IR 05000333/1988020: Difference between revisions

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{{Adams
{{Adams
| number = ML20151Q788
| number = ML20205Q355
| issue date = 07/12/1988
| issue date = 10/21/1988
| title = Insp Rept 50-333/88-20 on 880523-0603.No Violations Noted. Major Areas Inspected:Emergency Operating Procedures
| title = Insp Rept 50-333/88-20 on 880919-23.No Violations Noted. Major Areas Inspected:Nonradiological Chemistry Program, Including Measurement Control & Analytical Procedure Evaluations
| author name = Haughney C, Konklin J, Vandenburgh C
| author name = Kirkwood A, Pasciak W
| author affiliation = NRC OFFICE OF NUCLEAR REACTOR REGULATION (NRR)
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
| addressee name =  
| addressee name =  
| addressee affiliation =  
| addressee affiliation =  
Line 10: Line 10:
| license number =  
| license number =  
| contact person =  
| contact person =  
| document report number = 50-333-88-200, NUDOCS 8808110288
| document report number = 50-333-88-20, NUDOCS 8811090203
| package number = ML20151Q780
| package number = ML20205Q353
| document type = INSPECTION REPORT, NRC-GENERATED, INSPECTION REPORT, UTILITY, TEXT-INSPECTION & AUDIT & I&E CIRCULARS
| document type = INSPECTION REPORT, NRC-GENERATED, INSPECTION REPORT, UTILITY, TEXT-INSPECTION & AUDIT & I&E CIRCULARS
| page count = 24
| page count = 8
}}
}}


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U.S. NUCLEAR REGULATOP.Y COMMISSION OFFICE OF NUCLEAR REGULATORY REGULATION Division of Reactor Inspection and Safeguards  .
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Report N /88200 Docket No.: 50-333 Licensee: Power Authority of the State of New York 123 Main Street White Plains, New York 10601 Inspection At: James A. Fitzpatrick Nuclear Power Plant Inspection Conducted: May 23 through June 3, 198 Team Leader: Mh .'  T-n-t'I C.A.VanDenbufgh,TeamLeader  Date Signed Consultants: J. D. Wilcox, Jr., Prisuta-Beckman Associates D. Schultz, Comex, In D. Jarrell, Battelle-Pacific Northwest Labs 0. Meyer, EG&G-Idaho Other NRC Personnel Attending Exit Meet .gs: C. J. Haughney, Chief, RSIB, NRR; J. Konklin, Section Chief, NRR; D. Langford, Project Engineer, NRR; J. Johnson, Section Chief, RI; A. J. Luptak, Senior Resident Inspector; R. A. Plasse, Resident Inspecto Reviewed By:    7// 2./T7 ames E. Konklin, Chief  Date S'igned
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  & Integration Section, NRR Approved By:  1
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arl s J. Ha sial Inspe neyf Chie ion Branch 4R te g ned /
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PDR ADOCK 05000333    l 0  PNU    '
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U.S. NUCLEAR REGULATORY COMMISSION
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SUMMARY OF RESULTS Scope:
On May 23 througn June 3, 1988 an NRC inspection team conducted an inspection of the James A. Fitzpatrick Nuclear Power Plant (JAFNPP) Emergency Operating Procedures (EOPs). JAFNPP is a BWR-4 with a Mark I containment. The objectives of the inspection were to determine whether the E0Ps are technically correct, whether the E0Ps can be physically carried out in the plant, and whether the E0Ps can be correctly performed by the plant staf The team accomplished the inspection by performing a comp" ' the BWR Owners Group Emergency Procedure Guidelines (EPGs) to the Pa..< Specific Technical Guidelines (PSTGs); a comparison cf the PSTGs te 'ne E0Ds; a review of the calculations performed to develop the plant specific curves, values, and setpoints utilized in the E0Ps; a plant walkdown of all the E0Ps and the Abnormal Operating Procedures (A0Ps) referenced by the E0Ps; a simulation of two emergency scenarios using a full size control room mock-up; a human factors review of the procedures and plant operations, including interviews of nine licensed and non-licensed personnel who utilize the E0Ps and A0Ps; a detailed review of the containment venting procedures; and a review of your on-going program for evaluation of E0Ps. The inspection was primarily focused on the adequacy of the end prcduct and not on a review of the process to develop the E0P Results:
Based on a review of the E0Ps and supporting calculations, the inspection team concluded that although the PSTGs have not been controlled and maintained up-to-date as a design basis document, the E0Ps are a technically accurate incorporation of the EPG The plant walkdowr.3, operator interviews and tne simulated E0P scenario identified several minor deficiencies; however, the team concluded that the E0Ps can be accurately accomplished using the existing controls, instruments, and equipment. Based on the human factors review of the E0Ps, Writer's Guide implementation, plant walkdowns, and the E0P simulation, the team concluded that the E0Ps and associated procedures have the useability to provide the operators with an effective accident mitigation tool and can be correctly performed by the plant staf Several concerns were identified which will require further licensee action to resolve. The most sigr.ificant concerns were:
1.) The E0Ps and the PSTGs had not been maintained a.c a design basis document and theref ore have not been maintained up-to-date and appropriately controlled. This resulted in several discrepancies between the PSTGs and the E0P (Section 3.a and 3.2)
2.) Plant process computer setpoints did not correspond to the E0P entry conditions and potential confusion existed in the measurement and indica-tion methodology of suppression pool level. In addition, outstanding validation comments concerning the suppression pool level measurement methodology had not been satisfactorily documented. (Section 4.a)
i 3.) In a few instances, information or e uipment necessary for the performance of the E0Ps had not been provide Section 4.b)
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==REGION I==
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Report N /88-20 Docket N License No. DPR-59  Priority Category C Licensee: Power Authority of the State of New York P. O. Box 41 New York, New York 13093 Facility Name: James A. FitzPatrick Nuclear Power Plant Inspection At: Lycoming, New York Inspection Conducted: September 19-23, 1988 Inspectors: / to ct N  d[ O/,/hIh A. Kirkwood, Radiation Specialist  date r
Approved by:  M a. _  c '/ 2019.99'
W. J. Pasciak, Chief, Effluents Radiation dat'e Protect.on Section Inspection Summary: In;pection on Se.)tember 19-23, 1988 (Report N /88-20)
Areas Inspected: Routine, announced ins; ion of the non-radiological chemistry program. Areas reviewed included measuremar t control and analytical procedure evaluation Results: No violations were identifie '
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4.) The E0P simulation adequately demonstrated that the minimum shift crew described by Technical Specifications was sufficient to accomplish the required actions of the E0Ps. However the team could not conclude that sufficient personnel would be available to accomplish all of the actions j required in an emergency, such as implementation of the Emergency Plan or l activation of the Fire Brigade., coincidental with implementati6n of the  l E0Ps. In addition, a method of placekeeping was not used by the-operators  i during the performance of the E0P Placekeeping methods have not been l utilized during periodic training and were not supported by the  ,
  "DR  GS1008ADOCK 05000333 FDC Q
procedure [Section 5.c.1 and 5.c.3)
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5.) A response to the Safety Evaluation incorrectly indicated that action  ,
 
statements would not be carried over from one page to anothe (Section
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6.1.c)
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6.) Sufficient guidance was not provided in the E0P for Primary Containment Control to describe the calculation of the Heat Capacity Temperature Limi (Section 6.2.c)
7.) An evaluation had not been performed to demonstrate the capability of the Standby Gas Treatment System to operate under the anticipated accident conditions of hign pressure and temperature during containment ventin (Section 7)
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CETAILS
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  *R. Converse, Resident Manager    i G. Tasick, QA Supervisor
  *W. Fernandez, Superintendent of Power .
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; *E. Mulcahey, Radiological and Environmental Service Supervisor
  *A. McKeen, Chemistry General Supervisor
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W. Hamblin, Chemistry Supervisor
  *C. Boucher, Chemistry Supervisor    ,


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D. Johnson, Waste Management Supervisor    i O. Lindsay, Operation Superintendent    .
.     1 TABLE OF CONTENTS EMERGENCY OPERATING PROCEDURE INSPECTION at James A. FitzPatrick Nuclear Power Station (Inspection Report 50-333/88200)   . ;
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      '-Page IhSPECTION 06JECTIVE......................................... 1 BACKGROUN0................................................... 1 PROCEDURE REVIEW............................................. 2 S.1 EPG/PSTG REVIEW......................................... 3 3.2 PSTG/EOP REVIEk......................................... 3 3.3 CALCULATION REVIEW...................................... 4 PLANT WALKD0WNS.............................................. 5 E0P SIMULATION USING CONTROL ROOM M0CK-UP............ ....... 9 HUMAN FACTORS REVIEW AND INTERVIEWS.......................... 12 6.1 WRITER'S GU!DE IMPLEMENTATION........................... 12 6.2 OPERATOR INTERVIEWS..................................... 13 CONTAINMENT VENTING.......................................... 16 Oh-GOING EVALUATION 0F E0FS.................................. 18 EXIT MEETING................................................. 19 1 REFERENCES 10.1 PERSONNEL CONTACTE0..................................... 19 10.2 PROCEDURES REVIEWED..................................... 20
3 Analytical procedures Evaluttion
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During the inspection, .tandard chemical solutions.were submitted by
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the ins 1ector to the licensee for analysis. The standard solutions were prepared by CruoKhaven National Laboratory (BNL) for the NRC Region I, and were analyzed by the licensee using normal methods and equipment. The analysis of standards is used to verify the licensee's capability to  !
. monitor chemical parameters in various plant systems with respect to  ;
Technical Specifications and other regulatory requirements. In addition, the analysis of standards is used to evaluate the licensee's analytical procedures with respect to accuracy and practsio ,
J The results of the standard measurements comparison (Table 1) indicated  l
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that seven out of thirty-one measurements were in disagreement unust the  !
1 criteria used for comparing results (see Attachment 1). However, by  1
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redilution of the NRC standards, preparation of new standards calibrai. ions, !
3 and using different technicians to do the same analysis, all measurements  ,
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l The chloride disagreement, at approximately 10 ppb, was due to limitations of the statistical method for determining agreement. The actual Lic/NRC  ,
i ratios differed by only four percent (4%). Similarly, the silica Lic/NRC  I
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ratios differed by less than ten percent (10*4). This was resolved and  {
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Copper, nickel, chromium and iron disagreements were apparently due to
:  calibration of the inductive coupled plasma analyzer (ICP). Subsequent j  analysis after recalibrations brought all samples into agreement.


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I No violations were noted.
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i l      \ INSPECTION OBJECTIVE The inspection team reviewed the licensee's Emergency Operating Procedures l (EOPs), operator training and plant systems to accomplish the following objectives in accordance with NRC Temporary Instruction (TI) 2515/92:
(1) Determine whether the E0Ps conformed to the vendor generic guidelines and were' technically correct for the James A. Fitzpatrick Nuclear Power Statio (2) Assess whether the E0Ps can be physically carried out in th! olant using existing equipment, controls, and instrumentation, under tre expected environmental condition (3) Evaluate whether the plant staf f has been adequately trainto to perform the E0P functions in the time availabl . BACKGROUND Following the Three Mile Island (TMI) accident, the Office of Nuclear Reactor Regulation developed the "TMI Action Plan," (NUREG-0660 and NUREG-0737) which required licensees of operating plants to reanalyze transients and accidents and to upgrade Emergency Operating Procedures (E0Ps) (Item I.C.1). The plan also required the NRC staff to develop a long-term plan that integrated and expanded efforts in the writing, reviewing, and monitoring of plant procedures (Item I.C.9). NUREG-0899, "Guidelines for the Preparation of Emergency Opera-ting Procedures,' represents the NRC staff's long-term program for upgrading E0Ps, and descr40es the use of a Procedures Generation Package (PGP) to prepare E0P The licensees formed four vendor owner's groups corresponding to the four major reactor types ir. the United States: Westinghouse, General Electric, Babcock &
Wilcox, and Combustion Engineering. Working with the vendor company and the NRC, these owner's groups developed generic procedures that set forth the desired accident mitigation strategy. For General Electric plants, the generic guidelines are referred t.o as Emergency Procedure Guidelines (EPGs). These EPGs were to be used by licensees in developing their PGPs. Submittal of the PGP was made a requirement for the James A. Fitzpatrick Nuclear Power Plant by Confirmatory Order dated June 12, 1984. Generic Letter 82-33, "Supplement I to NUREG-0737 - Requirements for Emergt.nty Response Capability" required each licensee to submit to the NRC a PGP wnich includes:
(1) Plant Specific Technical GuidElires (PSTGs) with justification for safety significant ditterences from tne EP (2) A Plant Specific Writer's Guide (PSWG).


(3) A cescriptico of the program to be used for the verification and validation of E0P l (4) A description of the training program for the upgraded E0P '
. Measurement Control Evaluation f
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Verification of the licensee's measurement capabilities on actual plant water samples is done by splitting samples with the licensee and Brookhaven National Lab (BNL). Samples were taken to be sent to BNL for
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Plant specific E0Ps were to have been developed that would provide the operator with directions to mitigate the consequences of a broad range of accident and multiple equipment failure For various reasons, there were long delays in achieving NRC approval of many of the PGPs. Nevertheless, the licensees have all implemented their E0Ps. To determine the success of this implementation, a series of NRC inspections are being performed to examine the final product of the program: the E0Ps. A representative sample of each of the four vendor types has been selected for review by four inspection teams from Regions 1, II, III and I An additional 13 inspections have been scheduled at f acilities with General Electric Nark I type containments. The latter inspections are being conducted by the Office of Nuclear Reactor Regulation and include a detailed review of the containment venting provisicns of the E0Ps. This inspection at the James A. Fitzpatrick Nuclear Power Plant is the first of the 13 Mark I inspection . PROCEDURE REVIEW This portion of the inspection was performed to determine whether the JAFNPP E0Ps have been prepared in accordance with the current Procedures Generation Package (PGP) and the Plant Specific Technical Guidelines (PSTGs). The inspection team compared Revision 3 of the BWR Owners Group Emergency Procedure Guidelines (EPGs) to the PSTGs, and the PSTGs to the E0Ps. All differen m were identified and reviewed to ensure that safety significant deviations were identified in the PGP and that a documented basis existed for all deviation A review of selected calculations was performed to ensure that plint specific values utilized in the E0Ps are correct and have been performed 1. accordance with a documented engineering analysi Section 10.2 of this report lists the procedures reviewed, Reccrd Control in the process of reviewing the PSTGs and the E0Ps, the team identified that the original copies of the E0Ps, EPG Calculations (Appendix C)
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and the Plant Specific Technical Guidelines (PSTGs) were all being temporarily stored in the Operations Department Administrative Office as uncontrolled documents and were not being upgraded and maintained up-to-dat HUREG-0899, "Guidelines for the Preparation of Emergency Procedure Guioelines" indicates, in paragraph 4.4, that the PSTGs are the primary basis for plant E0Ps and, as such, should be subject to examina-tion under the plant's Quality Assudance (QA) Program, and are also required to be accurate and up-to-dat Further licensee action is l necessary to upgrade and maintain tne PSTGs as require ;
i A review of Administrative Procedure F-AP-1.4. "Control of Plant  l Procedures," Plant Standing Order PS0-4, "Quality Assurance and Plant '
Operating Records," and JAFHPP Records Retention / Turnover Schedule, Re .1 indicated that the location and length of storage of the PSTGs and supporting records were not in accordance with the plant's administrative requirements. The QA Department had identified similar concerns with record storage and the timeliness of turnover in QA Audit No. 584 dated February 20, 1987 and QA Audit No. 646 dated April 22, 1988. However, at
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the time of the inspection, noce of the E0P records had been turned over l l
 
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to Document Contro The initial upgraded (symptomatic) E0Ps were issued on December 29, 198 Further licensee action is necessary to maintain and control these records as require .1 EPG/PSTG Review    ,
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Four minor differences were identifieo between the EPGs and the PSTGs,'as detailed below. The team concluded, based on a review of the PSTGs and the differences identified, that significant discrepancies did not exist and that there were no adverse affects on the adequacy of the JAFNPP E0P Further action should be taken to ensure that future revisions or upgrades of the E0Ps correct these discrepancies, The selection of 1090 psig as the lowest Safety Relief Valve (SRV)
pressure in PSTG step RC/P-2 did not consider the setpoint tolerances of the SRVs. The EPG basis (Appendix B) indicates that the intent of RC/P-2 is to establish a control band for pressure control of the reactor at which the SRVs will not cycle. The use of the lowest SRV setpoint pressure without consideration of the setpoint tolerance of the SRV could result in cycling the SRV if the reactor pressure is controlled at 1090 psig. Additional operator guidance such as a lower pressure control band would ensure that reactor pressure is controlled below the pressure at which the SRV would lif Other steam driven equipment available at JAFNPP such as Reactor Feed Pump (RFP) turbines, RFP drains, steam seals, steam jet air ejectors and off gas heaters have not been incorporated into PSTG step RC/P- Amplifying information for Group 1 Isolation signals, such as initiating conditions and isolation valve identification, has not been provided in PSTG step RC/L- Alternate injection systems which are available at JAFNPP, such as demineralized water transfer to the Standby Liquid Control test tank, have not been incorporated into PSTG contingency Cl-1, "Level Restoration."


3.2 PSTG/EOP Review
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independent verification. The reactor building closed cooling water (RBCCW) was sampled. One sample was spiked with a standard solution of anions and split with the licensee and BNL. The second sample was not spiked and was also split with the licensee and BNL. The standard spike  i solution was prepared by-BNL for the NRC Region I. On completion of the analyses by BNL and the licensee, an evaluation will be mad . Quality Ast.urance/ Quality Control Program


Six minor differences were identified between the PSTGs and the E0Ps, as detailed below. The team identified that in each example the E0P w6s correct  ;
The inspector performed a selected review of the licensee's program for the quality control of analytical measurements made by the chemistry i department. The review was performed, in part, with respect to criteria
and concluded that the discrepancies were the result of not maintaining the  l PSTGs up-to-date as required. No conditions which wculd adversely affect the performance of the E0Ps were noted. Further action is necessary to correct
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contained in the following:
l these deficiencies and to control and maintain the PSTGs appropriatel I F-E0P-1, "E0P Cautions," Caution No. 6 had a different Reactor Pressure Vessel (RPV) water level specified than the PST6 Caution #6, because tne E0P was revised af ter an evaluation identified that the drywell tempera-ture referenced in the PSTG was incorrec F-EOP-2. "RPV Control (Boron Injection Not Required)," specified 335 psig for the nigh/ intermediate RPV pressure region vice 300 psig as specified l in PSTG Contingency Cl-E, because the E0P was revised after the PSTG was develope '
  * 10 CFR 50, Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants"   '
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Regulatory Guide 4.15. "Quality Assurance for Radiological Monitoring
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;  Programs (Normal Operations) Effluent Streams and the Environment''
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The following licensee procedures were reviewed to determine the adequacy ,
of measurement control:    '
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Quality Assurance / Quality Control Program, dated 7/19/88  l l
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CLI-17, Dionex Ion Chromatograph and Hewlet Packard Reporting  ;
Integrator, dated 7/19/88
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CLI-34, Routine Verification and Calibration of Laboratory Pipettes, dated 8/19/83 4       ,
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CLI-35, Plasma Spec 2-5 leeman Labs Operation and Calibration, dated  i 9/4/87      *
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! * CAP-19, Silica Determination, dated 8/5/88
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} Also the following records were selectively sampled to determine the  !
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extent of measurement control:    l
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1 l F-E0P-4, "Primary Containment Control," step 2.3 indicated that the suppression pool scram temperature is 100 degrees F vice 110 degrees F as l specified in PSTG SP/T-3 cue to a typographical error.
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QA/QC Data Log from 1/14/88 to 9/22/88, Intra and Interlab l Cross-check results    i
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Non-Conformance Reports from 3/17/88 to 9/18/88, both completed and g pending      '
  * Aiministrative Review, AM-88-01 for week of 8/4-12/88
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i F-EOP-5, "Secondary Containment Control," Table F-E0P-5.1 listed dif ferent I
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instruments and setpoints than specified in PSTG Table 1, becaus Subsequent plant modifications were not incorporated into the PSTG F-EOP-5, "Secondary Containment Control," did not have an entry conoition for Unit Cooler Temperature above 104 degrees F as specified in the secondary containment control guidelines of the PSTG, because subsequent plant modifications were not incorporated into the PST F-E0P-7, "RPV Flooding," did not utilize the Residual Heat Removal (RHR)
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Keep Fill System to maintain RPV water level as specified in PSTG C6-4, because the Keep Fill System had not been made operable at JAFNP .3 Calculation Review Specific values from the E0Ps were selected for review to determine if the values were correctly calculated based on the plant specific differences and the guioance of the EPGs. The team identified that the calculations were clear, orderly and performed in accordance with the guidance of the EPGs. Any deviations were noted and substantiated. The calculations were observed to include values for each of the cautions, steps and curves. However, as noted in Section 3.a, the calculations have not been controlled as a plant basis document in accordance with the requirements and guidance of NUREG-0899. The following calculations were reviewed: Hot Shutdown Boron Weight - This calculation was performed to support tne values incorporated as part of F-E0P-3, "RPV Control," and documented in Section 24, Appendix C, page C24-1 of the EPG Calculation Minimum Alternate RPV Flooding Pressure - This calculation was performed to support tne values incorporated as part of F-E0P-3, "RPV Control (Boron Injection Required)," and F-E0P-7, "RPV Flooding," and documented in Section 21, Appendix C, page C21-1 of the EPG Calculation Suppression Pool Load Limit - This calculation was performed to support tne values incorporated as part of F-E0P-2, "RPV Control (Boron injection Not Required)," F-E0P-3, "RPV Control (Boron injection Required)," and F-EOP-4, "Primary Containment Control," ano documented in Section 4, Appendix C, page C4-1 of the EPG Calculation Heat Capacity Temperature Limit for the Suppressior Pool - This calcula-tion was performed to support the values incorporated as part of F-EOP-4,
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  "Primary Containment Control," and documented in Section 3.3, Appendix C, page C3-1 of the EP6 Calculation Maximum Drywell Spray Flow Rate Limit - This calculation was performed to ensure that the evaporative cooling pressure drop was less than t'
permissible design to support a drywell spray flow rate of 10,000 gallons 1 per minut l l-4-
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  * Nanopure Con D H O conductivity measurements from 3/24/88 to 9/22/88
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I 1 Drywell Spray Initiation Limit - This curve was developed to determine the '
pressure limit wnen spraying at the maximum drywell spray flow rate to ensure that the reactor building to containment design negative pressure l differential limit is not exceeded by initiation of drywell or wetwell spray Maximum Primary Containment Water Level Limit - This calculatio'n~
determined that the maximum containment water level was based on the highest primary containment vent elevation as opposed to the design hydrostatic loading at the most limiting containment locatio . PLANT WALK 00WNS in order to assure that the E0Ps can be successfully accomplished, the team performed in-plant walkdowns for all the E0Ps and referenced A0Ps. The team verified that E0P instrument and control designations were consistent with the installed equipment and that indicators, annunciators, and controls referenced by the E0Ps were available to the operators. The location and control of E0Ps in the Control Room was verified. With the assistance of licensed operators, the team physically verified that activities which would occur outside of the Control Room during an accident scenario could be physically accomplished and that tools, jumpers, and test equipment were available to the operators. The post accident radiation survey map was reviewed to ensure that remote operations were not prohibited by environmental conditions. The procedures reviewed are listed in Section 1 During the performance of the plant walkoowns, the team identified a specific strength in that, general area and equipment cleanliness was exceptional. The plant walkdowns also identified several discrepancies which are broken down into the following six area E0P Entry Conditions The team reviewed the entry conditions and associated instrumentation for F-E0P-4, "Primary Containment Control," and identified the following methods for determining suppression pool water level in the control room: Utilizing the Safety Parameter Display System (SPDS) which indicates 13.95 feet normal torus water level at 100% powe . Utilizing the plant process computer which indicates level in Wide Range (+72 in to '-72 in) and harrow Range (+6 in to -6 in). Utilizing instruments at Main Control Board Panel 09-3, which iacicctes approximately 14 feet normal torus water level at 100%
powe The team no;ed that a narrow range level instrument which was not referenced by the E0P was available.on Standby Gas Treatment System Panel 09-75, which is located in the back of the control room. However an audible local or control room supervisory alarm was not availabl As a result of the review, the team identified that if the SPDS were not available as inferred by NUREG-0737, Item 4.1.c, the operator's ability to-5-
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    (daily checks)
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Reactor Water Data Sheets from 1/1/88 to 7/31/88
  * ICP H2 Cooler TBCLC side sludge sample analysis dated 9/21/88
  * Control Charts for Cr, Fe, Cu, Ni dated 9/13/88
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The inspector noted a few minor problems with the QA/QC program procedure
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identify entry conditions would be hampered because the plant process computer alarm setpoints were +2.5 and -2.5 inches vice the E0P entry conditions of 0.0 and -1.5 inches. The team also noted that E0P-4,
of 7/19/88. First, Section 5.6.7, "Use of Analytical Standards," allows l  the preparation of standards from the same lot. Thi., would not allow verification of standard quality and sample results. Also, in Section    ;
"Primary Containment Control," Section A, Item 6 listed drywell average temperature above 135 degrees F as a condition for entry into the E0 Again, if the SPDS were net available as NUREG-0737 Item 4.1.c i_nfers, the operator's ability to detect E0P entry conditions would be hampered because the alarm setpoints for plant process computer points M085 and M086 were 65 degrees F above the E0P entry condition of 135 degrees F (i.e. 200 degrees F) and the 09-75 Panel did not have a audible local alarm or a sur ervisory alarm in the control room. Further action is necessary to revise the computer alarm setpoints to values which support the E0P entry conditions, in addition, the team identified a human engineering deficiency which had the potential for cperator confusion with respect to the indication of torus water level. Technical Specifications 3.7.A.1.a and 3.7.A. specify maximum and minimum vent submergence levels of 53 and 51.5 inches respectively as the Limiting Conditions for Operation (LCO). However, the E0P entry conditions were specified as 0.0 and -1.5 inches. Although engineering correlatien exists between the TS LC0 (53.0 to 51.5 inches)
5.11, "Alternate Analytical Labs," the NRC is listed as an alternate lab    !
and the parameters monitored by the plant process computer and Panel 09-75 (0.0 to -1.5 inches) and SPOS (13.95 feet), the indication of torus water level in different units and methods is confusing. Interviews with opera-ting staff confirmed that the correlation was not immediately apparen The licensee has indicated that this deficiency will be resolved with the implementation of Revision 4 of the EPGs at wnich time all references to torus water level will be in feet of water in the torus. Additional action is necessary in the interim to ensure that operator confusion does not exis The team attempted to determine whether inadequacies in the licensee's program of E0P validation may have contributed to the above discrepanc)
that can be used for indepe.: dent sample splits. The licensee cannot fulfill its QC Prcgram Sample Split Schedule, for annual independent    .
related to tne torus water leve The plant specific validation of the JAFNPP E0Ps was performed in December 1984 by a shift supervisor in the j control room. The validation criteria of F-AP-2.2, "Procedures for Emergency Operating Procedures," Appendix 0, "EOP Validation Checklist,"
lab checks, since they cannot be assured of NRC splits at this frequenc :  The NRC should be removed from this list.
were used to perform the validatio The completed checklist included reference to the inconsistencies concerning the suppression pool level i instrumentation, however objective evidence in the form of written I resolution to the validation comment was not available in the licensee's record The team determined that val 106 tion comments of a typographical nature were incorporated into the E0Ps; however, the resolution of coments requiring engineering resolutions was not apparent. Future action is i necessary to resolve this discrepancy as well as any other unresolveo '
validation comments, b. Lack of Equipment or Inf ormation which could Af tect Performance of the Procedure F-E0P-4, "Primary Containment Control," Step 4.7, required the operators to vent the containment in accordance with F-A0P-35, "Post Accident Venting of Containment," to maintain pressure below the limits of Figures F-E0P-4.6a and F-EOP-4.6b. The curves were specifically applicable to-6-


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3  The QA/QC program is otherwise well documented and improvements in measurement control are evident. Analyses are completed in a timely fashion. Trends are easily detected on control charts and out-of-l
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specification situations are followed up and investigated promptl Corrective actions are technically sound and retraining of technicians, where needed, was noted. A thorough chemical contaminant control program was detailed to the inspecto No violations were noted.
- Management Controls 5.1 Facilities and Equipment      ;
;    Expanded laboratory space and improvements in the work environment    ;
were noted since the last inspection. A storage area has been converted to an analysis area for metals and anions, creating more
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workspace in the lab proper. Also, a dropped ceiling and new paint    i provide a brighter more professional work area. The ICP (inductive    ;
coupled plasma) analyzer is fully operationa '
5.2 Staffing l
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Staffing appaars adequate for meeting outage requirements. A special
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chemical decontamination of recirculation lines had required the
,    staff to work six, twelve hour days each week. This is the maximum allowed under the licensee guidance. Seven, ten hour days will be resumed for the remainder of the outage. The inspector noted an alert, technically competent group of eight fully qualified techni-cians and two first line supervisors, with many years of experience
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pressure instruments MENS 0R 16-1-PIT-104 and 27-PT-101A/B respectivel Figure F-E0P-4.6a was not provided in Procedure F-A0P-35. Therefore, the possibility for operator confusion existed, in that both pressure instru-ments and their respective figures were not provided for use in F-A0P-3 The licensee indicated that F-A0P-35 will be modified to control contain-ment pressure using MENS 0R 16-1-PIT-10 _
During the walkdown of F-A0P-34, "Alternate Control Rod Insertion," the team identified that tool cabinets containing the tools and equipment required to perform the control rod withdraw line venting portion of the procedure did not contain the necessary equipment for handling the venting components in the anticipated accident environment. Alternate control rod insertion is accomplished by venting the Hydraulic Control Unit (HCU) vent valves with a flexible stainless steel hose. Venting even relatively small flow rates of high temperature primary water would result in a dangerous two-phase steam-water mixture through the flexible hose. A caution in the procedure did not appear adequate in view of the potential adverse impact on the performance of the procedure. As a result of this concern, the licensee took immediate action to provide safety equipment (i.e. welder's gloves) in the cabinets. The team also identifieo that the venting procedure directed the tlexible hose discharge to a floor drain and that no provision existed for securing the discharge end of the hose against the reaction loading during venting. The licensee indicated that a modification has been initiated to fabricate and install permanent drain connections for this vent procedur During the walkdown of F-A0P-43, "Plant Shutdown From Outside the Control Room," a potential deficiency in the performance of the procedure was identifiec, in that a delay was experienced when the remote shutdown panel could not be opened using the on-shift key ring. Subsequent investigation identified that the necessary key was in the previously opened, staged equipment box. The inspection team questioned the benefit of using a separate key for the remote shutdown panel, in that no additional protection is provided and the potential for confusion is increase Further licensee action is necessary to resolve this deficienc F-A0P-43, also required the operation of various Emergency Diesel Generator (EOG) controls. Step D.2.2, which required the verification of control power availability by checking the indicator lights, did not indicate which lights would be energized to indicate that control power was available to the EDG synchronization circuits. The location of the inoicating lights was not apparent and should be clarified to prevent confusion, c. Uncontrolled Operator Aids As a result of verifying that operator aids posted on plant instrumenta-tion and control panels were the latest revision ano administratively controlled, the team identified that Operations Department Standing Order No. 21, "Posting of Operator Aids," paragraph 7.7, required an annual review of all operator aid The licensee was unable to demonstrate that this review had been performed in 1967 and 1988 and as corrective action initiated an immediate verification. As a result of this verification, six operator aids were identified which required revisio None of these-7-
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_ . , _ , - - , _ _ _ , , ~ _ ~ . . _ _ _ , . - _ . _ _ . . . .-,% .-.m, w-_ .g .- ,__,_.,s. ..m.,e p ,-~,,m- -- - , , . , . , . _ _ _ , .
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aids could have resulted in performing the E0P actions 'ncorrectly. The licensee indicated that the audit of operator aids will be incorporated into a new surveillance procedure to ensure the audit is accomplished as required, Referencing Errors in Procedures The following discrepancies in referencing were identified during the review and walkdown of the E0Ps. Based on a review of the context and effect of these deficiencies, the team concluded that their effect on the ability to adequately perform the E0Ps was mino . F-E0P-5, "Secondary Containment Control," Table F-E0P-5.1, incorrectly identified area temperature instruments, 23-MTU-202A and 23-MTU-202B (located at Panels 09-95 and 09-96 in the Reactor Building) as 23-MTU-201C and 23-MTU-201 . F-A0P-15, "Recovery from an Isolation," Paragraph I.C.2, incorrectly identified the location of alarming Reactor Building radiation monitors as Panel 09-12. The correct location for the alarmine function is Panel 09- . F-A0P-36, "Stuck Open Relief Valve (50RV)," Paragraph A.6, incorrectly indicated in the last parenthetical note that the symptoms of a stuck open reliet valve are four energized solenoid indicating light A single energized indicating light is the correct sympto . F-EOP-4,"Primary Containment Control," Step 5.2, incorrectly referred to F-EOP-2, "RPV Control (Boron Injection fiot Required)," Step 4 for emergency depressurization. The correct reference is Step 4.8 and 4.9. A similar reference occurred in Step 5.3.1 of F-E0P- . F-E0P-2, "RPV Control (Boron Injection flot Required)," Table F-E0P-2,1, incorrectly referenced step 4 in the third action item under pressure high / level decreasing. The correct reference is to Steps 4.8 and A similar reference existed for the fourth action item in the same section of Table F-E0P-2.1, in that the reference j should hve been to Steps 4.6 and 4.7 vice Step I Incorrect / Inadequate labelling The following minor examples or incorrect or inadequate labelling were identitied. The team concluded that these examples did not adversely
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l affect the performance cf the E0P l F-A0P-36, "Stuck Open Relief Valve (50RV)," required the removal of l fuses at Panel 09-47 in the Relay Room. Four fuses were not labelled and eight fuses were labelled with temporary marking (i.e. Dynotape).


The licensee indicated that the fuse location prevents permanent marking of the fuses and that further action would be taken to remove the temporary markings and ensure that the fuses are adequately identified by an cperotor ai _ _ _ _ - _ _ _ _ _ _ _ -
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' F-0P-37, "Nitrogen Ventilation and Purge; Containment Atmosphere Dilution (CAD); Containment Vacuum Relief and Containment Differ-l  ential Pressure Systems," Section G.1.b.4, required operation of the 27-M0V-121 valve on Panel 27PCP in the Relay Roc The valve switch was labeled. "Bypass Valve", however the correct name is ' Purge Exhaust fan Bypass Valve."  _
l Several additional minor examples of informal marking (i.e. use of black markers) and temporary labelling (i.e. adhesive labels) were ioentifieo during the plant walkdowns. The team observed that specific actions have been undertaken to upgrade the equipment labelling throughout the plant. Although no examples were identified which would prevent the procedures from being accurately performed, further actions are necessary to upgrade the labelling of instruments ano component In particular, the Hydraulic Control Unit (HCU) vent valves used in F-A0P-34, "Alternate Control Rod Insertion," shoulo be permanently labelled. These valves were identified witn temporary marking (i.e. magic markers). E0P SIMULATION USING CONTROL ROOM M0CK-UP To ensure that the E0Ps could be correctly implemented under emergency cer.ditions, two accident scenarios were developed and conoucted in a 6-hour session utilizing licensed operator The accident scenarios were accomplished to determine whether the E0Ps provide the operators with sufficient guidance such that their required actions during an emergency were clearly outlined; to verify whether the E0Ps could cause the operators to physically interfere with each other; to verify that the procedures did not duplicate operator actions unless required; and to verify that transitions from one E0P to another or to other procedures were accomplished satisfactorily, Control Room Mock-Up Because JAFNPP does not have a site specific simulator and the plant was operating at full power thereby making extensive control room access difficult, a full scale photographic mock-up of the control room was utilized for tne scenario. To ensure a realistic test of the E0Ps using the control room mock-up, the following provisions were verified: The photographic reproduction replicated the actual control room with enough fidelity 50 as not to cause confusion or detract measurably from the ability of the operating crew. To ensure fidelity, the mock-up was examined by the team and the license Although mincr differences were identified, the team concluded that the simulation would not be excessively impaire . Realistic scenarios which preserved a true time line were required to be developed and executed in a believable manne The licensee supported these efforts by providing a reactor analyst who assisted the NRC operator examiner to produce two scenarios based on an integrated analysis of computer generated transients. The time response of the reactor, primary, and secondary parameters were caref ully followed. The scenarios were then examined by tne  i licensee's training staff to confirm the adequacy and accuracy of the  l


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among the Two trainees and two contractors supplement this core of employee First line supervisory duties are transitioning such that one individual will handle radiological chemistry and the other will handle non-radiological chemistry. Both first line supervisors report to the Chemistry General Supervisor; who is responsible to the Radiological and Environmental Service Supervisor; who in turn reports to the Superintendent of Power and then to the Resident Manage The long-term, technically competent, staff is viewed as a positive program asset, as is the cooperation and attitude noted by the inspecto No violations were foun . Exit Interview The inspector met with the licensee's representatives (denoted by an asterisk in S(ction 1) at the conclusion of the inspection on   i September 23, 1988 and summarized the scone and findings of the  i inspectio ..' _,
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system response To implement and control the scenario, a certified training staff member and the NRC operator examiner functioned as controllers, providing data input as required by the operators in accordance with the timed and scripted accident scenario . Licensed operators were required to simulate actions and responses based on inputs from the controller The shift crew was requested and fully supported the mock-up requirements to simulate obtaining data from the appropriate instrumentation and location Three NRC l team members were used to monitor the operators' response and imple- l mentation of the E0Ps. Based on the operators' actions and the te6m's review, the use of the mock-up was determined to be a adequate l simulation for performance evaluation of the E0P l l Scenario Description The team developed two simulated accident scenarios. Both were selected i to exercise parallel E0P paths and contingency procedures, with a special emphasis on reaching and utilizing the containment venting procedure. The specific paths were designed to invoke PRA-based risk significant operator actions as a means to demonstrate E0P adequac The first scenario involved the loss of the main condenser with an )
ATTACHMENT 1 CRITERIA FOR COMPARING ANALYTICAL MEASUREMENTS This attachment provides criteria for comparing results of capability test In these criteria the judgment limits are based on the uncertainity of the ratio of the licensee's value to the NRC value. The following steps are performed:
anticipated transient without scram (ATWS) and delayed Standby Liquid Control System (SLC) actuation. The transient was initiated with an unrecoverable failure of the main condenser. The ATWS condition was then required to develop sufficient internal energy in the primary to produce a containment venting situation later in the scenario. The workability of the text-type E0Ps in a multiple path sequence was demonstrated by requiring the operators to simultaneously follow F-EOP-2, "Reactor Pressure Vessel (RPV) Control," for reactor power, reactor pressure, and reactor level control. As alternate rod insertion techniques were being pursued, primary containment precedures and later SLC injection (with SQUIB valve failure) were entered due to increases in torus temperatur Power control with vessel level was then utilized to minimize power while alternate SLC injection methods were tried. On approaching the containment venting pressure, alternate rod insertion (withdraw line venting) and CR0 boron injection were successfully implemented to allow reactor shutdown and plant recovery. Following containment venting, shutdcwn cooling was established and the scenario was terminate The second scenario involved an unisolable LOCA in the Secondary Containmen This sequence postulated that the inboard isolation valve in the secondary containment portion of the Reactor Water Cleanup (RWCU)
(1) the ratio of the licensee's value to the NRC value is computed Licensee Value (ratio = HRC Value );
system f t 51ed to close. No alarms sounced initially, but parametric values indicative of a 0.1 square f oot breach of the primary system were given to the operator This condition was intended to initiate a problem solving mcde (i.e. an event based evolution) prior to any E0P entry ;
(2) the uncertainity of the ratio is propagate If the absolute value of one minus the ratio is less than or equal to twice the ratio uncertainty, the results are in agreemen (ll-ratio l s 2 uncertainity)
c/ dition. Area high temperature, and high radiation alarms were used to )
Z= 5' then bZ = SI + by y Z2 x2 y8
e. ;ablish secondary containment and radiation control E0P entry with feergency Plan actions. Secondary containment conditions were severe enough that emergency depressurization was required to minimize the radioactive release. Rapid plant shutdown and cooldown proceeded with all l plant components operating normall l-10-
* ( From : Bevington, P. R., Data Reduction and Error Analysis for the Physical Sciences, McGraw-Hill, New York, 1969)
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c. Observations and Conclusions The operators' cooperation under the difficult simulation circumstances involving the control room mock-up was excellent. Cornmunications between crew members during the evolution were clear, and overall response to the scenario situations resulted in moving through a very complex scenario without any significant procedural errors. The scenario timeline was maintained with a deviation of less than five minutes at the one hour point, which resulted in a realistic event sequence. Under these real time conditions, the useability of tne E0Ps was demonstrated to be satis-factory. The shif t from event based (i.e. abnormal operating parameters)
to condition based E0Ps in the second scenario was made smoothly, com-pletely and without hesitation, as was the transition from secondary containment centrol to the emergency depressurization sequence. Although no significant procedural errors were evident, the following concerns were identified by the tea . Placekeeping Method - The team was concerned with the method utilized for placekeeping curing the performance of the E0Ps. The operators used multiple loose leaf copies of the procedures to follow the multiple paths required. Through direct observation and interviews, the team determined that loose leaf procedures would not be used in the control room. The control room copies are of a "lay flat" design and two copies of the E0Ps are maintained to support multiple entry conditions. Although the operators did not lose their place in the procedure during either scenario, the team was concerned that the Plant Specific Writer's Guide, Operating Department Standing Order s, operator training and the E0Ps themselves did not provide or support a preferred method of placekeeping. Further action should be taken to evaluate and identify a preferred method of placekeeping. This method should be procedurally supported and trained on a periodic basi . Use of Cautions - The team was concerned that at no time during the l performarice of the scenario were the operators observed to review F-EOP-1, "EOP Cautions." This E0P contained 23 Cautions, whicn were '
referenced throughout the E0Ps. Specific reference to these cautions were made by numeric and abbreviated reference. Althougn none of the l specific requirements of this E0P were violated or overlooked during l the scenario demonstrations, the team remained concerned that the collection of all cautior.s into one location outside the normally performed flow path of the E0Ps could result in overlooking a significant cautio . Minimum Snif t Staf fing - Corwnunications outsioe the control room Tdispatener, plant o.anagement, NRC, and Emergency Plan notifications)
were not simulated in the first scenario due to the circumstances of the control room mock-up. Corrcunications were adequately simulated in the second scenario, and crew communication was excellent. The Emergency Plan was not available at the simulator, but was called for by the operators. The E0P simulation demonstrated that the minimum shift crew could imGement all steps nf the E0Ps. However, the team was concerned that sufficient control room personnel might not be available to concurrently perform all required actions, including-11-
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l implementation of the Emergency Plan and activation of the Fire Brigad l The operatnrs confirmed that their.first responsibility was to perform the E0Ps and that additional actions would be required to be performed by personnel not specified in the minimum shift _ crew (i.e. ;
t  t t t t t t n n  n n n n n n o e  e e e e e e s mtt ttt ttt ttm mtt mtt mtt mtm t t t t i
security guards). In addition, the operator interviews (discussed in l Section 6.2.a) indicated that the operators believed that the present staffing levels were adequate. The team remained concerned that all actions required to be performed in an emergency would not be able to be accomplished by tne minimum shift crew defined in Technical Specifications. Further licensee and NRC action is required to resolve this concer . HUMN FACTORS REVIEW AND INTERVIEWS In order to determine the adequacy of the E0Ps witn respect to the guidance provided in NUREG-0899, "Guidelines for the Preparation of the Emergency Operating Procedures," a review of the Plant Specific Writer's Guide (PSKG) and tne E0Ps was performed to determine the extent to which the PSWG nas been implemented. In addition, structured interviews were conducted with relevant JAFNPP personnel. The results of these efforts are detailed belo .1 Writer's Guide Implementation Administrative Procedure F-AP-2.2, "Writer's Guice," was reviewed to ensure that the human factor's guidance provided was incorporated during the develop-ment of the E0Ps. Two specific strengths and four weaknesses were noted: Paragraph 4.3, page 27, provided concise, distinctive specifications for the content of Cautions and Notes which was similar to the guidance of paragraph 5.7.9 of NUREG-0899. This application of human factors principles was used by JAFNPP to transform Cautions #4, 5 and 10 of the EPGs into either Notes er Action statements in the E0Ps. This reduction in the total number of cautions creates a reduction in the eperator's buroen and is considered an improvement in the clarity of the E0P Tne use of miniature figures within the body of the E0Ps was identified as an innovative method to minimize branching outside the procedure withcut reducing the technical adequacy of the references. As identified in Section 6.2.e, the operator interviews confirmed that whenever extrapola-tions were required, the full sized figures attached to the E0Ps were use l JAFhPP's response to comment B.1 of tne Interim Safety Evaluation dated l September 11, 1985 states that an action step will be completed on tne l page where it begins. A review of tne E0Ps identifleo that operator '
enn nnn nnn nne e n enn enn ene n  n n n r eee rmm eee mmm eee eee eee eee eee eee e  e e e eee mmm mmr gee rmm gee rmm rmm rmr mgeg a        m m m p gee eee eeg gee    e e e e m
action statemer,ts were oftEN Continued from one page to another witn no consister4 in format. This practice sometimes resulted in part of a l logic statement on the first page and the remainder on the second pag '
o aee srr rrr eeerrr eeerrs eea srraee srraee srr aee srs aear e  e r
For example, F-E0P-4, step 4.4 (pages 21-23), step 4,6 (pages 23-25) and step 5.3.1 (pages 31-35) each contained examples or three different formats. Beseo on the response to the safety evaluation and t;ie operator ccncerns retarding continuity and placekeeping identified in Section-12-
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C igg ggg ggg ggi igg igg igg i gi g g g g DAA AAA AAA AAD DAA DAA DAA DAD A    A A A
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- ) 232 873 111 652 87 077 522 033 8  5 3 3 C 000 000 111 000 110 100 010 110 0 2 0 0 oR  .. . .
iN 000 000111000 000 000111000 000 000 0  0 0 0 ta /. 111 850 259 856 111 533 112 892 723 111 11t 874 121 527 7


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6. the Writer's Guide shculd be revised to provide a human factored format for the continuation of action statements to the following page when completion of the action statement on the first page is not possibl In addittoa, the incorrect response to the SER comment should be identifieo and resolve . Paragraph 4.8, page 29, specified that capitalization will be use'd for emphasis in specifted instance A review of the E0Ps indicated that the term "upper case type" is prob'bly meant instead of "capitalization" in the Writer's Guide and only the action statements within the contingency statements were to be in upper case typ Tne Writer's Guide should be clarified to provide consistent guidance, in addition, Caution No. 22 of F-EOP-1, "EOP Cautions," should be changed to conform to the requirements of paragraphs 4.2 and 4.8, such that the "if" logic term is in upper case and is located at the beginning of the logical conditio Paragraph 4.6, page 28, concerning the referencing and branching to other procedures or steps, provided no guidance as to how referencing or branching to Abnormal Operating Procedures (A0Ps) or Operating Procedures (ops) should be handled. The review of the E0Ps identified that the prac-tice utilized within the E0Ps was not consistent, in that the A0Ps or ops were not always referenced by both procedure number and nam NUREG-0899, paragraph 5.2.2 indicates that the specific system procedures should be referenced in the E0P Paragraph 4.7, page 26, indicated that the equipment names referenced in the E0Ps may not always match the engraved names on the panels, but will be complete and in operator language. Althougn there were several minor examples identified ouring the plant walkdcwns in which equipment names did not match their labels, on-going labelling efforts, identified in Section 4.e. are anticipated to correct this discrepancy. Further action should be taken to revise the Writer's Guide to reflect the current labelling philosoph .2 Operator Interviews interviews 'were conducted by the human factors member of the team with individual' members of the plant staff as classified below:
ifob Classification (License)  Number
  $nift Supervisors (SRO)  2 Asst. Shift Supervisors (SRO)  3  t Senior Nuclear Operators (RO)  1 Nuclear Control Operator (k0)  1 Auxiliary Operator (unlicensed)  1 Training Coordinator (SR0 certifieo) 1 l
A four page interview guide with 8 major topics was used for each interview and l was reviewed by both parties. Discussions were open-ended, in that the licensee representative was encouraged to volunteer comments which were relevant. Each person was advised that the objective of the interview was to develop information on the effectiver.ess of the E0Ps and not to examine th qualifications of the individual. The length of the individual interviews were-13-l
 
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approximately one hou Two majo changes were in progress at JAFNPP which the operators anticipate will have a positive impact on the effectiveness of the E0P The first is the stort-up of a site specific training simulator late in 1988 and the second ;s the change to a flow chart E0P format in 1989. However,
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I the interviews were confined tc the context of the presently implemented E0P The results of this process are identified belo Role / Task Definition - There was an established, uniform practice for the conduct of plant operaticns during the execution of E0Ps. These operations were governed by Operations Department Standing Order No. 2,
"Operating Principles ano Philosophy," and further developed during control room crew team training during simulator and on-the-job training (0JT). Ef fective execution of the E0Ps was also aided by crew stability (typically crew membership has been the same for 3 years) and by a recently initiated training program on effective oral communications. The control room task assignments were well defined. The consensus of opinions was that the current control room staffing level achieved the proper balance between assuring adequate staffing and avoiding confusio At JAFNPP an off duty control room crew was assigned to "standby" and may be called to the site upon the declaration of an Unusual Event. However the avail 6bility of off duty personnel does not resolve the team's concern with the minimum shift crew manning as discussed in Section 5. Use of E0Ps - The control room resources for use of the EUPs were considered adequate. Two sets of E0Ps were kept ir he control room. A space was assigned for the lay-down of the "open-fl ' E0P manuals. A cart was provided for abnormal and normal operating procedures, Technical Specifications, Emergency Plan, et Technical Adequacy - Concerns were expressed about the ability to reliably execute, under accident conditions, Step 5.2 of F-EOP-4, "Primary Contain-ment Control," which required the combined use of Figure F-E0P-4.1, "Heat Capacity Temperature Limit," and Figure F-E0P-4.7, "Heat Capacity Level Limit." There was no explicit direction in Step 5.2, or in the E0Ps overall, as to how to obtain the value of the abscissa for Figure F-EOP- Paragraphs 5.6.9 and 5.7.8 of NUREG-0899, "Guidelines for the Preparation of Emergency Operating Procedures," provide guidance which suggests that step-by-step direction should be added to F-EOP-4, Step 5.2, for the proper combined use of the two curves. This direction would improv[ethehumanreliabilityassociatedwiththeoperationofthesafety systems affected by Step Concerns were expressed about the ability within the time available to bypass the low RPV water level Main Steam Isolation Valve (HSIV) isolation interlocks in accordance with F-A0P-38, "EOP isolation / Interlock Overrides," when directed by step 2.2.2 of F-E0P-3, " RPV Control (Boron injection Required)." The operators suggested that F-EOP-3, "RPV Control (Boron Injection Required)," should specify controlling RPV water level at e control pcint above the automatic MSIV isolation setpoint to prevent unnecessary isolation of the RPV (and loss of the ultimate heat sink)
during the installation of override jumper Use of Cautions - The responses of the persons interviewed indicated that l the first three cautions of F-E0P-1, "EOP Cautions," were not readily !
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000 099 890 999  391 619 200  411 9 2 9 9 o 111 100 001 000  101
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recalled and were therefore not fulfilling their intended functio F-EOP-1, contained all cautions applicable directly to the E0Ps. The cautions were listed by serial number and consisted of a title ano tex Numbers and titles of the cautions were used throughout the E0Ps when reference to the specific caution was required, however the first three cautions were applicable to all E0Ps and were consequently not referenced within the texts of otner E0Ps. These cautions required the operators tc monitor overall plant conditions, nicnitor multiple indications and to confirm safety functions of automatic equipment. This concern is not to st6te ''that the requirements of the cautions would not be applied by the operators in the performance of the E0Ps, however the effectiveness of the use of a separate volume of cautions (as discussed in Section 5.b.2) is questionabl The effectiveness of the E0P Cautions would be enhanced if the texts on some of the associated figures were more explicit. For example, label the cross-hatched areas as "Prohibited Region," or label the figures associa-ted with Caution # 8 with a directly worded caution such as "Do not operate pumps unless..." instead of the non-specific, "Observe NPSH Limits...".
n io l
There was a general concern among the operators over the number of cautions. The PSTGs must be observeo with respect to the incorporation of tecnnical restrictions into the E0Ps, however application of human factors guidelines has reduced the burden on the operator of several technical restrictions (as discussed in Section 6.1). As previously identified, Cautions # 4, 5 and 10 of the EPGs were not incorporated as caution statements into the E0Ps because they did not meet the criteria for cautions in NUREG-0899, paragraph 5. Similar further application of the criteria of accuracy, conciseness, and consistency, could reduce the impact of a large number of cautions, e. Miniature Figures - The interviews confirmed that the use of Miniature figures within tne E0Ps were used by the operators to the extent that the figures can be safely interp-eted. If the small size of the figures ,
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caused any doubt, the full size figures of the E0Ps were used. The
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miniature figures did not appear to create any additional possible error mechariism and the operators considered them useful aids, l
f. Need for a Basis Document - The interviews identified the operator's desire to add the basis for the opEdating limits of the E0Ps into the l procedures. Although the addition of this information would generally '
interfere with the clarity of the operating instructions and is therefore rot recommenced, the need for this information in some form is apparen The basis of the operating limits were supported by operator training, nowever the identification of a neea for a "basis document" underscores the requirements (identified in Section 3.a) for a reference document which is traceable ar,d maintained up-to-date, g. Transitions and Placekeeping Methods - The interviews indicated that the metnoa of hanaling transitions witnin the E0Ps as well as place-keeping within the E0Ps and the reliability of the metnods utilized has been receiving considerable attention by the JAFNPP staff. A standard method of placekeeping has not been developeo or trained. Each crew was free to-15-
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lb a
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  ,.3 t00 000 92  1 4 5  5 011 000 611 681 212 512 4 1 0 1 T A V r i11 111 11 111 111 211 111 111 1 1 1 1 L L e 072 027 108 076 302  109 641 146 3 6 3 7
implement its own method of placekeepin The use of separate binders and color ccding of the E0Ps, as well as the "lay flat" capability were the operator aids in us Based on the suggestions for additional methods, the activity involved to improve tnis ability, and the concerns identified in Section 5.c.1, further consioeration should be given to revising the E0Ps to include support for check-off spaces and the adoption of- a practice of writing directly in the action copy of the E0P Communication - The interviews indicated that the present communications methods were adequate both within the control room and to the local stations, but that a recent program for improving communications w;s appropriate and productiv Expectations for improvements in communications training due to the use of the site specific simulator were !
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also note I Contro! Rocm Environraert - The interviews identifieo that the provisions l for lignting witnin the control room and at local stations in the event of a station blackout and the provisions for control room habitability in the event of an on-site raoiation release or in the event of smoke or toxic j gasses in the control room have been difficult to incorporate into 0JT l training and that further training with the site specific simulator should I be pursue ' Balance of Plant / Local Control Stations - The interviews identified the i use of an "operator aids" program wnicn produced validation and  ;
U P . p  31  820 446 728  830  7 S
improvement of the human f6ctors at local stations including significant i efforts to upgrade the labelling of components. In summary, the implementation of E0P operations at local control stations was considered adequat Validation and Verification - The interviews indicated that the operators were includeo to a limited extent in the process of developing, verifying, and validating the E0Ps. Operators were aware that there was a procedure for initiating suggested upgrades to the E0Ps and that training exercises were expected to help identify possible discrepancies. However, operators were unclear as to their personal responsibilities for initiating resolutions of possible discrepancies as evidenced by several instances in which potential discrepancies identified by the operators in the interview had not been formally identified by the operator for resolution. Further action is necessary to clarify the operator's role in E0P upgrade . CONTAINMENT VENTING The team reviewed the EPCs and the Appendix C, Calculation Procedure No. 14,
E R cL 098 1 13 088 5 6 594 12 11 11 12 8 7 1 R A 4,1 l_
  "Primary Containment Pressure Limit," to determine if the PST6 values were computed correctly. The team also reviewed tne method, flow path, and feasib1-lity of the containment venting procedur .. .
113 1 1    5  5 7
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  - 1 T E
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a S C p E E U L 1 N n  0 8      _
The attributes of the vent paths are detailed in Calculation Procedure No. 14
B  ei 26 742 769 242 955 815  5 555 9 8 5 A Y K u 1        5 _
"Primary Containment Pressure Liniit," and are summarized as follows:
T T I
Suppression Chamber Vent Paths
C I
      ,
l a
Path  A(v?) P(oi) P(ci) E(vi) --
s t o11 00 011 111 000 111 000 044 111 111 127 111 5.1 3 114 0 1 1 4 _
27-A0V-117,118 0.736 7 .3 2 MOV-117,123 0.021 5 .0 2 Drywell Vent Paths Path  A(vi) P(oi) P(ci) E(vi)
111 111 1 1 1 1 L R Vl t73 820 568 847 090 I T u    915 1 _
27-A0V-113,114 0.697 .2 105 27-MOV-113,122 0.021 5 .0 105 A(vi) - Minimum vent path area (f t2)
B A
P(oi) - Maximum containment pressure the vent valve can open against(psi)
A P
P(ci) - Maximum containment pressure the vent valve can close against (psi)
C s R e 3.8 8 999 0.0 5 2.0 5 000 051
E(vi) - Elevation of the vent patn containtnent penetration referenced to torus bottom (ft)
      .
From the above information, JAFhPP determined that the suppression chamber vent path via A0V-117 and A0V-118, is the only path capable of operating at the design pressure of the containment and meeting the criteria of removing decay hea Venting of the containment was controlled by F-A0P-35, "Post Accident Venting of the Primary Containment." Two flow paths are possible: a small bore path and a large bore path. Containment venting would be initiated via the small bore path, and if not effective in restoring pressure to less than the limit, venting would be continued througn the large bore path. The initial path is via valves MOV-117, M0V-123 and M0V-121. Due to piping and valve size and a flow restricter between valves MOV-117 and MOV-123, this flow path will not control the containment pressure under accident condition The large bore path is via valves A0V-117, A0V-118, and MOV-120. Valves A0V-117 and A0V-116 are 20 inch valves that discharge tc a 30 inch carbon steel header. Valves M0V-121 anc h0V-120 are 6 inch and 12 inch valves respectively which discharge in parallel through a 24 inch carbon steel header to one of two Stacdby Gas Treatment Systein (SBGT) trains. Tne SBGT trains are tested at 1 psig, and have a working pressure of 0.5 psig. The vent patn starts near the top of the torus. When torus level increases to 29.5 feet, the vent path becomes unusable due to flooding. An alternate path from the drywell air space is not available under hign pressure conditions because isolation valves in
4.5 1 595 0  0 9 5 913 13 611 511 628 628  91 517 6 6 $ 7 P Z NR A T 11 11 51 11  1 C l f
" potential vent paths are not able to be stroked (closed) under high differential pressure condition The team confirmed that figure F-E0P-4.6, "Primary Containment Pressure Limits," was a result of considering the maximum constant pressure condition during the air purge and during tne steam vent. During air purge, constant-17-
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        *
pressure is the result of volumetric air flow out of the vent being equal to ;
2    .
the volumetric addition of steam generated by decay heat. During the steam I vent, the steady-state condition exists where energy out of the vent is equal i to energy generated by decay heat. The limiting structural component in the primary containment is the 48 inch manway to the torus. For any conditions of vacuum that may develop in the torus or drywell, vacuum breakers are-provided between tne torus and the drywell and between the torus and the reactor buildin Calculations were not included as part of the EPG calculations to determine the pressure which the SBGT system would be subjected t under venting condition The SBGT filter units are located in an enclosure adjacent to the reactor building and 1solated from the environment by a non-seismically qualified door with ventilation louvers. Without further procedura l precautions or hardware modifications, it is possible that the SBGT train would rupture due to the high pressure steam being vented. In the event of a f ailure, the vent path would then release into the environment via this unmonitored path througn the SBGT room. Further evaluation is required to ensure that the SEGT train is not anticipated to rupture under the postulated pressures and temperatures associated with the containment venting sequenc The vent paths discussed above pass through readily iccessible portions of the secondary containment. Although venting would result in increased radiation levels, the team concluded that the operators could carry out other duties simultaneously with ventin F-A0P-35 woulo be clearly implemented by control room operators without further direction after the accident mitigation strategies of the E0Ps have faile ;
          *
The operators were directed by the procedure to vent, "... irrespective of radioactive release." Appropriate cautions were included concerning implementation of the Emergency Plan ano Dose Assessmen The team identified one concern with the implementation of the containment venting procedures, in that a Special Procedure, F-SP-02, "Post LOCA Venting of Containment &
TABLE 1 CAPABILITY TEST RESULTS FITZPATRICK NUCLEAR PLANT Chemical  Analyticas  Ratio Pa ra me te r Procedure , NRC value Lic. Value (Lic./NRC) Compa ri son Results in carts eer million f oom)
Operation of the Main Steam Leakage Collection System," was identified to be an active plant procedur The vent paths described in F-SP-02 were identical to those specified in F-A0P-35, however the initiation pressure for containment venting was approxim6tely 45 psig lower than the pressure specified in F-A0P-35. The licensee indicated that the containment venting portion of this '
Bo ron T i t ra t ion '.040tto 1038t29 Ag reemen t 31002100 3009163 0.9710.04 Ag reement
procedure was based on old event-based operating procedures and was l inadsertently issued during the last revision of the section applicable to the operation of the Main Steam Leakage Collection System. During the inspection period, F-SP-02 was withdrawn and revised to remove tne non-applicable portions 1 of containment ventin l ON-G0ING EVALUATION OF E0Ps hUREG-0899 Paragraph 6.2.3, indicates that licensees should establish a program for on-going evaluation of the E0Ps. This program should include: evaluations cf the technical adequacy of the E0Ps in light of operational experience and ,
    '>000190 49 % i30 0.9920.02 Ag reement
use, training experience, and any simulator exercises and control room '
        ,
walk-throughs; evaluation of the organization, format, style and content as a result of using the procedures during cperations, training, simulator exercises, and walk-throughs; and evaluation of staffing and staff qualifications relevant to using the E0P I-18-
*See report text Section _ - _ . . _ _ -  _ _ _ - _
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o s .
 
l
 
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The team reviewed the Administrative Procedures which control the use of  l procedures at JAFNPP. F-AP-1.4, "Control of Plant Procedures," established the
      )
requirements for initiation, review, approval, revision, temporary change,  I withdrawal, and control of procedures and was applicable to all operational i procedure F-AP-1.2, "Plant Operating Review Comiittee," specified a schedule for periodic review of procedures, which included a biennial review of EOP Plant Standing Order No. 28, described the procedure by which internally and externally generated operating experience is evaluated, reviewed and, if necessary, incorpcrated into plant procedures or design changes. Although each of these progran.s or procedures was considered to be applicable to tne E0Ps, a specific program for the on-going evaluation of the E0Ps did not exist. As a future upgrade of the E0Ps, the licensee is in the process of upgrading the E0Ps to Revision 4 of the Vendor's EPGs and has scheduled verification and implementation of flow chart E0Ps by July 198 . EXIT MEETING The inspection team conducted an exit meeting on June 3,1988, with licensee management to identify the inspection fir. dings and provide the licensee with an opportunity to question the observation The scope of the inspection was discussed and the licensee was informe' of the conclusions identified in the course of the inspection. Mr. C. J. Haughney, branch Chief, Special Inspec-tions Branch, NRR, and Mr. Jon Johnson, Section Chief, Division of Reactor Projects, Region I represented NRC management at the final exit meetin . REFERENCES 10.1 Personnel Contacted A large number of personnel were contacted during the inspectio The following is a list of the JAFNPP personnel involved:
  *R. Converse, Resident Manager
" Fernandez, Superintendent of Power
*D. Lindsey, Operations Superintendent
  *0. Burch, Reactor Analyst Supervisor
*R. Patch, Quality Assurance Superintendent
*V. Walz, Technical Services Superintendent
*0. Simpson, Training Superintendent
*J. Catella, liuclear Training Manager P. Brozenich, Shift Supervisor R. Pike, Asst. Shif t Supervisor G. Davis, Reactor Operator L. Shaffer, Reactor Operator G. Frank, Training Department D. Jonnson, Waste fianagement General Supervisor J. Lazarus, Assoc. Plant Engineer K. Moody, Plant Engineer    .
0. Ruddy, Plant Engineer Supervisor    l G. Tasick, Quality Assurance Supervisor    l B. Robinson, Quality Assurance Engineer    l J. Prokop Jr., Quality Assurance Engineer    .
O. Squires, Shitt Supervisor    '
    -19-
  .-.. _
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e r, o
*
.
I l
R. Thomas, Assistant Shift Supervisor W. Hendricks, Reactor Operator
  * Denotes those present at the Exit Meeting on June 3, 198 .
10.2 Procedures Reviewed    _
F-EOP-1, "E0P Cautions," Revision 3 F-EOP-2, "RPV Control (Boron Injection Not Required)," Revision 1 F-EOP-3, "RPV Control (Boron Injection Required)," Revision 1 F-EOP-4, "Primary Containment Control," Revision 1 F-EOP-5, "Secondary Containment Control," Revision 1 F-EOP-6, "Radioactivity Release Control," Revision 1 F-E0P-7, "RPV Flooding," Revision 1 F-AP-1.4, "Control of Plant Procedures," Revision 7 F-A0P-15, "Recovery from an isolation," Revision 9 F-A0P-33, "Alternate Shutdown Cooling," Revision 0 F-A0P-34, "Alternate Control Rod Insertion," Revision 0 F-A0P-35, "Post Accident Venting of the Primary Containment,"
Revision 0 F-A0P-36, "Stuck Open Relief Valve," Revision 3 F-A0P-37, "Boron Injection Using the CRD System," Revision 0 F-A0P-38, "E0P Isolation / Interlock Overrides," Revision 3 F-A0P-43, "Plant Snutdown from Outside the Control Room,"
Revision 8 F-AP-1.2, "Plant Operating Review Committee", Revision 4 F-AP-1.4, "Control of Plant Procedures", Revision 4 F-AP-2.2, "Procedure for Emergency Operating Procedures,"
Revision 5 F-SP-2, "Post LOCA Venting cf Containment & Operation of the Main Steam Leakage Collection System," Revision 8 JAFNPP Records Retention / Turnover Schedule, Revision JAfhPP Emergency Procedure Guide, Revision 3 0050-2, "Operating Principles and Philosophy," Revision 3 0050-4, "Posting of Operator Aids," Revision 1 P50-4, "Quality Assurance & Plant Operating Records," Revision 3 l
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Latest revision as of 04:54, 12 December 2020

Insp Rept 50-333/88-20 on 880919-23.No Violations Noted. Major Areas Inspected:Nonradiological Chemistry Program, Including Measurement Control & Analytical Procedure Evaluations
ML20205Q355
Person / Time
Site: FitzPatrick Constellation icon.png
Issue date: 10/21/1988
From: Kirkwood A, Pasciak W
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20205Q353 List:
References
50-333-88-20, NUDOCS 8811090203
Download: ML20205Q355 (8)
v • d • e


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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report N /88-20 Docket N License No. DPR-59 Priority Category C Licensee: Power Authority of the State of New York P. O. Box 41 New York, New York 13093 Facility Name: James A. FitzPatrick Nuclear Power Plant Inspection At: Lycoming, New York Inspection Conducted: September 19-23, 1988 Inspectors: / to ct N d[ O/,/hIh A. Kirkwood, Radiation Specialist date r

Approved by: M a. _ c '/ 2019.99'

W. J. Pasciak, Chief, Effluents Radiation dat'e Protect.on Section Inspection Summary: In;pection on Se.)tember 19-23, 1988 (Report N /88-20)

Areas Inspected: Routine, announced ins; ion of the non-radiological chemistry program. Areas reviewed included measuremar t control and analytical procedure evaluation Results: No violations were identifie '

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SG11090203

"DR GS1008ADOCK 05000333 FDC Q

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CETAILS

, Individuals Contacted  ;

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  • R. Converse, Resident Manager i G. Tasick, QA Supervisor
  • W. Fernandez, Superintendent of Power .

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*E. Mulcahey, Radiological and Environmental Service Supervisor
  • A. McKeen, Chemistry General Supervisor

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W. Hamblin, Chemistry Supervisor

  • C. Boucher, Chemistry Supervisor ,

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D. Johnson, Waste Management Supervisor i O. Lindsay, Operation Superintendent .

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3 Analytical procedures Evaluttion

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During the inspection, .tandard chemical solutions.were submitted by

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the ins 1ector to the licensee for analysis. The standard solutions were prepared by CruoKhaven National Laboratory (BNL) for the NRC Region I, and were analyzed by the licensee using normal methods and equipment. The analysis of standards is used to verify the licensee's capability to  !

. monitor chemical parameters in various plant systems with respect to  ;

Technical Specifications and other regulatory requirements. In addition, the analysis of standards is used to evaluate the licensee's analytical procedures with respect to accuracy and practsio ,

J The results of the standard measurements comparison (Table 1) indicated l

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that seven out of thirty-one measurements were in disagreement unust the  !

1 criteria used for comparing results (see Attachment 1). However, by 1

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redilution of the NRC standards, preparation of new standards calibrai. ions, !

3 and using different technicians to do the same analysis, all measurements ,

came into agreement during subsequent analyse !

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l The chloride disagreement, at approximately 10 ppb, was due to limitations of the statistical method for determining agreement. The actual Lic/NRC ,

i ratios differed by only four percent (4%). Similarly, the silica Lic/NRC I

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ratios differed by less than ten percent (10*4). This was resolved and {

came into agreement on another analysis, i

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Copper, nickel, chromium and iron disagreements were apparently due to

calibration of the inductive coupled plasma analyzer (ICP). Subsequent j analysis after recalibrations brought all samples into agreement.

i Procedures ware followed in all instances.

I No violations were noted.

. Measurement Control Evaluation f

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Verification of the licensee's measurement capabilities on actual plant water samples is done by splitting samples with the licensee and Brookhaven National Lab (BNL). Samples were taken to be sent to BNL for

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independent verification. The reactor building closed cooling water (RBCCW) was sampled. One sample was spiked with a standard solution of anions and split with the licensee and BNL. The second sample was not spiked and was also split with the licensee and BNL. The standard spike i solution was prepared by-BNL for the NRC Region I. On completion of the analyses by BNL and the licensee, an evaluation will be mad . Quality Ast.urance/ Quality Control Program

The inspector performed a selected review of the licensee's program for the quality control of analytical measurements made by the chemistry i department. The review was performed, in part, with respect to criteria

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contained in the following:

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Principles of Quality Assurance of Chemical Measurements, National

Bureau of Standards -

Regulatory Guide 4.15. "Quality Assurance for Radiological Monitoring

Programs (Normal Operations) Effluent Streams and the Environment

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The following licensee procedures were reviewed to determine the adequacy ,

of measurement control: '

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Quality Assurance / Quality Control Program, dated 7/19/88 l l

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CLI-17, Dionex Ion Chromatograph and Hewlet Packard Reporting  ;

Integrator, dated 7/19/88

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CLI-34, Routine Verification and Calibration of Laboratory Pipettes, dated 8/19/83 4 ,

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CLI-35, Plasma Spec 2-5 leeman Labs Operation and Calibration, dated i 9/4/87 *

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! * CAP-19, Silica Determination, dated 8/5/88

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} Also the following records were selectively sampled to determine the  !

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extent of measurement control: l

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QA/QC Data Log from 1/14/88 to 9/22/88, Intra and Interlab l Cross-check results i

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Non-Conformance Reports from 3/17/88 to 9/18/88, both completed and g pending '

  • Aiministrative Review, AM-88-01 for week of 8/4-12/88

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  • Nanopure Con D H O conductivity measurements from 3/24/88 to 9/22/88

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(daily checks)

Reactor Water Data Sheets from 1/1/88 to 7/31/88

  • ICP H2 Cooler TBCLC side sludge sample analysis dated 9/21/88
  • Control Charts for Cr, Fe, Cu, Ni dated 9/13/88

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The inspector noted a few minor problems with the QA/QC program procedure

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of 7/19/88. First, Section 5.6.7, "Use of Analytical Standards," allows l the preparation of standards from the same lot. Thi., would not allow verification of standard quality and sample results. Also, in Section  ;

5.11, "Alternate Analytical Labs," the NRC is listed as an alternate lab  !

that can be used for indepe.: dent sample splits. The licensee cannot fulfill its QC Prcgram Sample Split Schedule, for annual independent .

lab checks, since they cannot be assured of NRC splits at this frequenc : The NRC should be removed from this list.

3 The QA/QC program is otherwise well documented and improvements in measurement control are evident. Analyses are completed in a timely fashion. Trends are easily detected on control charts and out-of-l

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specification situations are followed up and investigated promptl Corrective actions are technically sound and retraining of technicians, where needed, was noted. A thorough chemical contaminant control program was detailed to the inspecto No violations were noted.

- Management Controls 5.1 Facilities and Equipment  ;

Expanded laboratory space and improvements in the work environment  ;

were noted since the last inspection. A storage area has been converted to an analysis area for metals and anions, creating more

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workspace in the lab proper. Also, a dropped ceiling and new paint i provide a brighter more professional work area. The ICP (inductive  ;

coupled plasma) analyzer is fully operationa '

5.2 Staffing l

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Staffing appaars adequate for meeting outage requirements. A special

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chemical decontamination of recirculation lines had required the

, staff to work six, twelve hour days each week. This is the maximum allowed under the licensee guidance. Seven, ten hour days will be resumed for the remainder of the outage. The inspector noted an alert, technically competent group of eight fully qualified techni-cians and two first line supervisors, with many years of experience

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_ . , _ , - - , _ _ _ , , ~ _ ~ . . _ _ _ , . - _ . _ _ . . . .-,% .-.m, w-_ .g .- ,__,_.,s. ..m.,e p ,-~,,m- -- - , , . , . , . _ _ _ , .

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among the Two trainees and two contractors supplement this core of employee First line supervisory duties are transitioning such that one individual will handle radiological chemistry and the other will handle non-radiological chemistry. Both first line supervisors report to the Chemistry General Supervisor; who is responsible to the Radiological and Environmental Service Supervisor; who in turn reports to the Superintendent of Power and then to the Resident Manage The long-term, technically competent, staff is viewed as a positive program asset, as is the cooperation and attitude noted by the inspecto No violations were foun . Exit Interview The inspector met with the licensee's representatives (denoted by an asterisk in S(ction 1) at the conclusion of the inspection on i September 23, 1988 and summarized the scone and findings of the i inspectio ..' _,

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ATTACHMENT 1 CRITERIA FOR COMPARING ANALYTICAL MEASUREMENTS This attachment provides criteria for comparing results of capability test In these criteria the judgment limits are based on the uncertainity of the ratio of the licensee's value to the NRC value. The following steps are performed:

(1) the ratio of the licensee's value to the NRC value is computed Licensee Value (ratio = HRC Value );

(2) the uncertainity of the ratio is propagate If the absolute value of one minus the ratio is less than or equal to twice the ratio uncertainty, the results are in agreemen (ll-ratio l s 2 uncertainity)

Z= 5' then bZ = SI + by y Z2 x2 y8

  • ( From : Bevington, P. R., Data Reduction and Error Analysis for the Physical Sciences, McGraw-Hill, New York, 1969)

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t t t t t t t n n n n n n n n o e e e e e e e s mtt ttt ttt ttm mtt mtt mtt mtm t t t t i

enn nnn nnn nne e n enn enn ene n n n n r eee rmm eee mmm eee eee eee eee eee eee e e e e eee mmm mmr gee rmm gee rmm rmm rmr mgeg a m m m p gee eee eeg gee e e e e m

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iN 000 000111000 000 000111000 000 000 0 0 0 0 ta /. 111 850 259 856 111 533 112 892 723 111 11t 874 121 527 7

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2 .

TABLE 1 CAPABILITY TEST RESULTS FITZPATRICK NUCLEAR PLANT Chemical Analyticas Ratio Pa ra me te r Procedure , NRC value Lic. Value (Lic./NRC) Compa ri son Results in carts eer million f oom)

Bo ron T i t ra t ion '.040tto 1038t29 Ag reemen t 31002100 3009163 0.9710.04 Ag reement

'>000190 49 % i30 0.9920.02 Ag reement

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