzed as required by REMODCM. Sample collection was performed according to the procedure and the sampling technique was good. The TLD program, including the quality assurance, was good. The R-chamber had been calibrated in January 1997 in response to an NRC finding during a REMP inspection at Millstone. The "as found" compared wall to the "as left" results of the R-chamber. The discontinuation of collecting and analyzing for air iodine was appropriate and the change made to the REMODCM prior implementation  .
was also appropriat R 1.4 Meteorolonical Monitorina Proaram (MMP) Inspection Scope (84750)
The Meteorological Monitoring Program (MMP) was inspected against TS Section 3.3.3.4, UFSAR Section 2.3.3 and Regulatory Guide 1.23 commitment The following activities were conducted to assess the licensee's ability to implement the progra Review of calibration procedures, calibration results, and channel check log Discussion of data acquisition, availability of data, and EDAN parameter Observation of coMition of meteorological equipmen Observations and Find Personnel from POSL are responsible for maintaining and calibrating the meteorological monitoring instrumentation. A member from POSL, familiar with the monitoring instrumentation and data acquisition, demonstrated Environmental Data
, Acquisition Network $ DAN) parameters. The licensee had recently upgraded the EDAN2 to EDAN3. Data acquisition is now personal computer (PC) base Meteorological data can be acquisitioned from any one of the many meteorological monitoring locations including the control room, emergency operating facility, the
  -
  -
POSL office, and from the Millstone meteorological towers. The UFSAR was
.      .
'
reviewed regarding the upgrade. No changes were made since the same data can be requested and the UFSAR does not stipulate a specific version of EDA Calibration checks were performed weekly in accordance with the weekly surveillance. The weekly calibration check was not observed during this inspection due to scheduling changes made by the licensee. However, the inspector reviewed the results of that particular surveillance and noted that no problems were encountered during the surveillance. Selected calibration checks from June 1995 through March,1997 were reviewed. The inspector noted that the licensee is
,
capable of resolving problems, when encountered. Most problems were minor in nature due to the high level of attention to the equipmen _________ __
*
l l
l
'
l
 
Calibrations were performed quarterly, more often than semiannual as required by TS. Calibration results from July,1995 through December 1996 were reviewe The results were within the licensee's acceptance criteria. Calibrations were performed according to the licensee's procedures. The meteorological monitoring procedures are PORC-approve Channel checks are required daily by the TS. Operations personnel performed the channel checks every shift, thereby ensuring compliance with TS. The inspector selected and reviewed the shift log from February 24 to March 19,1997, and noted that the channel checks were performed every shift during that time perio The inspector noted a new multi-point chart recorder located in the equipment house of the primary tower. The associated calibration procedure was updated to reflect the new chart recorder type. The inspector determined that a change to the UFSAR was not required because it is not prescriptive regarding the type of recorder. A new chart recorder is planned for installment in the control room after approval. All the equipment was in very good condition. The i?spector reviewed Section 2.3.3 of the UFSAR and noted no deviations in this are Conclusions Based on the direct observations, discussions with personnel, and examination of procedures and records for calibration of equipment, the inspector determined that overall, the licensee's performance of maintaining and calibrating the meteorological monitoring instrumentation was very good. The data were available as required and were easily accessed from severallocations including the control room and the EOF as specified in the UFSA R2 Status of RP&C Facilities and Equipment R2.1 Inocerable Effluent Monitors Technical Specifications 3.3.3.7 and 3.3.3.8 require that certain radiation monitors be operable at all times to monitor the status of liquid and gaseous releases from the site. The licensee declared all technical specification monitors inoperable on February 5,1997 as a result of an NRC inspection which found inadequacies in the calibration program. This issue was described in adverse condition report (ACR) 97-65. The radiation monitors (RM) affected included channels RM-18, RM-22, RM-19, RM-14A and RM-14B. Although inoperable per the technical specifications, the monitors remained functional and available to monitor the effluent pathway Section 01.6 of this report describes NRC review of licensee compensatory actions to comply with the action statement of TS 3.3.3.7 and 3.3.3.8, which included periodic sample and analysis of releases via the associated pathway Several other discrepancies in the effluent monitors were identified during the period as a result of continued testing and reviews by the licensee chemistry personnel, and through quality assurance audits and surveillances. RMS deficiencies included:
the failure of the RM-18 low flow alarm to reset following testing due to lower SW
 
.
.
 
system flows (ACR 97-28); a recurrent failure of the Scanrad Computer to lockup, resulting in involuntary entry into TS 3.3.3.7 and 3.3.3.8 on numerous occasions (ACR 97-47); RMS-22 failed during testing on February 5 (ACR 97-62); area radiation monitors R-20 and R-21 were found inoperable during testing per SUR 5.1-11 (ACR 97-72); the trip setpoint for liquid monitor RM-22 was not conservatively established (ACR 97-79); the lack of traceability of calibration sources used for RMS calibrations (ACR 97-96); multiple failures (channel cross talk) of area radiation monitors during testing per SUR 5.1-11 (ACR 97-133) and, the determination that the sample probe for stack gas monitor RM-14A was not isokinetic and did not meet the design basis to adequately sample the flow stream (ACR 97-110). This latter deficiency was found to be reportable; the licensee notified the NRC per 10 CFR 50.72(b)(1)(ii)(B) on April 4,1997 (ACR 97-169).
 
QAS audit CY-P-97-006 identified an adverse trend with the reliability of the RM The adverse trend was highlighted by many problems (22 issues noted) over the period from October 1996 to February 1997, which continued to occur despite corrective actions. QAS found that although actions were taken to correct immediate problems, the RMS reliability issue has not been addressed. Problems '
       ;
with failed calibrations, failed sensors, system lockups (scanrad computer), set point drift, improper use of procedures, inadequate procedures, no alarm response and annunciation failures have continued with no overall adequate resolution. QAS determined, based on the current conditions of the system, it was not apparent that j the design basis was being met by the system in the current configuration (ACR 97-73).
 
i Licensee plans and schedules to recalibrate the monitors were described in letters to the NRC dated February 24 and March 27,1997. A Special Report per TS 6.9.2 to address the status of RM-14B was submitted to the NRC on February 19,1997 (B16268). The licensee determined that the recent calibration deficiencies occurred as a result of recent changes in responsibility between departments to conduct radiation monitor calibrations. The change occurred in 1996 as the licensee implemented plant modifications to upgrade the radiation monitoring system. These modifications were partially implemented prior to the plant shutdown in July 1996, and then were not completed. To address the calibration deficiencies, the licensee actions included: revising calibration methods and procedures to conduct an insitu primary calibration of the channels; the use of new calibration test stands; the procurement of new sources (Cs-133 and Co-60) to improve the calibration method; and, additional reviews with the RMS vendor to improve the reliability of the Scanrad compute The licensee also committed to evaluate any potential inaccuracies identified by the new calibrations to determine the impact on past operations. During a calibration of channel RM-14B on February 26,1997, the licensee identified that the mid range detector channel was significantly out of tolerance low (ACR 97-101). When tested per SUR 5.2-69, the channel read 2400 cpm, which was below the required acceptance criteria to be within the range of 6777 to 8283 cp .  - . -
.
 
  .
  .
 
U. S. Nucinar R:gulatory Commission
Licensee actions to assess the as found accuracy of the radiation monitor channels relative to periods of past plant operations was in progress at the conclusion of the .
inspection period. As of April 7, the licensee had procured new sources,  l constructed new calibration equipment and revised the procedures to calibrate RM- !
22, RM-14A and RM-18. The licensee planned to complete the new calibrations and restore all channels to an operable status by May 2,1997. The licensee  i assembled a team comprised of l&C, chemistry, engineering personnel to develop l an Action Plan to address all RMS deficiencies in a comprehensive manner. The draft Action Plan was under review at the end of the inspection perio ;
R3 RP&C Procedures and Documentation l
R3,1 Whole Body Countina    )
I
        ^ inspection Scope (83750)
As part of the follow-up to the November 2,1996, fuel transfer canal airborne-radioactivity event, the inspector used an anthropomorphic phantom (Humanoid Systems), to qualitatively check the licensee's whole body counter and dose
% assessment capabilities. The whole body counter, a standup counter (Canberra Model 2250 FASTSCAN) maintained at the Emergency Operations facility, was
: used by the licensee, in addition to other data (e.g., fecal analyses), to quantify the
*
intake of airborne radioactivity following the November 2,1996, event. The phantom's lungs contained known quantities of radioactive material (including the principle gamma emitting radionuclide identified during the November 2,1996
,
,
. event) which the licensee was requested to estimate. The phantom was removed from and placed back into the counter and counted several times to evaluate geometry considerations. Further, the licensee was requested to perform a dose estimats based on inspector-provided radionuclide characteristics (e.g., transport class and particle size). Observations and Findinas The inspector's review indicated the licensee's whole body counter provided a reasonable qualitative estimate of radioactivity contained within the lungs of the phantom. Further, once the deposited activity was estimated, the licensee performed reasonable dose estimates based on assumptions provided by the
CY-97-058/ Attachment 1/Page 4 In the second case, the workers involved worked more than 72 hours in a single week due to inadequate management oversight.
*
inspector, Conclusions .
The whole body counter and procedures used by the licensee to estimate intake of
,
gamma-emitting radionuclides provided a good qualitative estimate of radioactivity contained within the lungs of a test phantom. Further, the dose assessment performed by the licensee was reasonable based on assumptions provide . - .
. _  _  __ . _ .


.
Corrective Steps that have been taken and the results achieved in the first case, the individual scheduling the operator was counseled and all Shift Managers reviewed the requirements for overtime controls for Nuclear Group Personnel, NGP 1.09. The specifics of these events have been given to operations personnel to convey management's expectation regarding the overtime practices and procedures.
.


R3.2 Contamination Controls (URI 97-0_1-11) Inspection Sg_qp_g The inspector reviewed selected aspects of the licensee's contamination control program. The review was in response to the identification by a vendor on February 27,1997, that video equipment, picked up by the vendor at the Haddam Neck facility as clean equipment on February 19,1997, was identified by the vendor to exhibit radioactive contaminatio Observations and Findinas identification of Contaminated Equipment at an Offsite Vendor Facility On February 26,1997, the licensee was informed that low level radioactive contamination had been detected on video cables that had been released from the station on February 19,1997. The contamination was detected by the vendor's personnel who normally perform routine contamination surveys of video equipment that is returned to the vendor's f acility from reactor f aci;ities, interim actions were taken (February 27,1997) to restrict the release of material from the radiological-controlled and protected areas pending evaluation of the event and development and implementation of long term corrective actions. The details of the NRC review of this matter will be included in a future inspection report. As a result of this event i I
Health Physics Management has directed all Health Physics Supervisors review
and other concerns involving radiological controls, t; a NRC issued a Confirmatory Action Letter (CAL No. 1-97-007) on March 4,1997, specifying actions to be taken to effect overall program improvemen Conclusions Contaminated material was detected outside the radiological controlled area including in the possession of a non-licensed individual. The details of the NRC review of this matter will be included in a future inspection report. The effectiveness of contamination controls is an unresolved item. (URI 97-01-11)
  " Overtime Controls for Nuclear Group Personnel", NGP 1.09, thus reiterating Station Management expectations and responsibilities to follow procedural requirements.
R5 Staff Training and Qualification in RP&C (URI 97-01-12) Insoection Scone (83750)
The inspector selectively reviewed the training provided to the licensee's radiation protection staff relative to the programmatic weaknesses identified during the November 2,1996, fuel transfer canal airborne radioactivity event. The inspector also reviewed training of personnel to perform contamination monitorin . . .. .. ---  . --  ._ = . . . .
      .
l t
.
      .
71 Observations and Findinos b-1. November 2,1996, Fuel Transfer Canal Event    i
      '
The licensee provided training to the radiological controls group on the circumstances surrounding the November 2,1996, fuel transfer canal airborne radioactivity event. The NRC inspection report, which detailed the NRC reviews of the event, was provided to the staff including the licensee's internal lessons learned document. Also, the licensee provided training on NRC Information Notice No. 92- -
75, " Unplanned Intakes of Airborne Radioactive Material By individuals at Nuclear ,
Power Plants," dated November 12,1992, which was referenced in the inspection repor !
b 2. Contamination Controls
      !
The inspector reviewed the training and qualifications of personnel who calibrated j and operated the waste sorting table. As discussed above, camera equipment, that i was apparently surveyed for contamination by use of the waste sorting table was ,
identified to be contaminated and located at an offsite vendor facility. The I inspector's review indicated the followin ]


        '
Corrective Steps that have been taken to avoid further violations Station management addressed this issue generically by reiterating the importance of observing overtime limits in an article in the plant daily newsletter, CY TODAY.
- The inspector was not able to identify a formal training program or training records for the individuals who calibrate the waste sorting table or bag i monitor. Calibration procedures were establishe l
- The inspector was not able to identify forrnal training records for several individuals who operated the waste sorting tabl The training and qualification of personnel for calibration and operation of contamination monitoring equipment is an unresolved item. (URI 97-01 12) Conclusions The licensee provided training of the radiological controls group on the  !
November 2,1996, event. An unresolved item was identified in the area of training and qualification of personnel who calibrate and operate contamination monitoring equipmen R6 Radiological Protection and Chemistry (RP&C) Organization and Administration Insoection Scoot The inspector selectively reviewed the licensee's organization and staffing relative to information contained within the Technical Specifications. The inspector also discussed the proposed organization to support decommissioning activitie j I
I i
i I
      ;
        )


      .
Date when full compliance will be achieved
.
_
.
Corrective actions have been completed and full compliance has been achieved.
72 Observations and Findinas The licensee developed a proposed interim and final decommissioning organization and staffing levels. At the time of the inspection, the licensee was continuing to maintain the normal radiological controls organization, but expected to transition to the interim decommissioning organization on or about April 9,1997, and transition to a final decommissioning organization on or about December 31,1997. Staffing levels for these proposed station organizations was expected to decrease from the current 322 personnel to 171 personnel (interim) to approximately 106 personnel (final). The radiation protection organization was expected to decrease from the current 34 person staff level to approximately 26 personnel. The licensee's proposed interim and final decommissioning organization included the establishment of associated responsibilities and authority for assigned positions. The licensee re-posted all positions essentially requiring current staff members to submit applications for decommissioning organization position The licensee was reviewing the proposed organizations and had contacted other stations undergoing decommissioning to attempt to determine appropr!ste organizational design and staffing levels to support decommissioning activities. The licensee's station services director indicated that, based on expected work reductions, the staffing levels in the area of chemistry personnel and radwaste technicians was expected to decreas As part of the proposed organizational structure, the licensee provided for enhanced attention to radiation protection and safety programs, in that a new safety manager position was established. In addition, an enhanced QA organization (to provide /obtain staff experienced in radiological control / decommissioning activities)
was establishe Conclusions The licensee developed proposed organizational structures and staffing levels to support decommissioning activities and was evaluating industry experience in this area. The licensee's initial efforts in establishing a decommissioning organization were considered adequat R6.1 Manaaement Controh Inspection SSone (84570)
The inspectcr reviewed organization changes and the responsibilities relative to oversight of the REMP and MMP, and the Annual Radiological Environmental Operating Report to verify the implementation of Section 6.9.1.6 of the T Observations and Findinas Organization changes affecting RAB occurred in 1996. The RAB was moved from Nuclear Engineering to Nuclear Design Engineering back to Nuclear Engineering for


  . . . . .
Restatement of Violation 1.d Work Control Manual WCM 2.1-2, Trouble Report / Job Scoping, Revision 3, requires in Step 1.2.3 and 1.3.3 that trouble report tags be removed and closed upon development of work packages, or for work that will be completed under blanket authorized work orders. Contrary to the above, on February 18,1997,as documented in adverse condition report ACR 97-90, a review of trouble reports by plant operators identified that (f 275 trouble reports hanging on plant equipment, 113 were associated with troublc reports or AWOs that had been either deleted, completed or canceled.


.
Reason for the Violation This violation was caused by inadequate personnel performance, i.e., failure to follow procedures. The primary reason cited for canceled tags still hanging was that the person canceling the Trouble Report (TR) was not going back out to remove the TR tag as required by WCM 2.1-2, Trouble Report / Job Scoping. Also, some TRs were not removed after work had been completed. A review of the various work control procedures indicated there were no programmatic problems with the tagging process. Rather, the main problem was inattention to detail.
j    73 a 6-month time period. In addition, the group was physically moved from the corporate office, located in Berlin to the Millstone station where they remained for a short period of time, and then were moved to an office in New Britain. The RAB personnel and responsibilities remained the same during these change The Annual Radiological Environmental Operating Report for 1995 and the analytical data from 1996 and 1997 were reviewed. The report included a summary of the j results of the radiological environmental monitoring program from the report period
'
including a summary of the analytical results of the environmental samples and radiation measurements taken from locations specified in the REMODCM. The report also included the land use census and the interlaboratory comparison results, as required by TS. No obvious omissions were noted. Inconsistencies between the REMODCM and the annual reports regarding distances documented in the land use consus, appeared to be administrative errors. No anomalous data were noted in the 1996 or 1997 data. The REMP reports met the TS and REMODCM reporting requirement ,
l Conclusions      !
*, -- Based on the above review, the inspector determined that the licensee implemented good management control and oversight of the REMP and MMP and effectively implemented Section 6.9.1.6 of the TS requirement R7 Quality Assurance in RP&C Activities incoection Scope (83750)
.The inspector selectively reviewed the licensee's quality assurance effort Observations and Findinas To oversee decommissioning activities, the licensee has hired individuals into the quality assurance group with extensive experience in radiological controls programs and decommissioning. This was considered a good initiativ )
      ! in i
_Conclu_tona The licenste enhanced the radiological controls and decommissioning oversight capabilities of the quality assurance group. This was considered a good initiativ d Qua'ity Assurance Audit Proaram Inspection Scope (84750)
The Quality Assurance (OA) audit and surveillance reports were reviewed to determine if the licensee was making an effort to identify deficiencies in the sampling, measurement and quality assurance program .
.
 
74
- QAS Audit Report, (95-4314), " Radiological Effluent Monitoring and Offsite Dose Calculation Manual (REMODCM) - 1995," October 5,1995
- Nuclear Safety and Assessment Audit Report, 96-A10-02,  l October 15-22,1996    l
 
-
Surveillance Report (SIP No. CY-P-97-024) Observation and Findinas The assessment of the REMP as documented in the 1995 QAS audit report, performed by the Quality Assurance Surveillance (QAS) group, was very limite There were no concerns regarding programs strengths or weaknesses. The report
'
combined audits of the effluent and environmental programs from both CY and Millstone thus making the audit report confusing and difficult to understand and interpret. Requirements and deviations from requirements were not clearly stated and issues were vague. No discussions of audit findings were evident. Based on the report, the 1995 audit appeared as an ineffective effor The OAS group have made noticeable irnprovement efforts to better identify deficiencies and assess the strengths and weaknesses of the environmental and- j
, meteorological monitoring programs. The 1996 QAS audit of t i REMP/RETS/ODCM was more critical than the 1995 rudit. The report identified l four ACRs (one REMP related) and several observations. The responses to ACRs - !
were timely and appropriate. The effectiveness of the corrective actions in response to the 1995 audit were reviewed and docemented in the 1996 audit report. All five audits (Seabrook, Millstone 1, 2 and 3, and Connecticut Yankee)
s were then compiled into one audit report. Discussions of audit findings were clearl documented and issues were easy to understand and interpret. Audit findings were appropriate and appeared reasonable.


. The CY Oversight group became effective in January 1997. This group will perform audits and surveillances of CY exclusively. The group recently completed a surveillance specific to implementation of the REMP. The findings of the surveillance were appropriate and appeared reasonable, issues were formally documented and processed. This group is scheduled to perform an audit of the REMP in October 1997, Conclusions
.
.
Based on the 1996 audit and 1997 surveillance, and discussions with the CY oversight group, the inspector concluded that the audits have improved significantly since the 1995 audit and the QA Oversight group will conduct audits and surveillances of the REMP at C .    - ,_ .


.. - - -. - - . -- - - .--
_ . _ _ _ _ _ . _ _ ._. _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _  _ _ _ . _
  .
'
- -
  .      .
  .
  .
 
U. S. Nuclear R:gulatory Commission
872 2 Quality Assuritoce of Analvtical Measurements
_    , Inspection Scope (84750)
The inspector reviewed the quality assurance (QA) and quality control (QC)
programs against Section E.3 of the REMODCM and recommendations of Regulatory Guide 4.15, " Quality Assurance for Radiological Monitoring Programs (Normal Operations) - Effluent Streams and the Environment" to determine whether the licensee had adequate control with respect to sampling, analyzing, and evaluating data for the implementation of the REMP. The following issues were discusse Participation of the Yankee Atomic Environmental Laboratory (YAEL) in the Environmental Protection Agency (EPA) Interlaboratory Comparison (cross-check) program (1995)
      ;
  - ~ Review the cross-check program provided by Analytics to determine Ine effectiveness of the program, Observations and Findinos The performance of the contractor laboratory, Yankee Atomic Environmental Laboratory (YAEL), continued to be excellent. During an inspection at Millstone, the inspector visited the laboratory and assessed the quality assurance program. See-Section R7.2 of the Combined inspection Report Nos. 50-245/96-09, 50-336/96-09, and 50-423/96-09 for details.
 
i The EPA discontinued the interlaboratory comparison program, with the exception
-
of drinking water,in January 1,1995. The licensee was timely in finding a laboratory, Analytics, Inc., to continue the interlaboratory comparison program, as required in the REMODCM. Spike samples, similar to those provided by the EPA cross-check program, were provided to YAEL where the analyses were performe i YAEL sends the analytical results to Analytics for review. Analytics compares the !
ratio of the observed result to the expected result. The inspector reviewed the analytical results of this program and noted the results were within the licensee's l
      '
acceptance criteria. The inspector also reviewed the results of the 1995 EPA cross-check program and noted that most of the results were within the EPA acceptance l
crit::ria. Overall, the licensee's performance in this area was very goo Conclusions Based on the above observations, the inspector determined that the performance of the contract laboratory was excellent and the interlaboratory program was effective.
 
,
  - . , - ,
 
  .  .. __ . ._. . _ _ _ . _ . _- __ . . _ _
.-
  ,
  ,
        {
CY-97-058/Attachm:nt 1/Paga 5    ;
        '
Corrective Steps that have been taken and the results achieved
1, -
:
i   l'
i All 113 TR tags have been removed. Guidance was provided to all Managers / >
76    >
Supervisors on March 4,1997, outlining the appropriate use of the TR system, ,
R8' . Miscellaneous RP&C lasues      1 Bil Decommissionina Proiect Plannina      _ -
including the proper hanging and removal of TR tags. Supervisors are expected to :
4 Insoection Scope
ensure that personnel understand the requirements of WCM 2.1-1, Work Control i
        {
        '
,
,
The inspector selectively reviewed the licensee's decommissioning project plannin ;
Process in the attachment for Work Performance, the Work Completion section delineates one of the responsibilities of the worker as well as the first line supervisor as being to " Verify TR tag removed" and attached to the work order, f
        -
, unless contaminated or lost.
t ,
i Observations and Findinas The licensee established a decommissioning project milestone schedule and flow chart. The document identified expected deliverables as well as task team leader At the time of the inspection, the manual contained approximately 28 separate
'
tasks and schedules. ' The tasks included such items as development of the Post  ,
'-  - Shutdown Decommissioning Activities Report, revision of the UFSAR, and various  .;
accident analyses task i
,
  .
l . Conclusions      ]
 
;r - Th'e licensee established decommissioning project plans and schedules to support  1
  - decommissioning project plannin !
R8.2 Followuo of the November 2.1996. Fuel Transfer Canal Event l'
; Insoection Scope
'
  -The inspector selectively reviewed immediate and interim corrective actions e implemented by _the licensee following unplanned personnel exposure event .  -
. discussed in NRC Inspection Report No. 50-213/96-12, dated December 19,1996.
 
. Observations and Findinos
 
The inspector's review indicated the licensee implemented the following immediate  !
and interim corrective actions for the fuel transfer canal airborne radio::ctivity even I i'
  - The licensee initiated work control measures for work within the radiologi::al controlled area to preclude similar events pending establishment and implementation of appropriate corrective actions to address program weakness identified by the licensee's two root cause evaluations and the findings of the licensee's independent review team. Work within areas designated as high risk areas was to be specifically reviewed ar.d approved by the work services director or the unit director to ensure that there was a  ,
*
need to perform the work and that the work had adequately been planned  !
  .and prepared. Memoranda detailing these controls were issued on
  . November 25,26 and December 12,1996.. The November 25 memorandum
;  detailed controls for backshift and weekend work and the latter two memoranda detailed work controls for high risk area !
    '
  ,_ ,,_ n, .. . .u __- _ . = - _ _ _ - - . -


i Corrective Steps that have been taken to avoid further violations The specifics of the violation have been reinforced to all personnel under the work !
;  control supervisor using additional guidance and department training to clearly ,
.
.
.
convey Station Management expectations regarding work control and tagging ;
practices.


The licensee provided details of the work restriction to the NRC in a letter dated December 9,1996. As of the date of this inspection, no radiologically significant work activities occurred. The licensee did floodup the reactor cavity and off-load the reactor core during the month of November 199 Constant health physics technician coverage was provided for this activity and no radiological concerns were note The licensee issued a Licensee Event Report (LER) No. 50-213/96-030-00 on December 6,1996. The LER discussed, among other matters, the event and provided preliminary dose assessments and a discussion of apparent root i cause for the November 2,1996, event. This LER is close During the week following the event, radiation protection management met with radiological controls department personnel to discuss the event and its apparent root causes and corrective actions. A follow-up meeting was held with radiction protection department personnel once it was determined that an apparent overexposure to airborne radioactivity may have occurred and to discuss findings of initial root cause analyse The unit director issued memoranda (November.4,1996) to station managers and supervisors regarding expectations for notification to the control room and notification of work supervisors of work stoppages and the need for effective pre-job briefing A station work stand down was held on January 28,1997, for purposes of l re-emphasizing the need for industrial and radiation safety. The licensee's work services director also met with station departments in January 1997, to discuss the November 2,1997, event and its root causes, and lessons learne Following the event, radiation protection management took a number of immediate and interim actions to strengthen radiological controls as follow ,
)
 
-
The licensee permanently assigned an additional radiation protection technician (day shift) to the RCA control point to challenge workers entering the RCA as to their purpose for entry and to ensure appropriate RWPs were implemented. (November 6,1996) As discussed above, the licensee prohibited backshif t and weekend work without specific approval Alllicensee and contractor radiation protection technicians were required to read (November 11,1996) NRC Information Notice 90-33, Sources of Unexpected Occupational Radiation Exposures at Spent Fuel Storage Pools, dated May 9,1990, in support of fuel movemen The licensee eliminated use of field counting / checks of air samples as a basis for reduction in worker radiological controls. (January 10,1997)
 
_ ~ . _ . . -  -  _ _. - . . . . . . _ . _ _ . . . . _ _ ._ . _ _ . _
          .
,
..
          :
.
I
 
78
  -- The licensee revised applicable radiatit x permits to ensure compliance -  _
'
'
with Technical Specification High Rac . Area requirements (i.e., provision of job area radiation surveys) or dear ..ed the RWPs to prevent their us l
Date when full compliance will be achieved
,
\
    (December 1,1996)      ,
*
          !
CYAPCO has completed the corrective actions and full compliance has been achieved.
  - The licensee implemented (January 1,1997) zone radiation protection  ,
technician coverage to provide for technician familiarity of work area  :
          !
-
radiological conditions and improved ownership of work within the RCA.


'  - - The licensee developed an action plan to implement corrective actions for the ;
!
findings of the Independent Review Team (IRT).
I Restatement of Violation 1.e.


i          !
Work Control Manual WCM 3.3-1, Technical Specifications Surveillance Tracking, '
;.
        ,
  - .The licensee initiated action to review the ALARA and alpha contamination  i control programs. The licensee established an alpha program review / revision  ,
Revision 2, requires in Step 1.4.2 and 1.4.3 that the station procedures be audited
y    task force. The task force developed a series of recommendations for  .;
revision of the alpha control progra i
          !
  - The licensee assigned (November 26,1996) radiation protection technicians  j to read / review procedure l I
The licensee issued guidance (January 7,1997) regarding inappropriate use
'
  -
  . of " clean" terminology when briefing personne On or about November 22,1996, the licensee required supervisors to approve all RWP The licensee implemented enhanced oversight and control of work and e-  developed a draft conduct of work guidenne based on radiation work risk -
  -
assessment (i.e., establishment of controis based on radiological conditions  ;
;    identified).
 
- The. licensee developed standard pre-job briefings for expected high risk work
;    that may occur (e.g., spent fuel evolutions, diving, resin liner transfers, and filter changes). (January 8,1997)
 
  -The inspector's review indicated the licensee planned radiological controls program
-  upgrades by the end of February 1997. Program upgrades, among others, were  j
          '
identified for the following areas:
          !
.
  - Develop and implement improved procedures for high risk evolutions, i
including performance of representative pre-job surveys.
 
'
  -
Enhance the ALARA program to address both external and internal exposure
*
control !
Require use of respiratory protective equipment in all high risk alpha areas
  ~
h  -
 
until acceptable airborne conditions were determine j
'
'
annually (initiated in January) to assure the technical specification requirements i  are met. Procedure ACP 1.2-6.5A, Station Procedures, Revision 0, requires that
!  technical specifications surveillances be conducted by procedure in the 4  Surveillance Procedure category. Contrary to WCM 3.3-1, as of April 4,1997, the Licensee had not completed the annual audit of procedures per WCM 3.3-1.


i
i
          !
^
, . _ , .  ._m _ , _ . _ _ - _ , -- -
Contrary to ACP 1.2-6.5A, on March 12,1997 the licensee identified that technical
      -_ - -- _ _ _ _ _ _ _ __________i
, specifications surveillances were completed using preventative maintenance
  ;
  (PMP 9.1-31) and environmental services (ESP 14.1-4) procedures.


      .-
Reason for the Violation The requirement to audit the Master List of Technical Specification Surveillances 2  was missed when the former program administrator was transferred to a different
.
,
.       e
location and the work was not reassigned to another individual.


- Upgrade the RWP program to provide for clear descriptions of authorized work and controls, including realistic dose controls with appropriate margins and limit As of March 6,1997, the licensee had revised and updated and issued six procedures to provide enhanced radiological controls including those controlling -
!        !
RWP initiation and risk assessment, health physics job coverage, ALARA reviews, non routine whole body counting, and radiological survey Conclusions The licensee initiated and took a number of corrective actions in response to the November 2,1996, fuel transfer canal event. However, a significant number of program improvements, although not yet due at the time this area was reviewed, have yet to be completed. The licensee placed all significant radiological work on hold pending program improvements as described in the licensee's  4 December 9,1996, letter to the NR l R8.3 (Ocen) URI 96-12-01,02: Exposure Assessment, Dose Calculations  )
i        i i
l Insoection Scoo_g The inspector reviewed the licensee's calculations of deep dose equivalent (DDE),
I ll M w
shallow dose equivalent (maximum extremity and whole body) (SDE), lens dose i equivalent (LDE), committed effective dose equivalent (CEDE), committed dose l equivalent (CDE), and total effective dose equivalent (TEDE) for the two workers j who entered the fuel transfer canal on November 2,1996. The workers unknowingly carried a bag of debris, later measured to indicate between 20 R/hr to 60 R/hr on contact. Also, the workers sustained an intake of airborne radioactivity which included transuranic contaminants. The review was against criteria contained within 10 CFR 20 and guidance contained within apolicable NRC Regulatory Guide Observations and Findinas Dose Attributable to Radiation and Contamination External to the Body The inspector's review indicated that relative to the external exposure of the workers, the licensee made measurements of the bag of debris with TLDs placed at various distances from the bag. The licensee calculated the potential unmonitored deep dose equivalent (DDE), shallow dose equivalent (SDE), and maximum exposure to the extremities, and lens dose equivalent essociated with collecting and handling the bag of debris and walking in the canal. The inspe:: tor noted that the licensee did not initially calculate, for Worker A, the dose attributable to peeling paint chips from the surface of the fuel transfer canal wall and did not calculate, for Individual B, the shallow dose to the skin due to potential contamination of the back of the coveralis when sliding against the wall of the fuel transfer canal. The licensee also did not calculate maximum lens dose equivalent associated with the activity. The
 
  . _ _ _ - __  _ _ _ _ _
.
       '
< .
.


licensee subsequently calculated the exposures for the individuals. The licensee '
. -. --  - .- . - - - -- --- . - .=. _.
provided the following radiation doses (as measured with TLD) and potential unmonitored dosec which were added to the workers' radiation exposure historie j Based on the inspector's review, the measured and estimated external doses were considered reasonable and well within applicable limits,  :


      '
! U. S. Nucl::ar Regulatory Commission *
      ,
,. CY-97-058/ Attachment 1/Paga 6    l
  !
'
.
The requirement to conduct surveillances with procedures in the surveillance procedure category was overlooked. for the preventive maintenance and ,
f
environmental procedures. This error was caused by inattention to detail. l


J
      .
 
_
                . .
 
Table 1 Deep Dose Equivalent (DDE) (millirem)
Worker DDE by TLD for    Additional DDE by Total DDE for Total DDE for 10 CFR 20.1201 11/2 event    evaluation for 11/2 event 11/2 Event  1996  annual limit A  233    152  385  396  5,000 (DDE + CEDE)
8  157      56  213  473  5,000 (DDE + CEDE)
Table 2 Shallow Dose Equivalent (SDE), Whole Body (WB) (millirem)
Worker SDE, WB by TLD for    Additional SDE, WB by Total SDE, WB Total SDE, WB  10 CFR  ,
11/2 event    evaluation for 11/2 event for 11/2 Event  for 1996  20.1201 annual limit
                  ~
A  233    154  387  398  50,000 B    157    56  213  473  50,000 Table 3 Shallow Dose Equivalent (SDE), Maximum Extremity (ME) (millirem)
Worker SDE, ME by TLD for    Additional SDE, ME by Total SDE, ME Total SDE, ME 10 CFR 20.1201 11/2 event    evaluation for 11/2 event for 11/2 Event  for 1996  annual limit A    233    920  1,153  1,164  50,000 B    157    284  441  701  50,000
--- . . - - . . _ _ _ - _ _ _ _ - _ - _ - _ - - _ _ _ _ _ _ _ _ - . - _ _ _ - - - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - .- .-.  . - . - _ _ - , - . . . . . . - -. - _
                . . _ - , .
 
_. -. . - _ - _ _ _ _ -
 
Note: The licensee calculated an additional 12 millirem to the face of Worker A (due to working in close proximity to the bag of debris) that is to be added to the lens of the eye dose and shallow dose to the whole body and extremity. Consequently the totallens of the eye dose to Worker A was 397 millirem for the event,399 millirem to the shallow dose equivalent to the whole body, and 1,165 shallow dose equivalent maximum extremit Dose Attributable to Contamination Internal to the Body The inspector reviewed the licensee's calculation of committed effective dose equivalent (CEDE), committed dose equivalent (CDE), and total effective dose equivalent (TEDE) for the two workers who entered the fuel transfer canal on November 2,1996. The licensee evaluated the dose to the workers using (among other data), whole body counts, air sample analysis data, fecal sample analysis results, and sample analysis results for samples of material collected within the fuel transfer canal. The following doses, attributable to intake of airborne radioactivity, were calculated by the license Table 4 Committed Effective Dose Equivalents (CEDE)
1996 TEDE    !
I Worker CEDE Previous for 11/2 1996 (DDE, from Table 1, 10 CFR 20.1201
,
,
event CEDE Column 5, plus CEDE for Annual limit (millirem) (millirem) 11/2 event (millirem) (millirem) *
Corrective Steps that have been taken and the results achieved The Unit Director has assigned an individual with the responsibility for
A 250 none  646  5,000  ;
'
8 440 none  913  5,000  j
; implementing procedure WCM 3.3-1.
* Note: The annual limit of 5,000 millirem is for the Total Effective Dose equivalent (TEDE) which is the sum of the deep dose equivalent (DDE) and the committed  i effective dose equivalent (CEDE).


Based on the above analysis, neither worker received radiation exposures in excess of 10 CFR 20.1201 limit Regarding committed dose equivalent due to intake of airborne radioactive material (e.g., dose to the bone surface attributable to transuranics), the inspector noted that the licensee's contractor calculated the doses by use of a combination of an inhalation and an ingestion model and assumed an intake of 1 micron diameter particles. The following doses were calculate .
Corrective Steps that will be taken to avoid further violations  ,
.
) The individual assigned will ensure implementation of WCM 3.3-1. The applicable
;
preventive maintenance and environmental services procedures will be converted '
; to surveillance procedures.


Table 5 Total Organ Dose Equivalents (TODE)
Date when full compliance will be achieved    l i
Worker CDE for 11/2  TODE  10 CFR 20.1201 event (CDE + DDE from Table Annual Limit (millirem) 1 column 5)  (millirem)
The individual assigned will complete implementation of WCM 3.3-1 and we will j convert the applicable procedures to surveillance procedures by September 1,  ,
A  3,000  3,396  50,000 B  5,400  5,873  50,000 j Based on the above analysis, the licensee concluried that neither worker sustained a TODE in excess of 10 CFR 20.1201 limit The individual worker's dose records (NRC Form 5) were updated to reflect the measured and calculated exposure At the conclusion of the inspection, the inspector had not completed an independent analysis of the licensee's dose calculation. This unresolved item
  '
continues to remain open pending additional NRC review of the licensee's CDE dose -
assessmen R8.4 Housekeepina The inspector noted that overall housekeeping was goo R8,5 Confirmatory Action Letter    i l
       '
       '
Since November 2,1996, the licensee experienced several radiological events / problems attributable to weaknesses in managing and controlling radiological work at the facility. These included the November 2,1996, reactor cavity airborne radioactivity event (Reference NRC Inspection Report No. 50-213/96-12), the programmatic deficiencies associated with calibration of effluent monitoring systems (Reference NRC Inspection Report No. 50-213/97-02), and the problems associated with release of contaminated material to an unrestricted area. As a result of these issues, the NRC issued Confirmatory Action Letter (CAL 1-97-007)
to confirm the licensee's actions and commitments to identify and effectively resolve weaknesses and deficiencies in the implementation of the radiological controls progra B3J UFSAR A recent discovery of a licensee operating their facility in a manner contrary to the Updated Final Safety Analysis Report (UFSAR) description highlighted the need from a special focused review that compares plant practices, procedures and/or parameters to the UFSAR description. While performing the inspection discussed in this report, the inspectors reviewed the applicable portions of the UFSAR that


    .- .- .
1997.
!
l
*
i i


related to the areas inspected. The inspectors verified that the UFSAR wording was consistent with the observed practices and procedures and/or parameters.
a      i
 
: Restatement of Violation 1.f.     ,
.
Security procedure SEC 1.3-8, Package and Material Control, Revision 26, requires in Step 6.2.1.b, that all packages be searched prior to entry into the i protected area. Contrary to the above, on March 13, 1997, six boxes were  l brought into the protected area without receiving the required security searches. j
P2 Status of EP Facilities, Equipment, and Resources f)2d Emeroency Plan Staffino
      !
        . Insoection Scope (71750)      !
Reascn for the Violation    j
        )
      !
The purpose of this inspection was to review licensee changes to the site emergency response organization, and the licensee response to deficiencies in  i meeting plant commitment ! Observations and Findinos      i
The reason for the violation involved poor personnel performance. The  ;
        ;
stockhandler failed to follow procedure, SEC 1.3-8, step 6.2.1b. i Corrective Steps that have been taken and the results achieved The individual was suspended pending investigation. Upon completion of the investigation, the individual's employment was terminated on March 13,1997.
Several staffing or training discrepancies in Emergency Preparedness area were noted during the inspection period. The deficiencies included: the failure to  '
complete timely remedial training of an individual in the site emergency response  !
organization (SERO) who had failed a requalification exam (ACR 96-1371); the poor  ,
4 SERO response to a quarterly radiopager test conducted on February 5 in which 4  !
on-site responders and 6 off-site responders did not callin (ACR 97-61); the poor SERO response to a followup radiopager test conducted on February 6 in which 4 on-site responders and 9 off-site responders did not callin (ACR 97-67); the  !'
inadvertent termination of site access of an employee who transferred to Millstone, which might have delayed the response to Haddam Neck to perform a SERO duty (ACR 97-88); the poor SERO response to a quarterly radiopager test conducted on February 26 in which 6 on-site responders did not callin (ACR 97-100); and, the  ;
findings by QAS regarding inadequate emergency plan training in the development of table-top exercises (ACR 97-122). i Most problems appeared to be due to occur during the period of transition from the  i full plant staff in place for power operations, and the reduced staff designated to remain for plant decommissioning. The licensee evaluated each discrepancy as it occurred and found that none would have precluded implementation of the emergency plan. Appiupriate corrective actions were taken for each discrepanc The licensee continued to review the causes for the poor responses during radiopager drills, which were attributed to equipment and personnel failures. NRC and licensee review of this area were in progress at the end of the inspection perio The licensee revised the emergency plan procedures during this period to delete the use of a Duty Officer at the station during the back shift periods (reference memorandum CY-GHB-97-031). The Duty Officer position had been established during plant operations to provide additional resources on-site to assist the Shift Manager classify events and implement actions under the emergency plan. This resource was found to be necessary to lessen the burden on the shift manager as he monitored th plant status and directed crew actions to mitigate the operational event. The Duty Officer resource was found to be no longer needed based on the
      .
      . . _ _ _ _ - _ - - _ _ _ - _ _
 
      - , - .
.
.


status of the facility and the decision to cease operations. The shift manager assumed responsibility event classification and notification under the new procedures, until the SERO was fully staffed. The change was evaluated per 10 '
Corrective Steps that have been taken to avoid further violations  ,
CFR 50.54 (q) was found acceptable since it did not decrease the effectiveness of the emergency plan. The change became effective on April 4,1997. The inspector reviewed the licensee's actions and identified no deficiencies relative to maintaining an effective emergency response capability relative to the duty officer position, Conclusions Licensee actions in response to deficiencies in meeting emergency plan  !
l The remaining stockhandlers were interviewed and the incident was determined to be an isolated case. Management expectations were reaffirmed with the remaining stockhandlers.
commitments were acceptable. The change to the SERO by deleting the use of the Duty Officer was acceptabl l S1 Conduct of Security and Safeguards Activities    j l
S1,1 Fitness for Duty      I The inspector reviewed licensee actions on April 2 in response to a fitness for duty
- event that was detected while conducting pre-access testing for a new employee . ..


      '
l i
(reference Security memorandum CY-3550). The event did not involve supervisory personnel. Licensee actions were appropriate to identify and investigate a potential problem, and to take corrective action. No inadequacies were identifie :
l
l
_S1,2 Guard Inattentive To Duty The inspector reviewed licensee actions to the discovery on April 3,1997 of a guard who was inattentive to his duties. The individual had not worked overtime in excess of the administrative limits. The guard was stationed within a vital area as a compensatory measure for other security equipment that was inoperable. The licensee took appropriate actions to immediately assess the area, re-establish -
      ~
appropriate security controls, and take corrective actions. The incident occurred within about two hours from when the guard started his daily shift. No inadequacies were identifie S1.3 Resoonse to Potential Threat The inspector reviewed licensee actions on April 2 in response to a telephone call from an individual located off site. Licensee actions were appropriate to assess the caller as a potential threat to the plant (the threat potential was not deemed
      ~
. credible) and to coordinate with offsite and local authorities in regards to the incident. No inadequacies were identifie S1,4 Eailure to Search Packaaes NIO 97-01 -02.f)
      ;
The inspector reviewed licensee actions on March 13 - 15 in response to the discovery that several boxes containing personal office supplies were brought into the protected area without receiving the required security search (ACR 97-132).


. - - .
_._..___.m.._- ....._.


,.  -  . .
'
,
  ..-- - -        ;
  .
U. S. Nuclear R::gulatory Commission  -
 
        !
The event occurred when a stockhandler, trained and assigned responsibility for searching materials that enter the protected area by the warehouse, failed to conduct that task. Following the discovery of the failure to follow security procedures, the licensee follow up actions were good to investigate the incident, search the site to assure the vital and security areas were secure, and to take corrective actions. The licensee terminated the individual's site access on March 13 pending the completion of a review of the event. The licensee subsequently terminated the individual for integrity issue Although licensee response actions to the event were good, this event in another example of a broader NRC concern in the failure by plant personnel to follow procedures. The f ailure to search packages prior to entry into the plant protected area was contrary to security procedure SEC 1.3-8, Step 6.2.1.b, which requires that all packages be searched prior to entry into the protected area by either x-ray . ;
CY-97-058/ Attachment 1/Page 7 -      l
      '
;: ,         l r l  Date when full compliance will be achieved    l
examination or visual search. This was a failure to follow procedure SEC 1.3-8, and the sixth of six examples of a violation of Technical Specification 6.8.1 (VIO 97-01- ;
        :
02.f).     l I
b Corrective actions were completed with the completion of interviews and  !
V. MANAGEMENT MEETINGS X1 Exit Meeting Summary The inspectors presented the inspection results to members of licensee management at the conclusion of the inspection on April 7,1997. The licensee acknowledged the findings presented. No proprietary information was examined during the inspectio In addition to the final exit surnmary, management briefings were conducted during the
reinforcement -of management expectations and full compliance has been   !
  .. inspection period as NRC reviews were completed. . Pre exit briefings were also conducted ,
! echieved.       J i
on the following dates:
:
Inspection  Reporting  Area Dates  Inspector  Inspected January 1317  Nimitz  Radiological Controls February 7-11  Jang/Echert  Effluent Monitoring March 17-21  Peluso  Environmental Monitoring
!- Restatement of Violation 2      !
 
i        i'
    . . - . - -
'
   ,
Technical Specifications 4.7.7 requires that reactor coolant system chemistry be  !
  }..
        '
87-PARTIAL LIST OF PERSONS CONTACTED Licensee -
} sampled every 72 hours to assure that chloride and fluoride concentrations remain
C. Bellamy, Chief of Security-
. G. Bouchard, Work Services Director J. Bourasse,-Quality Assurance Supervisor T. Cleary, Senior Licensing Engineer  '
_ R. Crandall, Supervisor-Radiological Engineering R. Gault, Radiation protection Supervisor
  , J. Goergen, Assistant Health Physics Manager
   .
   .
G. Goncarovs, Chemistry Manager
;  below the limits specified in TS 3.4.7. The mode of applicability for the ;
~ W. Eakin, Senior Engineer, Radiological Engineering D. Erickson, Acting Health Physics Manager
i specification 3.7.7 is "at all times". l
  ,
.         (
  .J._ La Platney, Unit Director T. Mc Cance, Nuclear Licensing
' J. Pandolfo, Security Manager R. Sachatello, Radiation Protection Manager J. Stanford, Operations Manager M. Sweeney, Radiation Protection Services Supervisor G. Waig, Maintenance Manager
.
G. van Noordennen, Licensing Manager J. Warnock, Quality Assurance Manager G. Wilson, Public Information NRC E. Conner, Haddam Neck Project Manager R. Nimitz, Senior Radiation Specialist L. Peluso, Radiation Physicist
      ,
t


l s
Contrary to the above, no RCS chemistry samples were taken from November 15,  l 1996 through February 6,1997.      !
t (.
        >
      .
:i        :
m , . . , v--- - -~
1-
'
Reason for the Violation      !
        !
$  The. apparent cause for this violation was poor judgment by Chemistry  )
j  Management. With the core offloaded and no flow through the reactor coolant
:  system, it was deemed acceptable to secure sampling on a system which was out
'        ;
of service.       '
.
i        i l Corrective Steps that have been taken and the results achieved  l l        l >


     . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ -
l  On February 6,1997, purification flow was re-established and the Chemistry  ;
         .
j  Department commenced sampling the reactor coolant system in accordance with
        ,
     ,
  .
j j  the Technical Specification frequency and specified parameters, at the most  :
;  representative location for the plant configuration. Chemistry sampling of the RCS  l
!_ has indicated that Technical Specifications chemistry limits were never  j l  challenged.      j i        t
'         >
j  Corrective Steps that have been taken to avoid further violations  l
!        . !
;
The Chemistry Department has and will continue to maintain reactor coolant  l
;  chemistry sampling until a revision to the Technical Specifications has been  l
'
issued. Defueled Technical Specifications to remove the sampling requirements  l, l- were submitted for NRC approval, on May 30,1997. l t      ~
i
i
        '
[  Date when full compilance will be achieved CYAPCO has completed the corrective actions and full compliance has been  ,
j  achieved.      .i


INSPECTION PROCEDURES USED    I IP 40500: Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing Problems IP 62703: Maintenance Observation IP 64704: Fire Protection Program IP 71707: Plant Operations IP 73051: Inservice inspection - Review of Program IP 73753: Inservice Inspection
. IP 83729: Occupational Exposure During Extended Outages IP 83750: Occupational Exposure
, IP 92700: Onsite Followup of Written Reports of Nonroutine Events at Power Reactor l  Facilities IP 92902: Followup - Engineering IP 92903: Followup - Maintenance IP 93702: Prompt Onsite Response to Events at Operating Power Reactors ITEMS OPEN, CLOSED, AND DISCUSSED Open 97 01-01 URI Control of systems in Defueled Mode 97-01-02 VIO Failure to Follow Procedures (Multiple)
97-01-03 URI Inaccurate Operator Training Records 97-01-04 URI Deficiencies in Material Conditions 97-01-05 VIO Failure to Complete TS Surveillance 97-01-06 VIO Failure to Correct Adverse Conditions 97 01 07 URI Actions to Address SW Waterhammer 97 01-08 URI Actions to Address SW Corrosion 97-01-09 URI Actions to Address Design Basis Discrepancies 97-01-10 IFl Compliance with 40 CFR 190 not verified 97-01-11 URI Review Release of Contaminated Material 97-01-12 URI Training to Operate Waste Sorting Table Closed 94-21-01' VIO Inadvertent Boron Dilution 94-27-01 URI Loss of Electrical Separation 94-011-00 LER. Unplanned Loss of Gpent Fuel Cooling 94-015-01 LER Main Steam Valves Exceed Lift Setpoints 96-08 01 IFl RHR Calibrations and Leakage 95-023-00 LER Failure to Prepare Special Report 96-201-10 URI Alternate Auxiliary Feedwater Sources 96 015-00 LER Containment Air Monitor Trip Valve 196-01-01 IFl Cable Vault Materials Condition 96-04-02 DEV Heavy Load Program Commitments 96-08-04 ' ICI Auxiliary Feed Water Overspeed Trip 96-08-05 IFl Steam Generator Hold Down Bolts
_ _ _ _ _ _ _ _ _
.- -
  - _ . _ . -- - . _- -..-.- . . ... . . . _ - ... . . . . - _ -
*
I
          .
          'l
*
*
l
    .,
  -      ;
  ,
;  96-08-06 lFl Observations of Procedural Quality
:  94-27-04 URI Surveillance Frequency Exceeded    ,
i  94-09-03 IFl HPSI Relief Valve Setpoint Drift    l
          '
96-04-03 VIO Inadequate Safety Evaluation 96-04-04 IFl Heavy Load Controls 4  96-201-12 URI Analysis Supporting LPSI Pump Shutdown 96-201-31 URI RWST 'astrument Calibrations 1  96-030-00 LER Woiker Contamination in Reactor Cavity I  Discussed
:  96-02-03 URI Control Room Habitability    l j  96-06-06 URI Battery Oscillations and Ground
'
96-08-15 URI Start-up issues (7/24/95 NRC Lauer)
;  96-12-01 URI Exposure Assessment
;  96-12-02 URI Review of dose calculations for two workers who entered the fuel transfer canal on November 2,-199 t
$
ll
!          i i          l l          I
!          l
          ;
q
!
i
i
        '
,        j
:        i
        '
.
-.s, y.. ., __...m., .- _ .-. .- . . _,-- . _ , -  - - , , - ,


i f
-
r
-
          ,
. , .
$
U. S. Nucl:ar Regulatory Commission
b
. CY-97-058/ Attachment 1/Page 8 Restatement of Violation 3 10 CFR 50, Appendix B, Criterion XVI, " Corrective Actions" requires that measures be established to assure that conditions adverse to quality are promptly identified and corrected. In the case of significant conditions adverse to quality, the measures shall assure that the cause of the condition is cetermined and corrective action is taken to preclude repetition.


i
"
          :
Contrary to the above, the licensee did not correct the recurrence of significant conditions adverse to quality, in that there have been multiple examples of a failure to adequately implement the technical specification (TS) operational surveillance program. Past deficiencies related to missed or late TS surveillance were the subject of inspection item URI 94-27-01, and licensee event reports 95-12,96-17 and 96-22. The licensee identified the following failures to complete technical specification surveillances in a timely manner: February 6,1997, the failure to test reactor coolant system chemistry per TS 4.4.7: on February 11, the failure to test the main station battery per TS 4.8.2.2 and 4.8.2.1.c: on February 12, the failure to test the service water pumps per TS 4.7.b.2; and on March 21, the failure to verify the positions of safety related valves per TS 4.5.2.c.
l
          !
-- - . - , , ,- , . - - - - . - - -- - __ ,--  .,-~
        -
          ..


_ _ . _ . _ _ _
Reason for the Violation Corrective action has been ineffective in preventing further missed or late surveillances required by Technical Specifications. Lack of accountability and low management standards contributed to a high tolerance for ineffective actions to correct previous problems.
,
u
,
90  ,
I LIST OF ACRONYMS USED ACM Administrative Control Manual  :
      '
ACP Administrative Control Procedure ACR Adverse Condition Report  ;
ADM Administrative Procedure AFW Auxiliary Feedwater AITTS Action item Trending & Tracking System ALARA As Low As is Reasonably Achievable  ,
,
ANN Annunciator Response Procedure  i ANSI American National Standards institute AOP Abnorma! Operating Procedure  l
      '
AOV Air Operated Valve ASME American Society of Mechanical Engineers CAL Confirmatory Action Letter CAR Containment Air Recirculation CBT Computer Based Training CEDE Committed Effective Dose Equivalent  i CFR Code of Federal Regulations CHDP Chemistry Department CHM Chemistry CMP Corrective Maintenance Procedure CMP Corrective Management Plan  i CO Control Operator CYAPCo. Connecticut Yankee Atomic Power Company CYDE CY Design Engineering DCR Design Change Request DCY Design Change Yankee DDE Deep Dose Equivalent DEV Deviation DNO Do Not Use ECCS Emergency Core Cooling System EDAN Environmental Data Acquisition Network EDG Emergency Diesel Generator  i ENG Engineering Procedure    l EOP Emergency Operating Procedure ESP Environmental Services Procedures EWR Engineering Work Request F Fahrenheit    ;
FSAR Final Safety Analysis Report  !
GL Generic Letter HP Health Physics HPSI High Pressure Safety injection IFl Inspection Followup item IR inspection Report IRT Independent Review Team ISAP Integrated Safety Assessment Program  i KPl Key Performance Indicators
    . _ - _ _


. . 4 -
Corrective Steps that have been taDn and the results achieved The Unit Director has met with all responsible managers to reiterate expectations.
  . . . .. . _ _ . _ . _ . _ . . . . _ _ __ .. __
#      1
      >
.
.      !
.-      ;
91    !
        ,
LDB Licensing Design Basis LDE- Lens Dose Equivalent '    i
*. LER Licensee Event Report    l L -LLD Low Level Dose    !
LLRT' Local Leak Rate Testing    .!
LNP Loss of Normal Power Event    !
LPSI- Low Pressure Safety injection Licensed Operator initial Training l
LOIT      !
LOR Licensed Operator Requalification Training LOUT- . Licensed Operator Upgrade Training-LPE = Liquid Penetrant Examination    !
MCC Motor Control Center    i MI Microbiologically influenced Corrosion  !
MM Meteorological Monitoring Program    j MRFF Maintenance Rule Function Failures    i NOP Normal Operating Procedure    l NGP- Nuclear Generation Procedure NOP Normal Operating Procedure    !
NOV Notice of Violation
. NPDES Nuclear Pollution Discharge Elimination. System -
NPSH' Net Positive Suction Head NRC Nuclear Regulatory Commission NRR - Nuclear Reactor Regulation    l NSO Nuclear Site Operator    i NS&O- Nuclear Safety and Oversight Organization  !
NUS Nuclear Utilities Service    !
ODCM Offsite Dose Calculation Manual    l ODM - Operations Department Memorandum    j OJT On the Job Training    ;
PAB Primary Auxiliary Building    !
PC Personal Computer    !
PDR Public Document Room    l PIR Plant inspection Report    !
PMMS Production Maintenance Management System  !
PMP Preventive Maintenance Procedure    l PORC Plant Operations Review Committee    {
POSL Production Operations Services Laboratory
      '
QA Quality Assurance QAS Quality Assurance Surveillance GC Quality Control    i RAB Radiological Assessment Branch    l RCA Radiological Controlled Area    :
RCS: Reactor Coolant System    !
REMODCM Radiological Effluent Monitoring and Offsite Dose Calculation Manual  !
REMP Radiological Environmental Monitoring Program  i REOR Radiological Environmental Operating Report  l
      '
RHR Residual Heat Removal
:RFO Refueling Outage
      .
  -
I
      ,
i
      -- _ _ ?


.. . . _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ .. .__ . .. . ._
Each manager reviewed the process being used to track surveillances in their
      :
- ??      ;
      '
.
      !
'
'
department and verified the process was acceptable.


l RM Radiation Monitors    l RP&C Radiological Protection & Chemistry  {
Corrective Steps that have been taken to avoid further violations The responsible managers have a clear understanding of what is expected and that they are responsible for this important program.
RWST Refueling Water Storage Tank  i
; SDE Shallow Dose Equivalent    -{
J SERO Site Emergency Response Organization  ,
SFB Spent Fuel Building    !
      '
SFP Spent Fuel Pool SFPCS Spent Fuel Pool Cooling System  ;
SM Shift Manager    j
,
'
SUR Surveillance Procedure    i SW Service Water TEDE Total Effective Dose Equivalent TLD Thermoluminescent Dosimeter  7 TMI Three Mile Island    j TPC Temporary Procedure Change l
TRM Technical Requirement Manual  ;
<
TS Technical Specification    i TSS Technical Specification Surveillance  !
UFSAR Updated Final Safety Analysis Report  l-UFSARCR Updated Final Safety Analysis Report Change Request -  .{
URI Unresolved item    !
UT Ultrasonic inspection    .
VIO Violation YAEL Yankee Atomic Environmental Laboratory YTD Year to Date    ;
-
WCM Work Control Manual
"
WL With Lockout


      '
Date when full compliance will be achieved Corrective actions have been completed and full compliance has been achieved.
t
:
.
    .
L
e
- . .  .  - - -
 
r
$'
 
93 Attachment 1 List of Procedures Reviewed The following procedures were reviewed during this inspection as part of the review of procedure quality and adequacy for plant shutdown operatio Operating Procedures NOP 2.6-1, Operation of the Control Air System, Revision 11 NOP 2.20-4, Hypochlorite System Operation, Revision 15 (TPC 97-10)
NOP 2.10-1, Spent Fuel Pit Cooling system Operation, Revision 14 NOP 2.9-3, Refueling Cavity Filling, Revision 24 (TPC 97-16)
ANN 4.7-23A, Spent Fuel Pit High Level, Revision 6 ANN 4.7-238, Spent Fuel Pit Low Level, Revision 6 ANN 4.7-14, Spent Fuel Pit High Temperature, Revision 6 NOP 2.24-1, Service Water System Startup, Revision 19 (TPC 96-787)
NOP 2.24-2, Service Water System Shutdown, Revision 14 (TPC 96-143)
NOP 2.24 3, Filtered SW System and Adams Filter Operation, Revision 17 (TPC 97-60)
NOP 2.15 3, Spent Fuel Building Ventilation Operation, Revision 14 Maintenance Procedures PMP 9.2-20, Calibration of IST Gages, Revision 12 PMP 9.1-36, Service Water Pump Strainer Operation, Revision 9 PMP 9.1-31, Diesel In-Leakage and Fuel Oil Transfer Pump Availability, Revision 11 PMP 9.1-27A, Electric Fire Pump (P-4-1 A) Test, Revision 10 Surveillance Procedures SUR 5.1-126, All Modes Locked Valve Checklist, Revision 24 (TPC 96-608)
SUR 5.1-104C, Boric Acid Flowpath Heat Trace Operabi!ity Test, Revision 1 ENG 1.7-114, inservice Test of Emergency Diesel Generator Heat Exchangers, Revision 9 SUR 5.1-153B, AC and DC Distribution Normal Configuration (Modes 5 and 6), Revision 5 SUR 5.7-148A, inservice Test of A, B, C, D Service Water Pumps, Revision 10 SUR 5.1-178, Emergency Diesel Generator EG-2B Manual Start and Loading, Revision 16 ENG 1.7-131, CY MIC Prevention, Monitoring, and Mitigation Program, Revision 3 ENG 1.7 65, Hydrostatic Pressure Tests, Revision 8 ST 11.7 201, Functional Testing of SW CV-963, Revision O ST 11.7-203, SFP Heat Exchanger Temporary SW Supply Flow Test, Revision O SUR 5.7-217, inservice Testing of SW Supply to SFP Cooling Check Valve, Revision 0,1 Administrative Procedures ACP 1.2-6.5A, Station Procedures, Revision 0,1
}}
}}

Revision as of 04:41, 25 October 2020

Ack Receipt of 970606 Ltr Informing NRC of Steps Taken to Correct Violations Noted in Insp Rept 50-213/97-01
ML20151L907
Person / Time
Site: Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date: 07/31/1997
From: Conte R
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To: Feigenbaum T, Mellor R
CONNECTICUT YANKEE ATOMIC POWER CO., NORTHEAST UTILITIES SERVICE CO.
References
50-213-97-01, 50-213-97-1, NUDOCS 9708080013
Download: ML20151L907 (2)
v • d • e


Text

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

p..

.

-.

July 31, 1997 Mr. Ted . Executive Vice President and Chief Nuclear Officer Northeast Utilities Service Company c/o R. A. Mellor,~ Director Site Operations and Decommissioning Connecticut Yankee Atomic Power Company 362 injun Hollow Road East Hampton, CT 06424-3099 SUBJECT: INSPECTION REPORT NO. 50-213/97-01

Dear Mr. Feigenbaum:

This letter refers to your June 6,1997 correspondence, in response to our May 8,1997 letter.

Thank you forinforming us of the corrective and preventive actions documented in your letter.

These actions will be examined during a future inspection of your licensed program.

Your cooperation with us is appreciated. .

!

Sincerely, Original Signed By:

Richard J. Conte, Chief Projects Branch 8 Division of Reactor Projects Docket No. 50 213 ')

i cc: w/o cv of Licensee's Response Letter {

B. D. Kenyon, President - Nuclear Group D. M. Goebel Vice President - Nuclear Oversight 4 J. K. Thayer, Vice President - Recovery Officer, Nuclear Engineering and Support F. C. Rothen, Vice President - Work Services

-

'

' R. Johannes, Director - Nuclear Training

.

L. M. Cuoco, Senior Nuclear Counsel G.= P. van Noordennen, Manager, Nuclear Licensing H. F. Haynes, Director - Nuclear Training j J. F. Smith, Manager, Operator Training  !

l$.dl

'

9709000013 970731 PDR ADOCK 05000213 a aa =

qui.sppui.ssu..

,

~ . - - , . . . .m.-, ,- - , -.. , , - - - _ . - -

,_,l

.

>

.

Mr. Ted cc: w/cv of Licensee's Resoonse Letter R. Bassilakis, Citizens Awareness Network J. M. Block, Attorney for CAN J. P. Brooks, CT Attorney Generals Office M. DeBold, Town of Haddam W. Meinert, Nuclear Engineer State of Connecticut SLO Distribution w/cv of Licensee Response Letter Region i Docket Room (with concurrences)

Nuclear Safety information Center (NSIC)

PUBLIC NRC Resident inspector R. Conte, DRP M. Conner, DRP C. O'Daniell, DRP W. Dean, OEDO S. Weiss, NRR, DRPM, PDND M. Fairtile, PM, NRR M. Callahan, OCA W. Travers, SPO R. Correia, NRR F. Talbot, NRR D. Screnci, PAO, ORA DOCDESK Inspection Program Branch, NRR (IPAS)

DOCUMENT NAME: G:\ BRANCH 7\REPLYLTR\HN97-01.RPY Ts receive a copy of this document, indicate in the boa: "C" = Copy without attachment / enclosure *E" = Copy with attachment / enclosure "N" - No copy OFFICE Rl/DRP / l NAME RConte DATE 7 /Jy97 / /97 OFFICIAL RECORD COPY

.

...

.

.

CONNECTICUT YANKEE ATOMIC POWER COMPANY HADDAM NECK PLANT 362 INJUN HOLLOW ROAD . EAST HAMPTON, CT 06424-3099 June 6,1997 Docket No. 50-213 CY-97-058 Re: 10CFR2.201 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Haddam Neck Plant Reply to Notice of Violation NRC Intearated Inspection Report No. 50-213/97-01 In a letter dated May 8,1997,N the NRC transmitted to Connecticut Yankee Atomic Power Company (CYAPCO) an inspection report covering the period from January 6, 1997 - April 7,1997, documenting three (3) Severity Level IV violations for the Haddam Neck Plant (HNP). As discussed in the report, the NRC Staff cited CYAPCO for violation of certain NRC requirements. The violations identified dealt with failure to follow procedures, failure to complete technical specification surveillances, and failure to complete effective corrective actions.

Pursuant to 10CFR2.201, Attachment 1 provides CYAPCO's response to the individual violations. The NRC Inspection Report 97-01 also requested information on how CYAPCO is addressing the issues of performance standards and procedure adherence at the HNP. CYAPCO recognizes that human performance issues and thus, performance standards, will play a key role in the safe and successful storage of spent

-

nuclear fuel and decommissioning at the HNP. The management team has initiated a concerted effort to raise the performance standards for all personnel on site and establish high standards for the decommissioning process. Included in this effort is reinforcing high standards for the core values of honesty, integrity and commitment to the job, including program and procedure ownership and adherence, safe working practices and a questioning attitude. These core values have been established in the management and supervisory team through group meetings and discussions. Group meetings to discuss these core values are also being conducted between senior site management and non-management / supervisory personnel.

(1) NRC letter from J. F. Rogge to T. C. Feigenbaum, "NRC Integrated inspection Report No. 50-213/97-01" dated May 8,1997.

l

'

_

D4Mr4J2/LL app - .

. _ _ _ .

- t

.

-

,

U. S. Nucisar R::gulatory Commission

,

CY-97-058/Paga 2 ,

-

The management team has worked with their individual department staff to establish department level standards that reflect these higher core values. All departments have

, completed draft standards and expectation documents that have been reviewed by l senior site management to assure consistency with the core values. Each of the :

department standards and expectations addresses the expectation of adherence to site !'

procedures. Site-wide and department level standards and expectations will be

published by June 30,1997. This process of ensuring consideration of the input from all site staff will help to ensure site-wide ownership of these core values.

,

Other actions that have been taken to improve performance standards include holding

'

individuals accountable for unacceptable performance through appropriate counseling i and/or disciplinary actions. Additionally, a mission statement has been developed and l Issued reflecting the high standards necessary for the safe storage of spent nuclear i fuel and conducting the decommissioning process. The work observation program at the HNP will be upgraded by June 30,1997 to provide increased management and l

supervisory observations of work activities and opportunities for coaching and positive

< reinforcement The improved Corrective Action Program has implemented a routine i trending report that provides management with timely data regarding performance

!

lssues.

In summary, CYAPCO takes these violations very seriously and is committed to implement and complete the broad scope corrective actions to improve station i performance. We will continue to keep the NRC Staff informed on our progress in these areas and are committed to demonstrating improving trends in areas of past

weakness prior to proceeding with major decommissioning activities.

The following are CYAPCO's commitments made within this letter. Other statements l within this letter are provided for information only.

!

. CY-97-058-1 Publish site-wide and department level standards and expectations by l June 30,1997.

,

l CY-97-058-2 Upgrade the work observation program by June 30,1997.

i CY-97-058-3 Complete the evaluatlon to replace the make-up valve to the

'

condensate receiver tank for the auxilisry boiler by July 31,1997.

CY-97-058-4 Complete implementation of WCM 3.3-1 by September 1,1997.

,

CY-97-058-5 Convert applicable procedures to surveillance procedures by

September 1,1997, i

! If you should have any questions, please contact Mr. G. P. van Noordennen at j (860) 267-3938.

Very truly yours,

.

_

m-

_ . _ - - - ._- . _ . . . . . . _ . . . _ _. . . _ . _ . _ _ . - - _ . _ . _ _ _ _ . _ . . _ _ . . . ~ -

- ' -

., .

. U. S. Nuclear Regulatory Commission -

_,

CY-97-058/Pago 3

.

CONNECTICUT YANKEE ATOMIC POWER COMPANY i

i fMAW T. C. Feigenba6m

Executive Vice President and

!

Chief Nuclear Officer i

cc: H. J. Miller, NRC Region i Administrator M. B. Fairtile, NRC Project Manager, Haddam Neck Plant W. J. Raymond, NRC Senior Resident inspector, Haddam Neck Plant

,

,

4

.

-

- - -

, .

,

.

.

Docket Nos. 50-213 CY-97-058 Attachment 1

'

Haddam Neck Plant Reply to Notice of Violations NRC Intearated Inspection Report No. 50-213/97-01

.

.

June 1997 m o

.

~ . -. _ -. .- -- .. .- . -. .- - -_ -

-

., .

.

U. S. Nuclear Rcgulatory Commission CY-97-058/ Attachment 1/Page 1

.

h

, Restatement of Violation 1 i

Technical Specifications 6.8.1 requires that written procedures and/or administrative  ;

policies be established, implemented and maintained covering the activities as  ;

recommended in Appendix A of Regulatory Guide 1.33. Regulatory Guide 1.33 3 requires that procedures be established governing plant operations and administration, f work controls, and security.

.

,

General Response i in addition to the station-wide efforts to raise performance standards and actions to

-

improve procedural adherence discussed in the cover letter, additional information is '

,

provided in the individual responses for each of the six examples noted.

.

j Restatement of Violation 1.a.

i Work Control Manual WCM 2.4-1, Equipment Tagging, Revision 7, was written ,

pursuant to the above and states in Step 1.2.2 that any equipment tagged with a l red tag shall not be operated by anyone. Contrary to the above, on February 19, 1997, a contractor health physics technician opened the red tagged prefilter

'

access door for auxiliary building filter FL-70-1 A.

Reason for the Violation i

(

l This violation was caused by inadequate human performance. l

An apparent cause determination was performed which concemed opening the

door to a filter housing without having the red tag cleared. The ventilation door

had a red tag on it that was not a part of the work order under which the work was :

being performed and which should have been removed prior to allowing work to  ;

l be performed on the filter unit. The technician opening the filter door assumed the l l red tag was part of the work order he was working on and did not check the red  ;

tag or his supervisor to determ' if clearance was required. The person showed a lack of understanding of the i ] process and a lack of a questioning attitude. i The technician had received trah...ig on tags as part of their initial site access and !

qualification training. l Corrective Steps that have been taken and the results achieved immediate corrective action was to counsel the technician and his supervisor on the importance of never operating red tagged equipment. Guidance was provided to station personnel, including contractors, stressing the importance of complying with the tagging program and never operating equipment with red tags. Each department head verified that a meeting with their employees had been held to discuss the event and reinforce the importance of following tagging procedures.

_

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

. .

,

'

U. S. Nucinar R:gulatory Commission

,

CY-97-058/ Attachment 1/Page 2 A Temporary Procedure Change (TPC) was written to WCM 2.1-2, Trouble Report (TR)/ Job Scoping. TPC 97-97 added a statement to the Tagging Required portion of the Evaluator's Checklist under TR/ Job Scoping Guidelines stating " perform walkdown of system / component to verify there are no conflicting tags on system / component that would affect work". This procedure change will identify tags from other clearances during the job scoping process.

A second TPC 97-98 was written to WCM 2.1-1, Work Control Process. This TPC added a statement to the step on verifying tagging under the Work Performance section for the First Line Supervisor / Designee ". . . to verify there are no conflicting tags on system / component that would affect work".

Corrective Steps that have been taken to avoid further violations Disciplinary action was taken against the employee. The TPCs have been l attached to the latest revision of the applicable procedures. A sign has been l permanently installed on the ventilation door to eliminate the need for future red tagging. .

Date when full compliance will be achieved The above corrective action is complete and full compliance has been achieved. l l

Restatement of Violation 1.b.

Procedure NOP 2.19-8A, Auxiliary Boiler Operation, Revision 2 dated January 30, 1997, requires in Step 6.1.12a that the operator maintain the condensate receiver tank one-half to two-thirds full when operating the condensate rnakeup pump manually. Contrary to the above, on January 5 and 11,'1997, an operator began a manual fill of the auxiliary boiler condensate receiver tank, and then left the area. i The condensate receiver tank overflowed on both occasions, resulting in an j ur. planned discharge to the environment.  !

Reason for the Violation:

<

.

The reason for this violation was a failure to follow procedures.

,

in these two events the operators did not follow instructions to keep the tank at the prescribed level. The operators had chosen to manually fill the tank in each case  ;

because of reliability concems of the automatic makeup valve.

.

-

. .

,

,

U. S. Nucl::ar Regulatory Commission

,

CY-97-058/ Attachment 1/Page 3 Corrective Steps that have been taken and the results achieved The involved NSOs were counseled, and guidance was issued to all Operations Department personnel on the subject of heating steam condensate receiver tank overflow. NOP 2.19-8A, "A" Auxiliary Boiler Operation, has been revised to include enhanced instructions for manual filling of the Condensate Receiver Tank. l A trouble report had been issued on this valve on February 11,1997. The Maintenance Department investigated the reliability concems related to the valve j and concluded no repair would be effective. l l

Corrective Steps that have been taken to avoid further violations l The specific events and management expectations regarding compliance with operating procedures have been reinforced. The corrective actions have been incorpcrated into the latest revision of NOP 2.19-8A, "A" Auxiliary Boiler Operation ;

l and NOP 2.19-88,"B" Auxiliary Boiler Operation.

l An evaluation to replace the valve is ongoing.

Date when full compliance will be achieved CYAPCO will complete the evaluation on replacing the makeup valve by July 31, 1997. There have been no new events for the auxiliary boiler operation related to this violation.

Restatement of Violation 1.c.

Technical Specifications 6.2.2.f requires that administrative procedures be developed and implemented to limit working hours of facility staff. Procedure NGP 1.09.was developed pursuant to the above, and requires in Step 4.6 that workers have no less than 8 continuous hours off between scheduled work periods and that workers do not work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period,24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in a 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period, and 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in a 7 day period. Contrary to the above, on February .

6-6,1997, an operator worked two shifts without 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> off between scheduled work periods. In addition, the results of a licensee investigation on January 17, 1997, identified 6 instances over a four week period in July and August 1994 in which work hours for plant staff exceeded the overtime limits.

Reason for the Violation An operator worked two shifts with less than eight hours off between the shifts without the prescribed time off because the individual scheduling the operator was unaware that tumover time is included in the total time worked.

.

m

- -

-

. .

.

U. S. Nucinar R:gulatory Commission

,

CY-97-058/ Attachment 1/Page 4 In the second case, the workers involved worked more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in a single week due to inadequate management oversight.

Corrective Steps that have been taken and the results achieved in the first case, the individual scheduling the operator was counseled and all Shift Managers reviewed the requirements for overtime controls for Nuclear Group Personnel, NGP 1.09. The specifics of these events have been given to operations personnel to convey management's expectation regarding the overtime practices and procedures.

Health Physics Management has directed all Health Physics Supervisors review

" Overtime Controls for Nuclear Group Personnel", NGP 1.09, thus reiterating Station Management expectations and responsibilities to follow procedural requirements.

Corrective Steps that have been taken to avoid further violations Station management addressed this issue generically by reiterating the importance of observing overtime limits in an article in the plant daily newsletter, CY TODAY.

Date when full compliance will be achieved

_

Corrective actions have been completed and full compliance has been achieved.

Restatement of Violation 1.d Work Control Manual WCM 2.1-2, Trouble Report / Job Scoping, Revision 3, requires in Step 1.2.3 and 1.3.3 that trouble report tags be removed and closed upon development of work packages, or for work that will be completed under blanket authorized work orders. Contrary to the above, on February 18,1997,as documented in adverse condition report ACR 97-90, a review of trouble reports by plant operators identified that (f 275 trouble reports hanging on plant equipment, 113 were associated with troublc reports or AWOs that had been either deleted, completed or canceled.

Reason for the Violation This violation was caused by inadequate personnel performance, i.e., failure to follow procedures. The primary reason cited for canceled tags still hanging was that the person canceling the Trouble Report (TR) was not going back out to remove the TR tag as required by WCM 2.1-2, Trouble Report / Job Scoping. Also, some TRs were not removed after work had been completed. A review of the various work control procedures indicated there were no programmatic problems with the tagging process. Rather, the main problem was inattention to detail.

.

_ . _ _ _ _ _ . _ _ ._. _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ . _

'

- -

. .

.

U. S. Nuclear R:gulatory Commission

,

CY-97-058/Attachm:nt 1/Paga 5  ;

Corrective Steps that have been taken and the results achieved

i All 113 TR tags have been removed. Guidance was provided to all Managers / >

Supervisors on March 4,1997, outlining the appropriate use of the TR system, ,

including the proper hanging and removal of TR tags. Supervisors are expected to :

ensure that personnel understand the requirements of WCM 2.1-1, Work Control i

'

,

Process in the attachment for Work Performance, the Work Completion section delineates one of the responsibilities of the worker as well as the first line supervisor as being to " Verify TR tag removed" and attached to the work order, f

, unless contaminated or lost.

i Corrective Steps that have been taken to avoid further violations The specifics of the violation have been reinforced to all personnel under the work !

control supervisor using additional guidance and department training to clearly ,

.

convey Station Management expectations regarding work control and tagging ;

practices.

)

'

Date when full compliance will be achieved

\

CYAPCO has completed the corrective actions and full compliance has been achieved.

!

I Restatement of Violation 1.e.

Work Control Manual WCM 3.3-1, Technical Specifications Surveillance Tracking, '

,

Revision 2, requires in Step 1.4.2 and 1.4.3 that the station procedures be audited

'

annually (initiated in January) to assure the technical specification requirements i are met. Procedure ACP 1.2-6.5A, Station Procedures, Revision 0, requires that

! technical specifications surveillances be conducted by procedure in the 4 Surveillance Procedure category. Contrary to WCM 3.3-1, as of April 4,1997, the Licensee had not completed the annual audit of procedures per WCM 3.3-1.

i

^

Contrary to ACP 1.2-6.5A, on March 12,1997 the licensee identified that technical

, specifications surveillances were completed using preventative maintenance

(PMP 9.1-31) and environmental services (ESP 14.1-4) procedures.

Reason for the Violation The requirement to audit the Master List of Technical Specification Surveillances 2 was missed when the former program administrator was transferred to a different

,

location and the work was not reassigned to another individual.

!  !

i i i

I ll M w

. -. -- - .- . - - - -- --- . - .=. _.

! U. S. Nucl::ar Regulatory Commission *

,. CY-97-058/ Attachment 1/Paga 6 l

'

The requirement to conduct surveillances with procedures in the surveillance procedure category was overlooked. for the preventive maintenance and ,

environmental procedures. This error was caused by inattention to detail. l

.

,

Corrective Steps that have been taken and the results achieved The Unit Director has assigned an individual with the responsibility for

'

implementing procedure WCM 3.3-1.

Corrective Steps that will be taken to avoid further violations ,

) The individual assigned will ensure implementation of WCM 3.3-1. The applicable

preventive maintenance and environmental services procedures will be converted '

to surveillance procedures.

Date when full compliance will be achieved l i

The individual assigned will complete implementation of WCM 3.3-1 and we will j convert the applicable procedures to surveillance procedures by September 1, ,

'

1997.

a i

Restatement of Violation 1.f. ,

Security procedure SEC 1.3-8, Package and Material Control, Revision 26, requires in Step 6.2.1.b, that all packages be searched prior to entry into the i protected area. Contrary to the above, on March 13, 1997, six boxes were l brought into the protected area without receiving the required security searches. j

!

Reascn for the Violation j

!

The reason for the violation involved poor personnel performance. The  ;

stockhandler failed to follow procedure, SEC 1.3-8, step 6.2.1b. i Corrective Steps that have been taken and the results achieved The individual was suspended pending investigation. Upon completion of the investigation, the individual's employment was terminated on March 13,1997.

Corrective Steps that have been taken to avoid further violations ,

l The remaining stockhandlers were interviewed and the incident was determined to be an isolated case. Management expectations were reaffirmed with the remaining stockhandlers.

l i

l

~

~

_._..___.m.._- ....._.

'

..-- - -  ;

U. S. Nuclear R::gulatory Commission -

!

CY-97-058/ Attachment 1/Page 7 - l

, l r l Date when full compliance will be achieved l

b Corrective actions were completed with the completion of interviews and  !

reinforcement -of management expectations and full compliance has been  !

! echieved. J i

!- Restatement of Violation 2  !

i i'

'

Technical Specifications 4.7.7 requires that reactor coolant system chemistry be  !

'

} sampled every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to assure that chloride and fluoride concentrations remain

.

below the limits specified in TS 3.4.7. The mode of applicability for the  ;

i specification 3.7.7 is "at all times". l

. (

Contrary to the above, no RCS chemistry samples were taken from November 15, l 1996 through February 6,1997.  !

>

i  :

1-

'

Reason for the Violation  !

!

$ The. apparent cause for this violation was poor judgment by Chemistry )

j Management. With the core offloaded and no flow through the reactor coolant

system, it was deemed acceptable to secure sampling on a system which was out

'  ;

of service. '

.

i i l Corrective Steps that have been taken and the results achieved l l l >

l On February 6,1997, purification flow was re-established and the Chemistry  ;

j Department commenced sampling the reactor coolant system in accordance with

,

j j the Technical Specification frequency and specified parameters, at the most  :

representative location for the plant configuration. Chemistry sampling of the RCS l

!_ has indicated that Technical Specifications chemistry limits were never j l challenged. j i t

' >

j Corrective Steps that have been taken to avoid further violations l

! . !

The Chemistry Department has and will continue to maintain reactor coolant l

chemistry sampling until a revision to the Technical Specifications has been l

'

issued. Defueled Technical Specifications to remove the sampling requirements l, l- were submitted for NRC approval, on May 30,1997. l t ~

i

'

[ Date when full compilance will be achieved CYAPCO has completed the corrective actions and full compliance has been ,

j achieved. .i

i

'

, j

i

'

.

-.s, y.. ., __...m., .- _ .-. .- . . _,-- . _ , - - - , , - ,

-

-

. , .

U. S. Nucl:ar Regulatory Commission

. CY-97-058/ Attachment 1/Page 8 Restatement of Violation 3 10 CFR 50, Appendix B, Criterion XVI, " Corrective Actions" requires that measures be established to assure that conditions adverse to quality are promptly identified and corrected. In the case of significant conditions adverse to quality, the measures shall assure that the cause of the condition is cetermined and corrective action is taken to preclude repetition.

"

Contrary to the above, the licensee did not correct the recurrence of significant conditions adverse to quality, in that there have been multiple examples of a failure to adequately implement the technical specification (TS) operational surveillance program. Past deficiencies related to missed or late TS surveillance were the subject of inspection item URI 94-27-01, and licensee event reports 95-12,96-17 and 96-22. The licensee identified the following failures to complete technical specification surveillances in a timely manner: February 6,1997, the failure to test reactor coolant system chemistry per TS 4.4.7: on February 11, the failure to test the main station battery per TS 4.8.2.2 and 4.8.2.1.c: on February 12, the failure to test the service water pumps per TS 4.7.b.2; and on March 21, the failure to verify the positions of safety related valves per TS 4.5.2.c.

Reason for the Violation Corrective action has been ineffective in preventing further missed or late surveillances required by Technical Specifications. Lack of accountability and low management standards contributed to a high tolerance for ineffective actions to correct previous problems.

Corrective Steps that have been taDn and the results achieved The Unit Director has met with all responsible managers to reiterate expectations.

Each manager reviewed the process being used to track surveillances in their

'

department and verified the process was acceptable.

Corrective Steps that have been taken to avoid further violations The responsible managers have a clear understanding of what is expected and that they are responsible for this important program.

Date when full compliance will be achieved Corrective actions have been completed and full compliance has been achieved.

.

e

Retrieved from "https://kanterella.com/w/index.php?title=IR_05000213/1997001&oldid=2966520"