IR 05000312/1986038: Difference between revisions

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{{Adams
{{Adams
| number = ML20210M567
| number = ML20212K950
| issue date = 01/12/1987
| issue date = 02/27/1987
| title = Insp Rept 50-312/86-38 on 861117-21 & 1208-23.Violation Noted:Failure to Include Appropriate Acceptance Criteria for Snubber Lockup Velocity
| title = Ack Receipt of Informing NRC of Steps Taken to Correct Violations Noted in Insp Rept 50-312/86-38
| author name = Clark C, Melfi J, Richards S, Wagner W
| author name = Kirsch D
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
| addressee name =  
| addressee name = Ward J
| addressee affiliation =  
| addressee affiliation = SACRAMENTO MUNICIPAL UTILITY DISTRICT
| docket = 05000312
| docket = 05000312
| license number =  
| license number =  
| contact person =  
| contact person =  
| document report number = 50-312-86-38, GL-85-05, GL-85-20, GL-85-5, IEIN-85-042, IEIN-85-42, IEIN-86-056, IEIN-86-063, IEIN-86-56, IEIN-86-63, NUDOCS 8702120575
| document report number = NUDOCS 8703090469
| package number = ML20210M507
| title reference date = 02-23-1987
| document type = INSPECTION REPORT, NRC-GENERATED, INSPECTION REPORT, UTILITY, TEXT-INSPECTION & AUDIT & I&E CIRCULARS
| package number = ML20212K952
| page count = 26
| document type = CORRESPONDENCE-LETTERS, NRC TO UTILITY, OUTGOING CORRESPONDENCE
| page count = 1
}}
}}


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U.'S." NUCLEAR REGULATORY COMMISSION
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"* Report N /86-38
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Docket N ;  .    .  "
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License N DPR-54
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Licensee: -Sacramento Municipal Utility District  '
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P. O. Box 15830-    .
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Sacramento, California 95813;        ;
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  = Facility:Name: Sacramento Municipal Utility District (SMUD)
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Inspection Condtieted:  November 17-21 and December 8-23, 19867 Inspected by: [M   ,     ' / - 8-- 8 2 -
C. Clark, Reactor Inspector      Date Signed M    .
b Y-h E Melfi, Reactdr Inspector
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          . Date Signed k?d. // lasani      / - 7-B7 Date Signed'
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W.pner,Reac 'r~faspector .    ,
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,  Approved by: . / - /2- 97-S. Richards, Chief, Engineering Section    Date Signed
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Summary:            ,
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Inspection on November 17-21 and December 8-23, 1986 (Report No'.' 50-312/86-38)-
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Areas Inspected: Routine unannounced inspection by. regional based inspectors
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of licensee action on inspe'ctor-identified items, Licensee Event. Reports,.
!  open items, I.E. Information Notices, Part 21 and generic letters.
 
;  Inspection procedures 30703, 92700,.92701, 92701-1, 92702, and'92703 were covered during this inspectio '
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Results: In the areas inspected, one violation was identified for     failure to include appropriate acceptance criteria for snubber' lock-up velocity
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DETAILS Personnel Contacted
*D.'Poole, Plant Manager
*B. Croley, Deputy Plant Manager
*G. Coward, Deputy Restart Implementation Manager
*S. Knight, QA Manager
*D. Army, Nuclear. Maintenance Manager
*R. Colombo, Regulatory Compliance Superintendent
* Shewski, Quality Engineer    *
H. Heckert, Staff Assistant (Acting)
*J. Browing, Regulatory Compliance Engineer
*J. Robertson, Nuclear Licensing Engineer
* Denotes those who attended the exit meetings. Licensee Action on Previously Inspector Identified Items (Closed) Unresolved Item No. 50-312/83-22-02: ' Approval of Proposed Amendment 97 to the Technical Specifications Table 4.1-1, " Instrument Surveillance Requirement," in Amendment 54, of the Rancho Seco Unit 1 Technical Specifications, contained typographical errors in the test column for items 48.a and The test column for item 48.a should have read NA and item 48.b should have read M, but instead they were reversed. To correct the above typographical errors, proposed Amendment 97 tc. the Technical Specifications required approva The proposed Amendment 97 was approved February 21, 1985, and issued as Amendment 60 to the Technical Specifications on March 8, 198 The inspector reviewed items 48.a and b of Table 4.1-1 of Amendment 60, and found the typographical errors have been correcte This item is close (Closed) Followup Item No. 50-312/84-26-02: Program for Changing Procedures to Reflect Technical Specification Amendments The licensee had been previously requested to examine their existing program for updating operating procedures and to evaluate any program modifications necessary to ensure that procedures are implemented in a timely manner to Technical Specification change This evaluation resulted in the recently issued Administrative Procedure AP.72, " Technical Specifications Amendment Procedure,"
effective date of October 15, 1986. The inspector's review of AP.72 revealed that Section 4.6 requires additional actions of the Plant Review Committee (PRC) with respect to processing a Proposed Amendment to Technical Specifications. Essentially,-it requires the cognizant engineer to present the change to-the PRC. Section 4. then requires any PRC members, for which their department procedural changes will be required upon NRC approval of the Proposed Amendment, to make known any desired issuance delays after NRC
 
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approval. LIf no requests for delays are made, then-the NRC_ approval date and the effective date of the operating procedure will be
' coincident. In order to ensure these procedures are implemented in a-timely manner, Section 4.6.2 requires PRC members to retain copies !
of the Proposed Amendment and to use the interim.between the proposal and approval to draft procedural changes for immediate implementation after amendment issuance. The licensee's evaluation and subsequent issuance of AP.72' adequately addresses the inspector's reques This item is close (Closed) Followup Item No. 50-312/84-31-01: Quality Assurance Review of Bulletin Response System The licensee did not provide a response to a March 10,.1983, bulletin until July 30, 1984. The bulletin requested a ninety-day response. The inspector was concerned about the timeliness of the licensee's respons In January 1985, the licensee's Quality' Assurance (QA) department committed to audit the system controlling bulletin responses and provide some action to prevent further delinquent response During January 14-17, 1986, the licensee performed audit No. 0-777 of NEP 3104.1, .2 and .3 as they apply to the control of and response to NRC I&E Bulletins. The summary for this audit report stated in part, "The procedures used to receive, control and respond to NRC correspondence are apparently adequate to control the
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response to I&E Bulletins." The audit reviewed the Coordinated Commitment Log (CCL) for I&E Bulletin commitments and overdue commitment responses. No overdue responses were foun In response to QA Audit 0-777, the licensee stated in a memorandum NL-B6-127, dated April 11, 1986, "The District has pledged increased management emphasis on commitment tracking and submittals-to regulatory agencies....This process will'be considerably streamlined ,
and improved with the development within the next month of a new commitment tracking system. Licensing has contracted with Stone and Webster ' Engineering Corporation (SWEC) to develop and implement the -
new system, which will be fully operational by May 1, 1986."
 
The inspector reviewed the available documents identified above and noted that the new commitment tracking system had its start date changed until after September 1, 1986. On December 15, 1986; the licensee signed out directive ND-86-19-A titled, " Commitment Management," which implemented a new coordinated commitment tracking system (CCTS). The new CCTS included all the information originally contained in the CCL system, and should improve the licensee bulletin response syste This item is close .
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    ' (Closed) Followup Item No. 50-312/85-04-01: -AFW' Start with-MFW Pressure Signal Testing The inspector identified tha'tone of three automatic: start signals, low main feedwater header pressure, was not being tested during the
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eighteen month shutdown surveillance. .However,'the licensee does
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use this signal on their monthly auxiliary feedwater pump surveillance test. ; Therefore, although the pumps :are-not:being-started with the low main feedwater header pressure signal during shutdown, per the technical specifications, the-licensee has(shown operability of-the start signal on a monthly basis. -In order to clarify the auxiliary feedwater surveillance testing requirements the licensee submitted to.NRR on: June 13, 1986,. Proposed Amendment-No. 148. The Proposed Amendment > revises Technical Specification to permit system testing of the auxiliary feedwater pump under conditions of either power operation or plant. shutdown. : Subsequent -
to this amendment request, the licensee:has proposed a change to the
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' Technical Specifications which'will permit operation of the Emergency Feedwater Initiation and Control .(EFIC) System. This Proposed Amendment No. 152, . submitted to NRR on December 5,1986
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  (letter JEW 86-713) will incorporate the auxiliary feedwater tests requirements of the previously submitted Amendment 1148. The inspector reviewed the documents submitted to NRR and is satisfied with the actions taken by the licensee to address this ite This item is close (0 pen) Followup Item No. 50-312/85-04-02: Review and Verification of Past Commitments and Design Implementation This item was generated.as a result of a commitment to install hydrogen monitor vent valves as a NUREG-0737 requirement. Since the hydrogen monitors penetrate containment, and do not receive a Safety'  '
Injection Actuation Signal (SIAS), they were required to be locked closed. This hardware change was performed by Engineering Change Notice (ECN) 2938. The Design Basis Report (DBR) found in the' major portion of this ECN states that the valves shall be administrative 1y locked closed. The actual work done for_this item is by sub-ECNs as per procedure NEP 4109 (Rancho Seco Configuration Control Procedure). 'The part of the commitment that failed was not'the ~
hardware installation, but the administrative controls (a software item).
 
The inspector talked with licensee personnel about ensuring that'
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past commitments were implemented. .The licensee is currently
. writing a procedure to-identify,. address, track and assure completion of all commitments madelpreviously. The-commitment evaluation program project procedure is . currently;in a draft for This item will be closed when this procedure'is finalized and inspected for its adequacy in verifying that hardware and software commitments are complete _, , _ _ _ .
 
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4 (Closed) Notice of Violation No. 50-312/85-08-01: Battery Maintenance Procedure and Data Errors The licensee's. response to this violation was previously reviewed in Inspection Report 86-25. The item which remained open concerned'the finding that during the initial inspection (Inspection Report N /85-08), Procedure EM.106, Revision 4, did not specify the step to be used when starting an equalizing charge without-performing'a discharge first. Licensee electrical maintenance personnel were using applicable'section of EM.106 to equalize the battery when required. Also, during this initial inspection, a review of battery test results identified errors.in the recorded data. A followup inspection in July of this year found that the latest issue of Procedure EM.106 had not been revised to address the weakness identified in NRC Inspection Report No. 50-312/85-08 and the licensee could not identify what actions were taken in response;
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to errors identified in battery test results dat ,
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During this inspection, the licensee agreed with the inspector that they did not have an existing written procedure, with specific steps that maintenance personnel should follow to place a battery or battery cell on equalize. Procedure EM.106 will be' replaced with EM.106 A, 106 A2, 106 B, 106 B2, 105 C, 106 C2, 106 D, 106~D2, 106-E, and 106 F, to cover battery testing. The' licensee stated that a new procedure (EM.151 - Equalize Charging of Batteries) will be issued prior to restart, for maintenance personnel to follow when charging a battery or battery cell, if required by battery surveillanc In response to the errors identified in the initial review of the battery test results, the licensee provided a memorandum dated July 15, 1985, from C. Linkhart to S. Crunk, which stated in part the following:
(1) " Existing procedures shall be rewritten to upgrade them, to eliminate a majority of the incorrect data seen on old procedure data sheets."
 
(2) " Maintenance Engineers will review all procedure enclosures /
data sheets instead of a maintenance foreman. This not only allows more time for review, it provides a fresh look at the data by a completely independent person."
 
(3) " Rough data from the field will no longer be copied onto a fresh enclosure. This practice was instituted with the good intention of providing nice, clean, presentable data for the history file. Unfortunately, it has too often resulted in transcription errors."
 
(4) "The comparison of new data to old data.has not been a formal process up to now. Our rewritten procedures will provide formal accounting of this comparison with guidelines for action to be taken when a negative trend is discovered."
 
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Based on the inspector's procedure changes review surveillance data and maintenanceof new changes in methodsand proposed proc this violation were equat adthe inspector c ofconc,ludedand handling
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th that the licensee  corre ccommitments ve tiabove actions for, This item is close k (Closed) Followu Re uired to Delete 50-312/85-27-02: RefereItem N AP-27 Revision b
Action," dated August 10CI-7Revision    1 to Quality A QCI-7, issue " Corrective suranceAction Procedure
     " whi h (QAP) No
    ,1984,
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c referenced Quality Contr l27, " Corre o
The inspector 198 had been cancelled ever and n Procedur This Action." e:, revision
  "Procedur  deleted step 7 ireviewed Revision 2 to
  " which referencednthe  the paragraph titledo. 27 dated Jan
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original QCI-7 titled "Co This item is closed . rrective  ~
  (0 en) Followu Em sProcedures  to Assure PrItem N o 50-312/86-07-07:    .
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Inspectio  er Control of NoncondensiblLicensee e Gases in an to Re ..
followup, n Report 50-312/86-07 operators which required examinatiidentifi ed an gases whenever the  pressu r zer a eiare on of aware of actions to procedures to assuropento t k item 3 for The inspector  emptie e that  h\
control noncondensible  /
p (1) Operatin reviewed the following do    4
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j { t System,"g Procedure (0P), A.74, Revisi    -
(2)  dated June 4, 1986c - Se ti OP A-1, Revision 20  on 20, " Control Rod Driv on e 1 {1 September 5,1986
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(3)
    ,Sec ti" Reactor Coolant Syste on 3.2 m," dated thI t  OP B.4, Revision 40 y  September 5, 1986 ,Se ti" Plant Shutdown c    U The  on 3.2 n !
and Cooldown," dated event re above documents had the f l
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tripped) quiring RCS venting occurso lowing instructions add d to the Therefore, prevent operation the in limit (see inspector , all CRDs shall be e , "If an run in (not
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licensee's of gas bound CRDs emergency procedur er had
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measures to been icould <inot
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t This item remains open pending further mplemented in the
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Based on the inspector's review of new and proposed' procedures,
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Docket No. 50-312:
procedure changes and maintenance changes in methods of handling surveillance data, and the licensee's specific commitments above,
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the inspector concluded that the licensee corrective actions for this violation were adequat This item is close (Closed) Followup Item No. 50-312/85-27-02: QAP-27 Revision Required to Delete Reference to a Voided QCI-7 Revision 1 to Quality Assurance Procedure (QAP) No. 27' " Corrective-
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Action," dated August 10, 1984, referenced Quality Control Procedure QCI-7, " Corrective Action," which had been cancelled and never issue '
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The inspector reviewed Revision 2 to QAP No. 27 dated January 1,~
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1986. This revision deleted step 7 in the paragraph titled '
" Procedure:," which referenced the original.QCI-7 titled." Corrective, Action."     ,
This item is close > (0 pen) Follow 9 Item No. 50-312/86-07-07: Licensee'to Reexamine Procedures to Assure Proper Control of Noncondensible Gases in an Emergency Inspection Report 50-312/86-07 identified ao_open' item 3 for-followup, which required examination of procedures to assure that operators are aware of actions to take to control noncondensible gases whenever the pressurizer emptie The inspector reviewed the following documents:
  (1)' Operating Procedure (OP), A.74, Revision 20, " Control Rod Drive System," dated June 4, 1986 - Section (2) OP A-1, Revision 20, " Reactor Coolant System," dated September 5, 1986 - Section 3.2 (3) OP B.4, Revision 40, " Plant Shutdown and Cooldown," dated September 5,1986 .Section 3.2 The above documents had the following instructions added, "If an event requiring RCS venting occurs, all CRDs shall be run in (not tripped) to the in limit (see...for venting requirements)."
 
Therefore, the inspector could not determine whether measures to prevent operation of gas bound CRDs had been implemented in t' e licensee's emergency procedur This item remains open pending further revie _
 
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    , (Closed) Notice of Violation No. 50-312/86-08-02: No Control of '
Measuring and Test Equipment (M&TE)
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The licensee's Administrative Procedure AP12 (Plant Housekeeping ~and Inspection) required that " Tools and test equipment shall be stored in their proper location'at the;end of the workday and anytime when not in ose."    <
Contrary to these requirements, an inspector found, on two occasions, calibrated tools uncontrolled at the= work site, when no work was being accomplished. On March 5, 1986, and again on March 7,1986, an inspector observed unattended calibrated tools placed on a tool cart, in cardboard boxes, and on the floor of the computer room adjacent to the control roo In response to this item, the licensee had each piece of reported equipment checked to ensure that inadvertent damage had not occurred. Each item was subsequently found to be in proper working order. Additionally, to ensure that greater care will be exercised over calibrated equipment in the future, the electrical maintenance superintendent issued verbal instructions to place M&TE within carts, tool boxes or cabinets while not physically in use. These verbal instructions-have also been included in Revision 5 (issued June 30, 1986) to Administrative Procedure AP.33 (Calibration and Control of M&TE), in paragraph 6.3.2 which states ". . . All individuals and supervisors must not allow M&TE to be left in any area when it is not being used and must ensure that M&TE is returned to the appropriate storage area after use."
 
The inspector reviewed the applicable licensee documents (Administrative Procedures, responses, etc.) and it appears the licensee has taken the necessary' corrective action to prevent recurrence of this ite This item is close (Closed) Notice of Deviation No. 50-312/86-18-08: Failure to Satisfy Commitment This deviation addresses the licensee's failure to satisfy * commitment to make a permanent revision to Procedure I.103 by -
February 28, 1986. The licensee responded to the Notice of Deviation in a letter (JEW 86-223) to Region V dated July 21, 198 The response pointed out that there has not been a need to physically perform the power range nuclear instrumentation calibration since the plant was' shutdown on December 16, 1985. In addition, the Power Range CalibrationfP rocedure, I.103, can only be
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done when the reactor is generating enough power to be measured by the ex-core detectors. The inspector verified that the licensee had
, revised Procedure I.103 as they previously committed to do. Also, the inspector reviewed a draft Management Directive which, when approved, will apply to the identification, tracking, implementation
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and closure of commitments by the District to regulatory and other i
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external agencies.' This Management Directive appears to provide the necessary instructions to ensure timely completion of commitment This item is close Unresolved Item No. 50-312/86-21-02: Licensee Acceptance of Snubber Test After First Test Failed Snubber No. 129 successfully passed its surveillance test after having failed the same test the previous day, June 26, 1986. The inspector expressed concerns regarding justification for: declaring the snubber operable, and why an NCR was not generated when the snubber failed to meet the acceptance criteria when first teste The inspector reviewed QA Procedure No. 26 and verified-that the procedure was revised to include the requirement that an NCR be written when surveillancejtest results are not in conformance to acceptance criteri In regards to the operability of the snubber, the inspector has determined that at the time these concerns were identified,.the licensee was utilizing a procedure which contained an inappropriate t
acceptance criteria. Specifically, the acceptance criteria of Procedure SP 201.10B did not compensate for the effects of temperature when performing snubber functional tests. Also, at this time information on temperature compensation requirements was available in vendor manuals located in the licensee's Technical Manual Library. A calculation performed by the inspector revealed that Snubber No.129 fails to meet the acceptance criteria when the effects of temperature are taken into consideration. That.is,'the snubber lock-up velocity at test temperature of 78*F was 20 inches per minute (ipm) whereas the lock-up acceptance limits are between 1 and 18 ip Failure of the licensee to include, appropriate acceptance criteria in their procedure for functional testing of snubbers is an apparent violation (50-312/86-21-02). , (Closed) Unresolved Item No. 50-312/86-21-08: Decay. Heat Removal j (DHR) System Put Into Service Without Initiating Operation of the Nuclear Service Raw Water (NSRW) System As an example of a lack of attention to detail in the performance of routine personnel activities, a train of the DHR system was put into service without initiating operation of. the NSRW system as required by plant procedures. This error was identified to the NRC by licensee personnel after they discovered i ~
In memorandum NL 86-936 dated December 8, 1986, from H. Sims to A. Little (Subject "CCL #R8608180056, clarification of response"),
the licensee identified what they had considered as the cause of this occurrence. Operation Procedure (OP) A.8 (Decay Heat System)
stated in paragraph 4.3 (DH removal during RC system cooldown) the following:
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  " Initial Conditions
  .1 The primary system temperature is <290 F, and~one or two RC pumps running preferably in Loop .2 The Nuclear' Service Cooling Water System in service to the DH Cooler as per OP A.2 .2 The Nuclear Service Raw Water System in service to the Nuclear Service Water Coolers as per OP A.25."
 
In order to meet these " Initial Conditions," the Operator should .
follow OP A.24 and start NSCW Pump P482A (P482B), then follow OP A.25 and start NSRW Pump P472A (P472B). .It was felt that these steps should ensure the proper operational mode of the-system before actually beginning the " Procedure" steps of OP Obviously, however, there was a case when this did not occur and-resulted in the noted problem. Therefore,:to eliminate future recurrences the licensee issued Revision 29 to OP A.8 which added the following first step to the procedur "4. Start NSRW Pump P472A (P472B) and NSCW Pump P482A (P482B)
and verify proper operation."
 
While the licensee considered paragraph 4.3.8 redundant, it did provide a " double-check" to ensure both available water sources were in service before proceeding with decay heat removal step The inspector reviewed the applicable licensee documents (operating procedures, responses, etc.) and it appears the= licensee has taken action to prevent a recurrence of this ite This item is close . Licensee Action on Licensee Events Reports (LER) (Closed) LER 83-24, Revision 0: 'B' Nuclear Service Raw Water Pump Tripped Due to Cable Grounding This LER reported the licensee's actions in response to a ground fault that occurred May 19, 1983, in the B phase of the breaker supplying the nuclear service raw water pump. The B phase cable-was repulled and spliced to eliminate the ground as part of the initial corrective action. The other two unaffected phases were also identified to be repulled at a later date as a precautionary measure to ensure no additional problems would be encountered with the pum The inspector reviewed the applicable licensee documents and noted ECN (ECN) No. A 4905 was prepared December 13, 1983, to replace the existing spliced cable with a new 3-I/C 250 MCM, SKV cabl According to work request (WR) No. 92879 issued November 27, 1984, the work required to accomplish ECN No. A 4905 was completed May'23, 1985, and the ECN-was signed off completed on June 1, 1985. It
 
  ' appears the licensee has completed the original precautionar corrective action identified to ensure no additional problems would be encountered with the pum This item is close b. (Closed) LER 84-11, Revision 0: Incorrect Configuration Tables in ,
Surveillance Test Procedures On March 6,(1984, the licensee identified that the configuration tables in Surveillance Procedures-SP 203.02 A, B,-and C (Quarterly I and Annual Inspections and Surveillance Tests for_ Hi>I Loop A, HPI Loop B, and Makeup System Pump and, Valve) were misleading and incorrect, with respect to the cross-tie isolation valves. The configuration tables allowed three (3) differentL configurations, one of which was contrary to the Technical Specifications,;but had never been used. The configuration tables are allowed to-be used a directed by the shift supervisor, but they are primarily used for-
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information purpose As a corrective measure the licensee stated they wouldl revise Procedures SP 203.02 A, B and C to delete the configuration tables and reference Operating Procedure A.15 (Makeup, Purification and Letdown System) for the allowable breaker and valve configurations for the makeup and high pressure injection pumps. The inspector reviewed the applicable procedures and verified.they had been revised as require This item is closed.-    !
c. (Closed) LER 84-24, Revision 0: Simultaneous Plant Heatup and Deboration Violated Procedural Control of Reactivity Addition On November 7,1984, plant heatup was commenced during reactor coolant system (RCS) deboration. The RCS deboration resulted in 10 ppm reduction in boron concentration over a period of one ~ hour and 26 minutes, while heating up the RCS to 440 F. The core reactivity at the end of the event was -4.8% Delta K/K, which is 3.8% Delta K/K more negative than the required 1% Delta K/K shutdown margin. The event commenced on November 7, 1984 at 1750 when the swing shift stopped plant heatup in order to perform Surveillance Procedure SP 203.11 (Decay Heat / Core Flood Systems Stop Check / Check Valve Seat Integrity Surveillance Test) for the core flood tank check valves, which requires RCS temperature and pressure to be stabilize During this pause in the plant heatup, a deboration was commenced at 1835. At approximately 2330, the swing shift was relieve _
However, during shift turnover, the significance'of the plant deboration in progress was not emphasized-to the oncoming shift. On November 8, 1984 at 0021, the relieving shift supervisor, unaware of the deboration in progress, directed the control room operators to start heating up the RCS and then went into the shift supervisor's office to complete administrative paperwork. At approximately 0226 the shift supervisor noted that deboration was being performed simultaneously with plant heatup and secured deboratio '
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As corrective actions, thetlicensee performed the following:
  (1) Issued a memorandum to the operators emphasizing the importance and necessity 1.for proper transfer. of information during shift -
turnove (2) Conducted a. review of-the shift turnover practices,.which included discussions with INPO representatives, shift supervisors and many operators hired from other utilities. A number of ideas were brought up and ' incorporated in-licensee procedures. ~The inspector reviewed AP.23 (Revision 20), other applicable documents and the following memorandums:
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(a)~ D. Comstock to licensed operators,= dated December 10,
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1984, on" shift-turnovers;x (b) D.~Comstock'to G. Coward','dated Februa'ry-13, 1985, on LER s 84-24, CCL 85-0004;
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(c) B. Spencer to shift supervisors, dated March.28, 1985 (S0 5-85),on changes'to AP.23. Also: identified as Special Order.5-85; (d) G. Coward to.B.. Spence'r, dated' April 11, 1985, on Comstock's memo, dated February 13,~ 1985; and (e) G. C. Wallace to distribution,' dated May 9, 1986 (NOS
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86-147) 'on Revision 20 of AP.2 (3) Revised AP.23 in Revision 17'to add new relief / turnover
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checklists for shift supervisors to power plant helpers, and two equipment checklists for a control room operator to fill out during the shift. These documents were added to aid in a more complete shift turnover and increased communications between crew Based on the information reviewed, it appears that the licensee has taken applicable steps to ensure a detailed shift turnover, which should preclude a recurrence of this even This item is closed. (Closed) LER 84-25, Revision 0: Reactor Trip The inspector investigated the LER to ascertain whether the licensee's review, corrective action, reporting of the event and associated conditions were adequat This LER was generated when the reactor tripped on high pressure due to a Main Feedwater (HFW)-transient. The reactor trip occurred on November 18, 1984, during a power escalation. The feedwater transient was induced by the Integrated Control System (ICS)
attempting to keep up with rapid steam header pressure swing Following the trip, the large auxiliary boiler had trouble staying f
 
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   ' Rancho Seco Nuclear Generating Station 14440 Twin Cities Road-      .
on line to feed the Auxilary Feedwater Pump. Turbine (AFPT), which was needed due to steam loads exceeding core decay heat productio Also, there was some difficulty with the "B" Auxiliary Feedwater Pump P-318 (turbine driven) steam admission valve (FV-30801), which had stuck in midposition. In addition, the pump was secured at a pressure which should have reset the auto-start pressure switch (PSL-31758) but failed to do s The inspector reviewed the trip report and corrective action taken by the licensee. The root cause of the trip was failure of a control room operator to keep the governor valve limiter higher than actual valve position demanded by the ICS, and then rapidly raising the valve limiter higher which induced the transient. The corrective action taken by the licensee was procedural cautions in procedures A.46 (Main Turbine System), B.2 (Plant Heatup and Startup), and B.3 (Normal Operations) to keep the valve limiter at 100%. These cautions imply that the ICS will now have control over the governor valves to the turbine under most of the plant conditions. The inspector was also informed that the MFW controllers have been recalibrated and now respond faster and more accurately to changing flow conditions. The steam admission valve has been added to the preventive maintenance _(PM) program. The auto start pressure switch will no longer be used, since the AFPT will be controlled by the EFIC system when it is installed. The licensee is also doing work on the boilers to improite their reliabilit The corrective actions taken by the licensee should lessen the likelihood for a reactor trip from the same caus This item is closed.
Herald, California '95638-9799 Attention: Mr. John E. Ward '
 
e. (Closed) LER E5-01, Revision 1: H, Monitor System Containment Isolation Valves Found Open for 7 Days This LER was generated when the licensee discovered that four hydrogen monitor system containment isolation valves were apparently lef t open for seven days. 'This installation was to meet the requirements of NUREG-0737, item II.F.1, attachment 6. The purpose of this NUREG-0737 item was to provide continuous indication of the hydrogen concentration in the containment atmosphere to the control room. The work on the valves was performed under ECN-293 This LER has been addressed previously in inspection report 85-0 This report left this LER open and also generated three additional followup items and referenced another item from a previous inspection report as being similar in nature. These followup items have been addressed in other inspection reports as follows:
Followup Item Status Inspection Reports 84-19-05 Closed 85-04, 86-36 85-04-02 Open None 85-04-03 Closed 86-18 85-04-05 Closed 85-30
 
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The one remaining followup item (85-04-02), was for the licensee to ensure that past design commitments had been implemented. -This item is described in the followup section of this repor The corrective actions taken by the licensee include: adding the-four valves to the locked valve list, the addition of ~ a plexiglass cover over the controls for the valves, and addition of valve positions to the IDADS computer in the control room. These corrective actions were inspected by the inspector and found to be acceptable. The remaining followup item (85-04-02) will be tracked under that ite This item is close (Closed) LER 85-03, Revision 0: Incorrect Boron Concentration Technical Specification Limit On February 5, 1985, the licensee identified that the Technical Specification limit for boron concentration during reactor vessel head removal and fuel loading / unloading was incorrect. This discrepancy was the result of the fuel supplier basing the cycle 6
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refueling boron concentration on 'a Keff of .99 rather'than the specified Keff of .95, and this resulted in a Technical Specification limit of 1850 ppm versus the correct value of 1936 ppm. This condition had existed since the beginning of fuel cycle 6, on June 17,-1983. The refueling Keff was changed from 0.99 to 0.95 in the cycle 4 amendment to the Technical Specifications; the corresponding refueling boron concentration was calculated properly at that time by the fuel supplier (Babcock and Wilcox). For cycle 5, the fuel supplier engineers erroneously used outdated and uncontrolled Rancho Seco Technical Specifications and reverted to basing the refueling boron concentration on a Keff of 0.9 As a corrective measure to preclude further occurrences, the licensee required the fuel supplier to provide tighter control over its reload calculations by destroying all noncontrolled supplier copies of Rancho Seco Technical Specifications. Additionally, the licensee required the fuel vendor to audit the vendor's control of Rancho Seco Technical Specification The inspector reviewed the results of Rancho Seco audit reports, audit No. 0-725 (June 10-13,1985) and No. 0878 (October 22-24, 1986) which found that no uncontrolled /out of date copies of the Technical Specifications were in the possession of Babcock and Wilcox. Based on the' inspector's review, it appears that the licensee has taken'the applicable steps to prevent a recurrence of a-similar proble This item is close <
 
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- (Closed) LER 85-04, Revision 0: Fire Dampers not Installed as Required by Fire Hazards Analysis On February 11, 1985, the licensee identified that several fire dampers which were included in their August l', 1977 Fire Hazards Analysis (FHA) submittal to the NRC, had not been installe Amendment 19 to the licensee facility operating license
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(February 28, 1978) was written based on this analysis. The-implementation date for the fire dampers of concern was the end of the 1979 refueling outage. Thus, the licensee failed to implement, these provisions of Amendment 1 ,
The' licensee stated that: "Previously, the areas for which the fire dampers were'not installed had been designated,.for-other reasons, as fire watch areas requiring hourly surveillance; therefore, no'
immediate corrective action was required."
 
The licensee root cause analysis (incident No. 85-11) dated September 17, 1985, revealed the following information:
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(1) The lack of specific details in the'1977 FHA made tl$e determination and monitoring of the commitment difficul (2) The lack of an integrated, district-wide commitment tracking program did not provide sufficient commitment visibility to the personnel involve (3) The ongoing evaluation of the district's fire protection program has eliminated the need for some of the originally -
required dampers and has resulted in a revised fire hazards analysi This analysis determined that the root cause was "the lack of detailed engineering procedures to ensure commitments are properly implemented."
 
The licensee corrective actions identified to address the fire protection concerns were:
Provisions for making the FHA a "living" document undergoing periodic review and update *
The improvement of design control, by including a cognizant fire protection engineer in the review cycl *
Installation of fire dampers consistent with the FH *
Development of detailed engineering procedures to ensure commitments are properly implemente *
Additionally, the district developed an integrated, computerized CCL system to facilitate the logging and tracking of commitments. This system is now part of the new
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b coordinated commitment tracking system (CCTS)' issued December 15, 198 ~
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The inspector reviewed the following documents:-
* QA Surveillance Activities Reports Nos. 734, 735, 736, 737 (dated October 14, 1986) and 742 (dated October 21, 1986),which provided feedback ~that the licensee had satisfied the intent of the identified cbrrective actionsJto be taken'to address the 1 .
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fire protection concern * Applicable sections of ECNs Nos'. A-5514, A-5529, A-5767, R-0763, and R-0764 which were issued.to ensure the installation of fire dampers was consistent with the FH Based on information reviewed by' the insp'ector, it appears that the licensee has completed the appropriate actions'to address the fire protection concerns in this LE .This item is close (Closed) LER 85-05, Revision 0: Closed Boration Path During Fuel Movement On April 20, 1985, the licensee identified that manual valve BWS-041 was closed while fuel movement was in progress. BWS-041 is located in the flow path between the concentrated boric acid storage tank and the decay heat removal pumps, which provides the only method of borating the system in the event of a boron dilution accident during fuel shuffle. This condition existed for approximately one (1) hour before being detected and correcte The improper valve positioning resulted from a procedure error in refueling procedure (Refueling Equipment Checks and Core Component Handling) which opened BWS-041 in step 4.20.15 and then mistakingly closed it while performing the containment isolation valve line-up in step 4.2 As a corrective measure to ensure valve BWS-041 is open during fuel movement, enclosure 7.1. to Procedure B.8 (which is referenced in step 4.20.15) has been changed per Revision 17 to place a clearance control tag on valve BWS-041 after it is open. Step .4 of enclosure 7.1 (system check list) reads: " Concentrated boric acid system is lined up in accordance with OP A.12 with the exception that BWS-041,
' boric acid supply to decay heat' is open and under clearance to remain open during refueling operations." Once the clearance tag is installed on valve BWS-041, its position cannot be changed without obtaining the correct signatures. ' Based on information reviewed by the inspector, it appears that the licensee has taken responsible steps to prevent closure of the valve during refueling operation ,
This item is close ,
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   . (0 pen) LER 85-20, Revision 0: Essential INAC Flow Controller Design Error Prevents Auto Control After Return of Power On October 7, 1985, the licensee iden'ified that the flow
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t controllers for the control room / technical support center essential hVAC. filtration units (trains A and B) were notf functioning in accordance with the system's design basis report. Specifically, upon being reenergized following a loss of external power, the flow controllers would assume the manual mode of operation rather than the automatic mode. In the manual mede, the flow controllers would not respond to signals provided to control the air flow' rate within the Technical Specification. limits. The controllers for both trains are located in an enclosed box on the roof'of the auxiliary building. To switch from the manual to the automatic mode in the reported configuration would_ require that personnel be dispatched to the roof, remove the enclosure cover and depress the " auto" push butto The controller discrepancy was detected while' technicians were performing an investigation to determine why Surveillance Procedure SP 211.01A (CR/TSC Emergency Ventilation Systcm Loop "A" Monthly Surveillance Test) faile The discrepancy was believed to be a result of the licensee's failure during procurement, to explicity state the requirement for the controller (s) to assume the automatic mode following reenergization from a loss of external power. The licensee's Incident Analysis Group (IAG) will perform a root cause analysis of this even If the conclusions reached by the IAG differ from the conclusions of this report, the licensee will submit a supplemental repor The flow controllers of concern are the Foxboro 2AC type. To correct the discrepancy ard meet the system design criteria, instrument technicians added a jumper wire. to each controller to ensure that the controller remains in the automatic mode except when the manual push button is held ia a depressed condition. The correction was made with the concurrence of the controller manufac.turer and will not compromise the Class I qualification of the system. The corrective action was completed on October 10, 1985, through ECM R-017 The inspector reviewed applicable licensee documents and noted the following information:
(1) Rancho Seco Unit I Technical Specification 4.10 (Control Room / Technical Support Centar Emergency Filtering System)
states: "During an SFAS and a loss of offsite power, the "B" train of essential HVAC equipment is sequenced to automatically start upon its actuation signals approximately 6 minutes after the diesel generator breaker closes."
 
(2) The Foxboro Instructions MI 250-120 dated October 1984 state in a note on page 2 under Functional Description: "When power is-
 
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first applied, the station is forced into the manual-mode until another mode is selected by-the operator."2 (3) Memorandum from S. Crunk to MRT,. dated September 19, 1986,.
which stated: " Presently the IAG has three' individual.LERs relating to the CR/TSC essential HVAC (85-20, 84-13 and 86-07)
pending evaluation. Due to the pervasive'and diverse ~ nature of the HVAC problems, I propose that a combined investigation of
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the overall problem and causes be performed under one Root Cause Investigation (RCI). This investigation will follow the corrective action program and will be completed after the system has been 're-accepted' af ter successful testing." 7A handwritten note added to bottom of the memorandum stated:
" Approved at Management Review Team (MRT) meeting dated 10/1/86, will be officially signed at MRT meeting dated 11/4/86."
 
Af ter reviewing the available information, the inspector identified the following concerns / questions listed below to the licensee:
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It appears that the subject flow controllers were procured with inadequate purchase specifications. The inspector questioned whether any other similar controllers / equipment purchased without specification of mode of operation after reenergization from a loss of external power. According to a licensee representative, these two flow controllers were the only two of that model purchased for this sit The inspector questioned whether the installation of these two flow controllers passed the original system acceptance testing, or any other previous surveillance testing, whether the test procedures were inadequate or not correctly followed. A licensee representative stated that these questions have been passed on to the Incident Analysis Group (IAG) and will be addressed in the final rcot cause investigation repor *
The inspector questioned why over a year has passed since the occurrence of this Licensee Event Report of October 8,1985, and the licensee still has not identified a root cause or addressed corrective action to prevent a reoccurrence of a similar problem. Licensee representatives could not provide an answer to this at this time, other than it was a low priority item. Since the licensee has purchased a large amount of new
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equipment for maintenance, modification and repair work in this last year, this appears to be an example of insufficient corrective action. The licensee management apparently should have taken more prompt action to determine how this happened and what could be done to prevent a similar occurrenc This item remains open pending a more complete licensee respons ,
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  .- , (Closed) LER 85-21', Revision 0: Emergency Diesel Generator-(EDG)
Auto Start Due to Personnel Error While Trouble Shooting On November 19, 1985, while the plant'was operating at 83% power, Emergency Diesel Generator (EDG) "A".was automatically started when the 4A bus normal supply breaker tripped on an overvoltage ,1 condition. The EDG "A" output breaker closed to the 4A bus and the bus reloaded as designed; however, the EDG "A" supply fan tripped shortly thereafte ~
An investigation of the event revealed that the overvoltage condition was created when electrical: technicians, who were replacing a relay in the 4A bus voltage protection circuitry,.
improperly disconnected ground wiring to the circuit's active relays. Lifting of the ground wiring caused the relays to " timeout" (trip) on a loss of AC power, thereby resulting in.the bus normal supply breaker tripping on two-out-of-three overvoltage logic. The subsequent EDG "A" fan failure resulted from a fan breaker overload device setpoint being out of the specified rang To prevent a recurrence of this event, the licensee took the following corrective actions:
(1). The Incident Analysis Group (IAG) prepared Lessons Learned Report No. 85-01, which alerts plant maintenance personnel of-the actions leading to the event and outlined.the proper precautions to take. The electrical' technicians used elementary diagrams for this work, which did not reflect the detail of the connection diagra (2) A monthly check of the 3A17 breaker phase overload devices was implemented for the next 6 months to ensure that the 3A17 overload devices remain set within the specified range and to determine if drift of the devices was a genuine concer The inspector reviewed Lessons Learned Report No. 85-01 and other available documents. It appears the licensee has taken the applicable steps to prevent a recurrence of a similar proble This item is closed. (Closed) LER 85-24, Revision 0: Shutdown Due to Pressurizer Liquid Sample Isolation Valve Leak-On December 22, 1985, reactor coolant system leakage was calculated by a Control Room operator from frequent dumping of the 120 gallon reactor building accumulator tank. The leak rate was determined to be between 15 and 20 gallons per minute, shortly after initial detection. Unit shutdown action was initiated in accordance with Technical Specification 3.1.6 which requires the reactor to be shutdown within 24 hours of detection of a ~ reactor coolant ' leakage rate exceeding 10 gp During operator actions to identify and isolate the leak, an attempt was made to close "B" letdown cooler outlet isolation valve (HV-22008). The valve did not fully close; m
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however, it was later determined to not be"in the leak path. The plant'was brought .to a hot' shutdown condition. The pressurizer
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liquid sample isolation valve'(SFV-70001).was' closed a'nd the leakage
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An investigation of the event determined that the leakage originated from the packing gland of SFV-70001*. SFV-7.0001 is the inside containment isolation valve and;is normally closed during power operation. .It had been opened approximately four (4) hours prior to the event to allow testing of the Post-Accident Sampling System (PASS). The valve packing gland was disassembled and the stem inspected for damage. No damage was observed. Twelve (12) rings of  ,
packing were added and the packing gland was adjusted to eliminate leakage. This corrective action was completed on December 24,:198 Following the event, HV-22008 was' examined and found to operate properly. The valve was stroked from its motor control center (MCC), the position indications checked, and also positioned and timed from the. control roo It was further determined that the valve is not designed to close against system pressure; however, the valve could be closed under the condition of the event, if necessary, by first closing an upstream valve such as SFV-22006,
  -(letdown to cooler E-220).
 
The inspector reviewed WR No. 108882, Casualty Procedure C.19 (Letdown Cooler Coil Failure), Revision 8, and other applicable licensee documents on this subjec It appears the licensee has taken corrective actions to ensure valve SFV-7001 operation is acceptable and that Casualty Procedure C.19 has been changed to first close an upstream valve prior to closing either letdown cooler outlet isolation valve (such as HV-22007 or HV-22008).
 
This item is close . (Closed) LER 86-04, Revision 0: Missed Fire Watch On March 14, 1986, the licensee determined that three fire doors were inoperable and fire watches had not been posted within I hour in accordance with Technical Specification 3.14. The results of the refueling interval fire barriers Surveillance Procedure (SP 201.3L) were reviewed, and it was determined that these doors were
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found to be inoperable. If a fire door is found to be inoperable, the Nuclear Operations Fire Protection Coordinator should be informed by the surveillance procedure, and a fire watch initiated in accordance with an Administrative Procedure-(AP.60).
 
The root cause of-this item was a personnel error. The individuals-involved.in this event have been counseled'and retrained to adhere to procedure This item is close I r i e
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4. Licensee Action on I.E. Information Notices
    ' (Closed)'I.E. Information-Notice No. 85-42: Loose Phosphor in Panasonic 800 Series Badge Thermoluminescent Dcsimeter (TLD) ,
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This information notice alerted NRC licensees to a problem noted in some Panasonic 800 series TLD badges ~that has caused spurious high readings in one of the badges' TLD element The Panasonic 800 series,TLD badge contains a card that holds four TLD elements. Each TLD element consists of a thin film of TL'
phosphor attached to a disk backing with a clear teflon bubble cover. During reading, the phosphor is heated by converging infrared light on the backin The luminescence from the phosphor (which is proportional to the dose received) radiates through the teflon cover and is read with a photomultiplier' tub Several Panasonic TLD users have identified badges where crystals of the phosphor have detached themselves from the backing of the element, resulting in high erratic readings in that element. When viewed through a stereoscopic microscope, phosphor crystals can be observed sticking to the teflon cover (presumably by electrostatic-charge). In this position, the loose TL material is not in contact with the backing and does not get heated when the badge is rea These TL crystals remain at an elevated energy state and continue to accumulate dose. Apparently erratically high readings result when the loose crystals are shaken back onto the backing surface during a subsequent reading. They are then heated and luminescence proportional to the total doses received during several read cycle This process can cause the affected element to erroneously read as much as an order of magnitude higher than the other' elements in the same card. Although the frequency of occurrence is small (one licensee found only one problem badge in 30,000), there is evidence that the frequency increases substantially once the badges have been through 100-200- read cycle The licensee stated in a memorandum JR 85-82, dated June 21, 1985, that at their facility TLD badges were not used routinely for personnel use but were normally used for environmental trendin The licensee had not seen any abnormally high or low readings that would indicate loose TL phosphor. Additionally, the TLD #/ badges used at their facility had been purchased within the last two to three years and had not been through 100-200 readings, so they would not expect any loose paospho The licensee issued a revision 4 (dated December 12, 1986) to Administrative Procedure AP.308-8 (Panasonic TLD reader) which referenced this notice in paragraph 2.7 and added information to paragraph 3.10 for inspection for loose TL crystal The inspector reviewed the applicable licensee documents and it appears they have taken adequate action *
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This item ir. close (Closed) I.E. Information Notice No. 86-56: Reliability of Main Steam Safety Valves (MSSVs)
This notice provided additional notificacion of NRC's concern for the reliability of spring-actuated main steam safety valves following reports of multiple failures during testing and problems during power operations and scram recover In memorandum SRT 86-148 of October 3, 1986, the licensee acknowledged that the problems identified for MSSVs were' applicable to their valves, and they were working on the issue. To improve blowdown performance of these valves _ Dresser had issued recommendations which established new ring settings for 3707 and 3777 valves, which the licensee was incorporating into Procedures NT.004 and H.2 The licensee considers that their continued involvement in the B&W owner group secondary relief project will aid in their effort to improve pressure response on the secondary sid *
The valve performance after resetting these valves will be evaluated ,
by the licensee later during power operation Based on the inspector review of the above information and other applicable documents, it appears the licensee has taken appropriate action This item is close (Closed) I.E. Information Notice No. 86-63: Loss of Safety Injection (SI) Capability This notice alerted recipients to a potentially significant problem pertaining to the loss of SI capability as a result of common-mode failure of SI pumps from crystallization of boric acid or gas binding of the pumps. Leaky valves in the discharge line of the boron injection tank (BIT) could enable highly' concentrated boric acid to flow through the low pressure discharge line (SI pump suction) and to precipitate in the pumps which are not normally heat trace In memorandum No. 66-304 of October 29, 1986, the licensee stated that crystallization of boric acid in SI pumps (HPI pumps) may pose a concern for Westinghouse design plants utilizing highly concentrated boric acid solution (20,000 ppm), which crystallize at 126*F, but this is not a problem at Rancho Seco. At Rancho Seco, being a B&W design, the highest boron concentration expected in the system at BOL is approximately 1400 ppm. .At Rancho Seco the highest concentration of boric acid solution utilized, is in the concentrated boric acid tank (CBAT) and it would normally be 8500 ppm (which solidifies at approximately 40 F).
 
The possibility of concentrated boric acid solution leaking past normally closed valves and precipitating in the HPI pump (which are not heat traced), in sufficient concentration to result in
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crystallization at ambient temperatures, is ~not considered credible by the licens The inspector reviewed Technical Specifications A.12 (Reactor Coolant Chemical and Hydrogen Addition System), SP 203.02A(B) and other applicable documents, and it appears that the licensee has appropriate instructions issued to prevent the identified loss of safety injection capabilit This item is close Review of Licensee's Program to Review Information Notices /Information Bulletins The inspector reviewed the following interdepartmental procedures
; (IDPs) and the draft of a new procedure:
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(1) IDP No.-001, Coordinated Commitment Log (CCL)
(2) IDP No.-002, Control of incoming regulatory correspondence (3) IDP No.-003, Control of outgoing regulatory correspondence (4) A draft titled - Commitment Management (Issued December 15, 1986)
According to a licensee representative, the new draft procedure titled: " Commitment Management" will supersede the IDPs noted above, when issued. These IDPs provide instructions for review, distribution, and scheduling of performance of corrective actions as required. In a majority of the Information Notices and Bulletins reviewed, the actions taken by the licensee appeared reasonable and appropriate to the substance of the information documents. There were some I.E. Information Notices, such as No. 85-23: inadequate i
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surveillance and post-maintenance and post-modification system testing (issued March 22, 1985) which the licensee 'was unable to provide any documentation of what its status was at the time of this inspection. The licensee could identify who had been ' assigned responsibility, but that person did not have status within his group. The inspector stated that licensee management would be prudent to ensure that their new control system screened new items-for priority and then periodically tracked action being taken't address concern '
5. Licensee Action on Generic Letters  , (Closed) Generic Letter No. 85-05: Inadvertent Boron Dilution Events This letter informed the licensee of the NRC position, resulting from the evaluation of generic issue 22 (Inadvertent Boron Dilution Events), regarding the need for upgrading the instrumentation for detection of boron dilution events in operating reactors. There
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l were concerns addressed regarding the lack of' distinct, positive              /
alarms to alert the operators to boron, dilution event v, The licensee performed an analysis of this event and stated, " Based on the analysis performed in the Updated Safety Analysis Report
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occurring at this facility is minima The inspector reviewed the USAR Section 14.1.2.4'(Moderator Dilution Accident); licensee operational analysis (attached to memorandum EQC-85-427, dated April 29, 1985); Revision 20 to Procedure which added new substeps 6.1.1.1,_7.1.1.1, 7.2.1.1. and other applicable documents. The new substeps added this following
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information, " Place shif t ' supervisor's clearance on RCDST pumps, P-622A and B breakers 2E 116 and 2C 516, to prevent boron dilution              N of the RCS when drained dow If these pumps must be run for freeze l, protection, potential RCS boron dilution should be considered."              /
                        , -r Based on the information reviewed above, it appears the licensee has              f completed his actions for this ite ,
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This item is close . (Closed) Generic Letter No. 85-20: Verify Stress Analysis Performed on Modified Thermal Sleeve Designs for HPI Nozzle In 1982, inspections at B&W plants revealed that some of the high pressure injection / makeup (HPI/MU) check valve, valve-to-safe-end weld, safe-end and thermal sleeves'were cracked. A safe-end task force was formed by the B&W owners' group, which issued a report with its findings and recommendations to aid in resolution of generic issue 69. The NRC reviewed the task force recommendations and agreed that certain actions should be taken. One of the actions was to perform a detailed stress analysis of a nozzle with a modified thermal sleeve design to justify long-term operatio In this letter issued November 8,1985, recent review of operating experience for some B&W plants has, indicated that the expected              -
fatigue analyses could be substantially exceeded by_the end of plant              i life. For example, an increased number of HPI actuation transients              g could occur due -to, manual. actuation. after reactor trips to avoid
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losing pressurizer level. Therefore, the NRC has determined that it is necessary that the licensee ensures that valid stress analys6s have been performed. Each licensee was requested to verify that a valid stress analysis had been performed for HPI/MU nozzles.and that the cumulative fatigue usage for these nozzles is ,within the allowables based on a realistic projection of the_ thermal cycles g expected for the' life of the plan The licensee is tracking the requirement for verification of a valid stress analysis under item No. 20.0219A, in the Quarterly Tracking System (QTS), with a due date of December 31, 198 >    ,
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The inspector _ reviewed the applicable licensee documents on this subject and it appears that the licensee is taking responsible actions in following up on obtaining verification of a valid stress analysi Based on the licensee commitment to verify the stress analysis, this item will be close This item is close . Licensee Action on Part 21 Items    ; (Closed) Part 21 No. 85-24: Oil Level Device on Auxiliary Feedwater Pump is Not Reliable The licensee issued letter RJR 85-545 on November 8, 1985, stating that the porthole oil level gauge for auxiliary feedwater pumps manufactured by Babcock and Wilcox Canada, Ltd, were not reliable indicators. The licensee stated the gauges were designed with a metal insert behind the sight glass that could cause the oil to be trapped so that a false level of oil was indicated. As an interim corrective action, the licensee recommended removal of the metal insert. As a long-term corrective action, the recommendation was to install an oil level sight glas The licensee issued ECN No. R-0173 to install a new vertical level oil gauge to isolate the effect of surface perturbation on the level indication. This ECN was voided later because the Plant Review Committee (PRC) did not want any gauge connections on the outside of the pump bearing housing, that could cause a loss of oil, if damaged. The PRC considered that the initial interim corrective action of removing the metal insert, provided an acceptable oil level readin The supplier of this type of oil level gauge (bullseye) indicat.?
that the insert was used as a reflector when observing "Hard-to-see fluids," and removal of the insert would not cause difficulty in reading the oil level. Since the metal insert was not shown in any
, pump drawing or technical manual, the licensee considered it an
! optional item and removed it from the applicable sight gauges.
 
Based on the inspector's review of the above information and applicable documents, it appears the licensee has taken adequate corrective actions for this ite This item is close (Open) Part 21 No. 86-13: Anchor / Darling-Missing Lock Welds on Internal Components of Swing and Tilting Disc Check Valves The licensee received the three letters from Anchor / Darling identified below:
 
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2  '(1) . Anchor / Darling to N. Bradford, Contract Administrator,' dated . ,
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July 30, 1985, on the subject of cracked tack welds .which loc the hinge pin busing in place on the tilting dise-check valves.-
  (2) Anchor / Darling to N.' Bradford, Contract Administrator,' dated
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July 31,.1985, on the subject of missing lock welds on hinge  ,
pin set screws .of swing check valves at Palo Verde. Nuclear .
  ~
.. Generating Statio (3) Anchor / Darling to N. Bradford',TContract Administrator,' dated-
. December 11, 1985, on'the subject of lock welds also missing at ,
i the hinges support / hinge. support capscrews; interface and at the hinge support / bonnet interface on swing check valve ~
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Deputy General Manager, Nuclea Thank' you for your letter dated February 23, 1987, informing us of the steps
  'you have taken to correct the items which we brought.to your_ attention in our z
letter dated January 22, 1987. Your corrective actions will be verified
  'during a. future inspectio ;Your cooperation with~us is appreciate


        +
Sincerely, Origino! sign 4f Ny
After reviewing the th'ree Anchor / Darling letters, the licensee  <
    . D. F. K tsch Dennis F. Kirsch, Director Division of Reactor Safety and Projects bcc w/ copy of letter dated 2/23/87:
performed operational assessment 86-8-(signed out June 21, 1986) and   -
State of CA Project Inspector L  Resident Inspector-
,  generated attachment 1 of that document which identified those  <
,  RSB/ Document Control Desk (RIfG l  B. Faulkenberry l  J. Martin l  J. Zollicoffer
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valves that-could be potentially defective..' Assessment 86-8 reconnendations were to implement an inspection / repair program for the identified valves prior to restar The inspector reviewed Operational Assessment,86-8-and requested an inspection status on the valves in the inspection / repair progra After comparing the valves identified in attachment <1:and the inspection status report to the list provided with the:  .
docket file
Anchor / Darling letter of December 11, 1985, the inspector identifie .
to the licensee that it appeared they had failed to include Valve-SFC-002 (A/DV Assembly.DWG 1338-3) in the inspection / repair' progra The. licensee representative agreed with the inspector and stated '
that WR No. 120092 would be issued to include ~this valve inuthe inspection / repair progra ' '
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As of the date of this inspection,.six palves of the twenty-two'  '
valves (m + includes valve SFC-002) identified _in,'the  ' - ~
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inspection /rci '' program had been' inspected. There~ are seventeen ' -
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swing check valves and five tilting. disc check valves in this; . .
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program. One of the four swing check. valves inspected required two
. new welds between the. hinge support / bonnet interface and both'of the tilting disc check valves inspected had hinge pin bushing' retaining
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i  welds cracked, and new bushings ~of a new design were installed with-weld:. .
Based on the inspector's review of the above information, the.
 
l  addition of valve SFC-002 and other applicable licensee. documents,-
l  it appeared the licensee has an informal inspection / repair program  -e
  'in operation that should resolve the Anchor / Darling swing and tilting check valve concerns about missing lock welds on internal ~
components, prior to the restart of.the unit.
 
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This item and item 6.c below are unresolved pending review of the-adequacy of the licensee's reporting of the defects;1dentifie ,
      > -
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  -
; bcc'w/o copy of letter dated 2/23/87:
        ,
l  M. Smith
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WM 2/24/87 -7,/2/787
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    /A'787 y,/g.[/87 R QT COPY ] REQUEST AOPh] REQUESTJ0PY ] REQUEST OPY ]
ES j NO ] YES / (N0 /] YES ' (NO3 ] YES /0  ]
_
SENDTTO PDR ] g (YESf/ N0 ]
l[0 $ f5 2
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-v  25 Inspector Observation of Licensee Program for Part 21 Items, Not Generated by the Licensee It appears the licensee is having some trouble in this area identifying where these items are received in their organization and then getting them into a formal tracking system for review and action as required. When the inspector requested status on Part 21 items No. 86-15-P, 86-22-P and 86-25-P, a licensee representative stated one had been assigned with no due date and the other two were not assigned yet. While these Part 21 items are relatively new, they concern Limitorque valve operators and the. licensee-is now performing major inspections, repairs and maintenance on Limitorque valve operator It may be that the actual licensee personnel performing the Limitorque work are knowledgeable of these Part 21 items, but it is not identified formally by the license It appears this area requires additional management attention to ensure that a Part 21 item is not received by_the licensee, and then lost or delayed in the system while inspection or work is being performed on the identified equipment. Once a Part 21 is received, '
it would appear prudent to immediately review it' to see if it'
requires immediate action, or to determine if.it effects existing inspections or work being performed. Discussions on this subject were held with licensee representatives, and they agreed this was a problem and stated they were already working on thi However, no clear action plan with dates for corrective action had been >
developed. The licensee's resolution of this issue will be reviewed when the unresolved Part 21 Item 86-13 above is resolve .
7. Exit Meetings Exit Meetings were conducted on November 21, 1986, and December 12, 1986, with licensee representatives identified in paragraph The inspectors summarized the scope of these inspections-and findings as. described in this report.
 
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Revision as of 20:03, 18 December 2021

Ack Receipt of Informing NRC of Steps Taken to Correct Violations Noted in Insp Rept 50-312/86-38
ML20212K950
Person / Time
Site: Rancho Seco
Issue date: 02/27/1987
From: Kirsch D
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To: Julie Ward
SACRAMENTO MUNICIPAL UTILITY DISTRICT
Shared Package
ML20212K952 List:
References
NUDOCS 8703090469
Download: ML20212K950 (1)
v • d • e


Text

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< FEB 271987' ,

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Docket No. 50-312:

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' Rancho Seco Nuclear Generating Station 14440 Twin Cities Road- .

Herald, California '95638-9799 Attention: Mr. John E. Ward '

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Deputy General Manager, Nuclea Thank' you for your letter dated February 23, 1987, informing us of the steps

'you have taken to correct the items which we brought.to your_ attention in our z

letter dated January 22, 1987. Your corrective actions will be verified

'during a. future inspectio ;Your cooperation with~us is appreciate

Sincerely, Origino! sign 4f Ny

. D. F. K tsch Dennis F. Kirsch, Director Division of Reactor Safety and Projects bcc w/ copy of letter dated 2/23/87:

State of CA Project Inspector L Resident Inspector-

, RSB/ Document Control Desk (RIfG l B. Faulkenberry l J. Martin l J. Zollicoffer

'

docket file

bcc'w/o copy of letter dated 2/23/87

l M. Smith

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!~ REGION V fg urP *~

- WANG / dot iller AChaffed DKirsch W

WM 2/24/87 -7,/2/787

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/A'787 y,/g.[/87 R QT COPY ] REQUEST AOPh] REQUESTJ0PY ] REQUEST OPY ]

ES j NO ] YES / (N0 /] YES ' (NO3 ] YES /0 ]

_

SENDTTO PDR ] g (YESf/ N0 ]

l[0 $ f5 2

)

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