IR 05000498/1987008: Difference between revisions

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{{Adams
{{Adams
| number = ML20214Q561
| number = ML20236D718
| issue date = 05/29/1987
| issue date = 07/24/1987
| title = Insp Repts 50-498/87-08 & 50-499/87-08 on 870309-0410. Violations Noted:Failure to Follow Procedures for Testing & Inadequate Cleanliness Controls Over Open Rcs.Major Areas Inspected:Tmi & Generic Ltr 83-28 Action Items
| title = Ack Receipt of 870625 Ltr Informing NRC of Steps Taken to Correct Violations Noted in Insp Repts 50-498/87-08 & 50-499/87-08
| author name = Bundy H, Carpenter D, Chamberlain D, Constable G, Cummins J, Hildebrand E, Johnson W, Jones W, Luehman J, Madsen G, Pick G, Reis T, Tapia J, Taylor R
| author name = Gagliardo J
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
| addressee name =  
| addressee name = Goldberg J
| addressee affiliation =  
| addressee affiliation = HOUSTON LIGHTING & POWER CO.
| docket = 05000498, 05000499
| docket = 05000498, 05000499
| license number =  
| license number =  
| contact person =  
| contact person =  
| case reference number = RTR-NUREG-0737, RTR-NUREG-737, TASK-***, TASK-TM
| document report number = NUDOCS 8707310038
| document report number = 50-498-87-08, 50-498-87-8, 50-499-87-08, 50-499-87-8, GL-83-28, NUDOCS 8706050075
| document type = CORRESPONDENCE-LETTERS, NRC TO UTILITY, OUTGOING CORRESPONDENCE
| package number = ML20214Q539
| page count = 3
| document type = INSPECTION REPORT, NRC-GENERATED, INSPECTION REPORT, UTILITY, TEXT-INSPECTION & AUDIT & I&E CIRCULARS
| page count = 80
}}
}}


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APPENDIX B U.S. NUCLEAR REGULATORY COMMISSION
      .. _
 
jd#"%  UNITED STATES y" - ~, NUCLEAR REGULATORY COMMISSION b  REGloN IV
==REGION IV==
  #
NRC Inspection Report: 50-498/87-08  Construction Permits: CPPR-128 50-499/87-08    CPPR-129 Dockets: 50-498 50-499 Licensee: Houston Lighting & Power Company (HL&P)
8  611 RYAN PLAZA DRIVE. SUITE 1000
P. O. B)x 1700 Houston, Texas 77001 Facility Name: South Texas Project, Units 1 and 2 (STP)
Inspection At: STP, Matagorda County, Texas Inspection Conducted: March 9 through April 10, 1987
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N Inspectors:
  .
  - -
  .Ma7penter, Senior Resident Inspector  D(te /
      /f'7 Project Section C, Reactor Projects Branch w
  -    flD
  ,T. heis, Resident Inspector, Project
      '
Da Section C, Reactor Projects Branch
      .fl2f/S7 H. F. Bundy, Project lYispector, Project  Date Section C, Reactor Projects Branch WP J. I. Tai >1a, ReactorInspector, Operations f/17/N 7 Date Section, Reactor Safety Branch G706050075 870529 PDR ADOCK 05000498 E_ __0  PDR  __ _ _ _ _ _ _ _ _ _  _
 
._ _ ___-________________ - - ________ ___ ___________
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        .
        . ..
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h J.~ G. Lu man, Seniof Resident Inspector 587/77 Date
        '
Project Section C, Reactor Projects Branch I
r    J. E. Cummins, Senior Resident Inspector D(ate /2Fb 7 Project Section B, Reactor Projects Branch l
fW D. D. Chamtierlain, Serrior Resident Inspector Tb W97 Date
    / Project Section A, Reactor Projects Branch
        .fbfA'?
    ~
W. D. Johns'on~~Senio Resident inspector Date Project Section B, Reactor Projects Branch i
    .
fAllfff7 G. L. Madsen, Reae dr Inspector, Operations (Tate ~
Section, Reactor Safety Branch
    . B. Jon s, Reside ( Inspector, Project ate Section A, Reactor Projects Branch
    - . Y% Al,14 A G. c , Reactor inspector, Operations Dat(
Secti n, Reactor Safety Branch I
 
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   / $
C P. fit Tde' brand, React 6r Inspector Date 29 ~ ~7 Operations Section, Reactor Safety Branch P
R. s b, Project Inspector, Project 4/o Dite/
Section , Reactor Projects Branch Consultants: M. Bishop, F. Jagger, N. Jensen, R. Picker, J. Stachew, J. Seherman, J. McGhee; EG&G Idaho In NRC Coop Student: J. Lara Approved:
i _ -c<
onstable, Chief, Project Section C ("
Da(e e ~)
Reactor Projects Branch Inspection Summary Inspection Conducted March 9 through April 10, 1987 (Report 50-498/87-08;
$0-499/87-08)
Areas Inspected: Routine, unannounced inspection of Technical Specifications (TS),
the structural integrity and integrated leak rate tests (SIT and ILRT),
preoperational test procedures, preoperational test results, the startup testing program, the as-built plant to documentation reconciliation, the residual heat removal (RHR)/ component cooling water (CCW) water hammer incident, operational staffing, training and qualification programs, the reactor coolant system loss of cleanliness recovery program, the Three Mile Island (TMI) and GL 83-28 action items, licensee action on previous inspection findings, site tours, review of the manual trip circuit, and procedures revie Results: Within the areas inspected, two violations of NRC requirements were identifled (failure to follow precedures for testing and inadequate cleanliness controls over an open reactor coc!}qt .S/ stem, paragraphs 8 and 11, respectively).
 
L.__._ . .
 
DETAILS Persons Contacted
*R. W. Chewning, Special Assistant Nuclear Group  l
*S. M. Head, Lead Engineer, Licensing
*D. L. Smith, Management Services Manager
*G. L. Jarvela, Manager, Health and Safety Services
*M. A. Ludwig, Maintenance Manager
*T. E. Underwood, Chemistry Manager
* L. Parkey, Technical Support Manager
*J. W. Loesch, Plant Superintendent
*M. T. Sweigart, General Supervisor, Operations Quality Control (QC)
*W. H. Kinsey, Flant Manager
*J. J. Eldridge, Operations Supervisor
*W. P. Evans, Project Compliance Engineer J. T. Westermeier, Project Manager F. A. White, Lead Licensing Engineer R. J. Daly, Startup Manager J. D. Green, Operations Quality Assurance (QA) Manager V. E. Geiger, Nuclear Assurance Manager M. Robinson, Director, Independent Safety Evaluation Group D. L. Cody, Manager, Nuclear Training M. E. Smith, Outage Manager J. Hooper, Employment Counselor D. Leazur, Reactor Performance Supervisor T. Godsey, Technical Support Engineer
* Denotes those individuals attending the exit interview conducted on April 10, 198 The NRC inspector also interviewed other personnel of HL&p, Bechtel Power Corporation, and Ebasco Service, In . TS Review The NRC inspectors and EG&G Idaho consultants reviewed the Proof and Review copy dated February 12, 1987, of the STP Unit 1 TS. In performing this review, the following techniques were employed:
o Comparison with NUREG-0452, Revision 5, " Standard Technical Specifications for Westinghouse Pressurized Water Reactors."
 
o Comparison with TSs for other recently licensed Westinghouse plants (Wolf Creek Generating Station and Byron, Unit 1).
 
o Walkdown of selected systems and components to verify as-built configurations were reflected in the T o Verification that numerical values of setpoints, operating criteria, and equipment operating parameters agreed with FSAR and/or engineering specification value _ _ _ - ._ _
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Resulting comments were discussed with the licensee and forwarded to the NRR project manager by memorandom on March 25, 198 l i
No violations or deviations were identified, t SIT and ILRT An inspection was conducted of the containment SIT procedures, test performance and test results in order to determine consistency with regulatory requirements and licensee commitments. The purpose of the SIT is to demonstrate the ability of the containment structure to withstand internal loads imposed by pressurizing to 1.15 times the design pressure of 56.6 psig or 65.0 psig. Bechtel Specification Nos. 20019S50009,
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Revision 2, " Instrumentation For Structural Integrity Test of Containment Structures," and 2C0195S1013, Revision 1, " Specification For The Performance of Structural Integrity Test on Reactor Containment Building, Unit 1," were reviewed by the NRC inspecto Prior to pressurization of the containment, the NRC inspector toured the containment and inspected the placement of instrumentation for the SI The containment was subsequently pressurized in five equal pressure increments. During the 1-hour hold periods between pressure levels, strains, and deflections were recorde Surface crack patterns of cracks
. larger than 0.01 in width were recorded at atmospheric pressure before the test, at the maximum pressure level, and at atmospheric pressure after the f tes The NRC inspector monitored the acquisition of data during the maximum pressure level holding period. Subsequent data analysis determined that the deflection pattern and strain measurements of the containment were within predetermined design acceptance criteria. No reportable cracks were identifie I At the completion of the SIT, the containment was depressurized to 31.8 psig to perform the ILR The preoperational containment ILRT conducted using the Absolute Method (as described in ANSI N45.4-1972, " Leakage Rate Testing of Containment l Structures for Nuclear Power Plants," and ANSI /ANS-56.8 1981, " Containment i
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System Leakage Testing Requirements,") was also addressed during this inspection. The inspection involved procedure and records review, test witnessing, and independent calculations by the NRC inspector. This ILRT was conducted in accordance with approved procedures and satisfied the specified acceptance criteria contained in 10 CFR Part 50, Appendix J,
" Primary Reactor Containment Leakage Testing for Water Cooled Power Reactors," and in the Plant T Preoperational Test Procedure No. 1-RC-P-03, Revision 0, " Containment Integrated Leak Rate Test," incorporates the referenced requirements and
;
criteri This procedure was reviewed by the NRC inspector and no
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discrepancies from the specified requirements and criteria were note The review provided verification that the following test attributes were correctly addressed:
o Containment interior and exterior requirements specified o Instrument locations justified by area surveys o Instrument calibration requirements specified o Instrument loss / test abort criteria delineated o Instrument error analysis performed o Type B and C test results correction to Type A test results specified o Venting of internal isolated volumes required o Isolation valve closing mode specified to be the normal mode o Proper postaccident system alignment to prevent creation of artificial leakage barriers specified o Quality control inspection specified o Test log entries required for repairs needed to complete test o Acceptance criteria specified o Data acquisition requirements specified o Data analysis technique specified o Method of depressurization specified The NRC. inspectors verified that the instrument calibration certifications traceable to the U.S. National Bureau of Standards for the resistance temperature detectors, humidity measuring devices, pressure gauges, and the flowmeter used in the verification test had been reviewed. The guidelines of ANSI /ANS-56.8-1981 were used to select the instruments for the ILRT. The formula from the Instrumentation Selection Guide (ISG) was used during the ILRT to ensure that the data acquisition system accuracy was sufficient to provide reliable test result This formula utilizes the systematic error of each sensor to determine an overall value for the data acquisition syste The instrumentation system for the ILRT was based on a computer controlled data acquisition system capable of reading all sensors rapidly, storing the information and then outputting to the computer for conversions and calculation of the data. Bechtel Calculation No. 2R569MC5887, Revision 0, "CLRT Volume Fractions," provided the
 
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calculational basis for the volumetric distribution of the resistance temperature defectors located throughout the containment. This calculation was reviewed by the NRC inspecto Af ter a period of 24 hours at 31.8 psig to allow for degassing of structures and components inside containment subsequent to the SIT,
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pressurization of the containment for the ILRT commenced. After the
:    internal pressure reached 37.5 psig, the compressors were shut down and
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isolated and the stabilization period commenced. The atmosphere is considered stabilized when the rate of change of containment temperature
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averaged over the last 4 hours minus the rate of change in containment temperature averaged over the last hour is less than 0.5 F/ hour. After the stabilization criteria was satisfied, the ILRT test director declared the start of the official 24-hour test. Continuation of the test indicated convergence of the calculated leak rate and the upper confidence limit below the allowable leakage. At completion of the 24-hour test, the-4-hour superimposed leak verification test was performed. The NRC
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inspector also witnessed this portion of the ILRT and the result between
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the calculated and imposed leakages was found to be within the 25 percent l
allowable leakage (La) limi r 1      .
Subsequent to the performance of the test, the NRC inspector obtained the
!    raw data and computed the leakage rate in accordance with the Mass Point i
Data Analysis technique. The computations performed were compared with
!  the licensee's results for the purpose of verifying the calculational    [
 
procedure and confirming the results. This analytical technique confirmed
.
the acceptability of the results obtained by the license The data
!  providing the as-left values for the type B and C tests were also
,    reviewe No violations or deviations were identified, f Preoperational Test Procedures
,  The NRC inspector reviewed preoperational test procedures which
,
demonstrated the response of the plant's engineered safety features under
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both normal accident and accident coincident with the loss of offsite l    power conditions. The tests were scheduled to be performed in late j  April 1987. The tests were reviewed for compliance with FSAR commitments
!    and adherence to Regulatory Guide 1.68 principles. Within the scope of i    the inspection, the procedures were found to comply with the stated requirements. The specific tests reviewed were:
1-SF-P-01, Safeguards Systems Response - No Blackout 1-SF-P-02, Safeguards Systems Response - Plant Blackout
:
No violations or deviations were identified.
 
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    .__ ___________________ - _ - _
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8 Preoperational Test Results Review The NRC inspectors began the review of results for selected completed preoperational tests. The reviews were done to verify that the testing of systems met specified acceptance criteria, problems encountered during testing were properly resolved, appropriate reviews of tests results were performed by the licensee, and approved administrative controls were followed during the conduct of testing. The tests reviewed included:
1-MS-P-02-01, Main Steam Isolation Valve Logic 1-MS-P-03-01, Main Steam Power Operated Relief Valves and Main Steam Dump 1-PK-P-03-01, IE AC Power Distribution Train C 1-RC-P-01-01, RCS Cold Hydrostatic Test l-  1-RC-P-05-00, RCS Pressurizer Relief Tank l  1-RC-P-13-00, RC Pump Check l
l 4 1-RH-P-01-01, RHR System Train A 1-RH-P-02-01, RHR System Train B 1-RH-P-03-01, RHR System Train C 1-RM-P-01-00, Reactor Makeup Water System 1-RS-P-02-00, Rod Control System 1-SI-P-01-01, SI Train A/B/C and Common Logic 1-SI-P-02-00, SI Accumulators 1-SI-P-04-00, SI Train (A,B,C) Performance 1-SP-P-01-00, Reactor Protection Logic
  '
  , 1-SP-P-02-00, Reactor Protection Master Relay 1-VA-P-02-01, 120V AC Class IE Vital Power Channel II Based on the reviews of the above tests, the NRC inspectors had the
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following comment The NRC inspectors noted a large number of instances in which the
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administrative requirements of Startup Administrative Instruction (SAI) 18, "Preoperational Testing," were not followe The instances included failure to use the test change notices (TCN)
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to correct procedure errors, failure to make chronological log entries for events specified in SAI 18, and a person other than the person making a verification entry in a test procedure correcting that entry. These administrative problems were discussed with the licensee's startup manager and members of his staff. They stated
  }thattherequirementsofSAI18havebeengraduallyputinplace through the five revisions of the procedure, and that future
,
preoperational test procedures should demonstrate stricter adherence to the present administrative requirement In at least two procedures (1-SI-P-01 and 1-PK-P-01), TCNs were generated to change test requirements, but the TCNs made no references to the document or drawing that justified the chang Justification for changing Steps 7.9.67 and -71 of 1-SI-P-01 from 15 to 19.4 and Step 7.5.95.2 of 1-PK-P-01 from black to red is considered an open item pending HL&P providing the needed justification and is identified as 498/8708-0 The acceptance criteria for Steps 7.3.2, 7.3.41, 7.3.44, 7.4.20, 7.6.2, and 7.9.41 of 1-SI-P-04 were not met during testing, and the NRC inspectors had two concerns with the dispositioning of these nonconforming conditions. First, in each case, the test engineer signed-off the step even though the acceptance criteria were not me The NRC inspectors were told by the licensee that the accepted practice is to sign off the step as having been performed whether or not acceptance criteria were met. The NRC inspectors could not find any procedural guidance to support or prohibit this practic In the case of each step specified above, once unacceptable test results were obtained, testing continue SSP-8 requires that for testing to continue after discovering a nonconforming condition, a
'\  conditional release must be obtained from the startup manager and the quality assurance manager or a determination must be made that the nonconformance will not affect continued testing. There was no documentation of either of these conditions being met, and the NRC inspectors could find no procedural guidance concerning who could make the above determinatio The licensee explained that, although not documented, in each case the test engineer made such a determination and, though not proceduralized, a decision on the part of the test engineer was what was intended by SSP-8. The NRC inspectors emphasized the need to properly document the basis for continuing testing. This relates, in part, to a and b abov The NRC inspectors noted that in almost all procedures reviewed, the
   " Witness" blocks, used for verifying the removal of temporary jumpers used for testing, were signed days, weeks, and, in one case, months after the tes This practice raised two question First, how could someone witness jumper removal days after it was performed?
Second, how was the licensee realizing the full intent of the second
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verification of jumper removal if a missed jumper could remain in the system for long periods, potentially invalidating further testing?
The licensee responded to the first question by saying that the initial blocks for jumper removal should have been more properly labelled " Verification" rather-than " Witness." In response to the second question, the licensee will review the practice of delaying verification of jumper remova Two specific cases of contradicting verification dates will require formal resolution. In 1-RC-P-13, QC verification signatures were made 1 day later than the test engineer's verification in Sections 7.5.28-30 and -32, and 7.6.73, .74, .76, .77, .79, and
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  .8 In 1-MS-P-02, QC verification signatures were made 1 day i
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earlier than the test engineer's verification in Sections 7.1, 7.16.32 .39, 7.17.1 .36, 7.19, and 7.2 These two cases of contradicting entries are considered an open item (498/8708-04). Startup Testing Program i
During this inspection period, the NRC inspectors began reviewing the startup testing' program. The NRC inspectors crovided the licensee a listing of 23 Regulatory Guide 1.68 line items which could not be
 
closed. Major issues were as follows:
. Licensee's system for taking credit for preoperational testing and i  some surveillance-tests as coverage for RG 1.68 line items.
 
' The need for additional involvement of operations and Plant
;  Operations Review Committee (PORC) in the evaluation of acceptability
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of preoperational tests for satisfaction of RG 1.68 requirements.
 
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The licensee committed to developing additional information which will be
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tracked as Open Item 498/8708-05.
 
f No violations or deviations were identified.
 
. As-Built Plant to Documentation Reconciliation The NRC inspectors inspected selected systems to determine that they were L  installed in accordance with commitments contained in the FSAR and
'
referenced drawings and specifications. In performing this inspection, mechanical and fluid system walkdowns were performed. Also, draf t TS surveillance test procedures were checked to be sure they could be accomplished for the as-built system. Portions of the controls and instrumentation were verified to conform to the descriptions contained in the FSAR. The following systems were selected for walkdowns:
o Reactor Coolant System (RCS)
 
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o : Chemical Volume Control System o Reactor Water Makeup System o RHR System o .High. Head Safety Injection System o Low Head Safety Injection System o Accumulator Injection System o Containment Spray & Spray Additive Systems o Control Room Emergency Air Cleanup System o Auxiliary Feedwater System o Emergency Diesel Generator Support Systems - Air, Lube Oil, and Fuel Oil o 480 VAC MCC EIA1 o Fuel Building Exhaust (HVAC) System o -Containment Ventilation Subsystems o Essential Cooling Water System Selected systems generally conformed to the FSAR descriptions. For the few discrepancies observed, the NRC inspectors verified that appropriate design change packages existed and action had been taken to update the FSAR drawings and descriptions.
 
; No violations or deviations were observed.
 
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'8. RHR/CCW Water Hammer Incident Description of Incident
; During hot functional. testing (HFT) on Unit 1, the RCS was being
; maintained at 350 F and 350 psig with four reactor coolant pumps in i operation. CCW was in service with all three pumps operating and all
! three RHR heat exchangers on line. Trains A and C of the essential
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cooling water (ECW) were in service supplying only Trains A and C ECW/CCW heat exchangers. Minimal heat loads and mixing of CCW in the i common portions of the system allowed operation of the B train CCW i pump without its associated ECW train in operation. Preparations r
were underway to place Trains B and C RHR in service to support Preoperational Test 1-RH-P-04, "RHR Thermal Performance Test."
 
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The incident occurred while an operator was placing two trains of RHR in service to support the performance of the preoperational tes >-  .The controlling HFT procedure directed that this operation be performed in accordance with Operations Work Order Request (0 WOR) 1-RO-RH-236. This 0 WOR is a modified version of the normal Operating Procedure IP0P02-RH-0001, Residual Heat Removal System Operation, for use during HFT only. While performing this    i procedure, the operator failed to close the RHR Heat Exchanger Outlet    -
Valves (RH-HCV-865/866) as directed by the procedure. This resulted in initiation of flow of hot RCS fluid (350 ) through the RHR Heat Exchangers when the Cold Leg Injection valves were subsequently opene Since CCW to the heat exchangers had been terminated at the verbal direction of the Startup Test Director to enhance heatup of the RHR loops, no cooling medium was provided and rapid heatup and void formation of the CCW in the heat exchangers occurre L  When the RHR Heat Exchanger CCW Outlet Isolation valves were reopened, CCW flow was established to the C Train RHR Heat Exchanger only. This occurred because CCW Pump B had been secured to clear flow alarms on CCW Trains B and C. This resulted in Train B RHR Heat Exchanger being without cooling water for a longer period of tim This prolonged no flow condition permitted a greater degree of heating and void formation in the B RHR Heat Exchanger which contributed to a' greater degree of water hammer on that train.
 
f This incident was caused by the operator's failure to follow procedures in that he skipped two steps in Procedure 1-RO-RH-236 and the Startup Test Director's failure to follow procedure by verbally directing a change.in the approved preoperational test procedures instead of using the authorized TCN method. These actions constitute
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an apparent violation (498/8708-01). This incident resulted in minor water hammer in Train C CCW and a greater water in Train B CCW. The Train B water hammer resulted in damage to pipe hanger, supports, and the CCW piping itself. The system was not breached nor was it overstressed, based on engineering analysis. The licensee took prompt corrective action on problem analysis and system repair Further review of this incident will be documented by the licensee in the final 10 CFR 50.55(e) Report and Station Problem Report.
 
, Technical Review l  The NRC inspector examined two runs of CCW system piping supplying i
the jacket side of RHR Exchanger 1B that was involved in a water hammer incident on March 11, 1987, during a preoperational tes Based on the examination and interviews with licensee personnel, the NRC inspection initially concluded that there was no apparent damage
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to either the piping or connected components. Several pipe supports l
were damaged to varying degrees which have been identified and
;  addressed on nonconformance reports. The licensee reported the event
;  pursuant to 10 CFR 50.55(e) on March 12, 1987. Pending conclusion of
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the licensee's evaluation of this matter, which is to be documented in a report to the NRC, this matter will be considered an open item (498/8708-06).
 
No additional violations or deviations were observe . Operational Staffing The NRC conductad a review of the organizational staffing and staff
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qualifications of personnel assigned to the operational phase for Unit Discussions with appropriate personnel, reviews of organization charts, reviews of 55 personnel resumes, and reviews of certification records indicated that the operational staffing was in accordance with the STP FSAR and the licensee's commitments to ANSI N18.1 with the following noted exceptions: The Manager of Reactor Operations position is vacant. This vacancy and the overall level of nuclear plant operational experience requires further evaluatio The Independent Safety Engineering Group (ISEG) has not been fully staffed. The Director of ISEG has been named; however, the additional four engineers have not been selected. The licensee is committed to the formation of the ISEG prior to fuel loading of Unit The Operational QA organization described in Amendment 54 to the FSAR differs from the in place organization. Discussion revealed that an Amendment 58 to the FSAR is in preparation for submittal to NRR. A review of the draft Amendment 58 revealed the proposed organizational submittal to be in accordance with the existing organizational structur Filling the Reactor Operations position, completing evaluation of operational experience, and implementation of ISEG will be tracked collectively as Open Item 498/8708-0 No violations or deviations were identifie . Training and Qualification Programs The NRC inspectors reviewed the licensee's training programs for licensed operator and nonlicensed staff to verify that regulatory requirements and license commitments are being met or that programs have been developed to implement the training requirements and commitments. The administrative programs to ensure that classroom and simulator training is based on up-to-date training materials that reflect the as-built condition of the plant and approved procedures were also reviewed.
 
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The licensee has clearly established responsibilities for administering the training programs including evaluating, scheduling, assigning
 
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qualified instructors, examining, retraining, and record keeping. The Manager, Nuclear Training oversees the implementation of the training program and is assisted through direct line responsibility by the Training Manager, Operation Training Division; Training Manager, Staff Training Division; and the General Supervisor, Program / Administration Support Section. The NRC inspector noted after discussions with training management personnel that the licensee has established a strong commitment to provide qualified personrel to operate and maintain STP. One example of this commitment is that the licensee is presently greater than 70 percent complete with the program development for the 10 Institute for Nuclear power Operations (INPO) training programs needed for accreditatio The NRC inspectors reviewed the licensed operator training program and verified that the following program elements have been established and implemented when required for:
o new reactor operators (RO);
o upgrading R0 licenses to senior reactor operator (SRO) licenses; o qualifying instructors and shift technical advisors (STAS); and o requalification of R0s and SR0s in accordance with the requirements of 10 CFR 55, Appendix The licensee has demonstrated their commitment to provide an effective
' licensed operator training program as evidenced by the NRC first cold license exams. The results of the exams were that 41 of 46 SRO candidates passed the exam and the only R0 candidate passed the exam. The first requalification cycle for these individuals began in April of this yea The licensee has completed the first cycle training material for the requalification program. The first group of plant equipment operators is presently going through the RO training program. The progress of each license candidate is being evaluated by weekly written and/or oral exam The results of these evaluations are used to determine what additional training the individual requires or if any other actions should be take The licensee also utilizes a fully operational reactor simulator located at the STP site. The presence of the simulator should enhance the overall effectiveness of both the initial and requalification training programs.
 
j l The NRC inspectors reviewed the licensee's program for nonlicensed staff i training and verified that personnel are being trained in the areas of administrative controls, industrial safety, fire fighting, and QA. The licensee demonstrated a strong commitment to the training of nonlicensed
; staff personnel; however, development of the on-the-job training (0JT)
l program has not been completed. The complete development of this program j will require a concerted effort between the applicable user groups and the t training department. The NRC inspectors noted that the licensee has dedicated several training personnel to complete the OJT program in the j
past few months.
 
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Because of the manhours required to complete the 0JT program, the licensee does not believe the task can be completed adequately prior to their projected fuel load date. The NRC staff places great importance on licensees' programs which provide training for personnel in the performance of specific tasks such as complicated surveillance tests, major equipment repair, major plant systems tests, and other special procedures for which the DJT program is an integral part of that training program. Based on the importance of the OJT program, the licensee must provide supplemental training to plant personnel in the area that will be covered by the DJT program until the OJT program has been developed and implemented. The implementation of the supplemental training program pending completion of the OJT program is an open item (498/8708-08). The training of QC inspectors is presently being carried out through the QA department. The training department has planned to take over the training of QA/QC inspectors and to also provide training for engineering support personnel such as ISEG. The development of these training programs will be reviewed during subsequent NRC inspection The NRC inspectors also reviewed the licensees administrative programs for ensuring that classroom and simulator training is based on up-to-date training materials that reflect the as-built condition of the plant and approved operating procedures as well as events or conditions identified by the NRC, INPO, or other facilities. The licensee has established a program for implementing changes to the facility and procedures into the lesson plans. In addition, STP personnel can request training through their supervisor through the use of the " Request For Training Assistance" form, Modifications to the simulator were also reviewed and found to lag only 5 months behind changes to the Unit 1 control panel which demonstrated that plant modifications were being incorporated into the simulato No violations or deviations were identifie ,
1 Reactor Coolant System - Loss of Cleanliness Recovery Program On April 6,1987, the resident inspector accompanied by a regional reactor inspector entered the reactor coolant system to independently verify damage incurred by the resistance temperature detectors (RTDs) and to inspect for screen remnants lost during HFT. During their entrance, the NRC inspectors noted a general disregard of the principles of foreign material exclusion and personnel and material accountability. At that time various work, internal to the reactor coolant system, was being performed via a startup work request (SWR). Rework via an SWR subsequent to system turnover from the construction activity requires the implementation of SSP-22. Section 2.2 of SSP-22 clearly defines the applicability of the procedur Section 5.4.2 of SSP-22 applies to cleanliness controls of
,
systems / components after turnover from the construction activity and requires the following:
_ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
 
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o 5.4. Performance of activities associated with a SWR shall require additional attention to internal cleanliness as outlined within this sectio o 5.4. Upon receipt.of a SWR, the appropriate discipline field engineer (DFE) shall determine the following to preclude entry of debris / contaminants into a system, if activity exposes internal surfaces:
  (a) Specific area controls require (b) Special methods to be utilized for entering the syste (c) Special methods required for maintaining system cleanliness during performance of the activitie The NRC inspectors discovered these requirements were clearly violate The reactor cavity and all hot and cold legs were open to a cluttered construction type environment. Workmen entering the system did not have their personal items and tools lanyarded to themselves and area controls were insufficient. This is an apparent violation of Criterion V of Appendix B of 10 CFR 50, which is implemented by the STP project GA plan Section 5.2.1, in that the licensee failed to follow established procedures for maintaining cleanliness (498/8708-02).
 
This situation was brought to the attention of project management the same evening and management's response was that controls were deliberately dropped because recleaning would have to be performed subsequent to further work on the system. On April 7, 1987, Operations QC was notified of the situation. They agreed that the ongoing work was not in conformance with SSP 22 and issued Nonconformance Report (NCR) SN-03315 on April 8, 198 The NCR reiterated the deviations from SSP 22 and ANSI N45.2.1-1973. On the evening of April 7, 1987, the NRC inspectors further discussed the situation with the project manager, the deputy project manager, and the startup manager. The NRC inspectors acknowledged understanding of the logic behind the decision to abort cleanliness and accountability control, but expressed concerns that cleanliness level B was not formally relaxed and that present cleanliness controls were inadequate and not in compliance with approved procedure No other violations or deviations were observe . TMI and GL 83-28 Action Items The NRC inspector and consultants examined the licensee's conformance with the requirements set forth in NUREG-0737 and GL 83-28. The following describes the NRC's position, observations, conclusions, and current status of the TMI and GL 83-28 action items:
 
  . .
 
(Closed) Generic Letter 83-28 Item 1.1 Post-Trip Review Licensees and applicants shall describe their program for ensuring that unscheduled shutdowns are analyzed and that a determination is made that the plant-can be restarted safel A review was conducted of the applicant's response to item 1.1 of GL 83-28 (Salem ATWS) Post-Trip review process. It was determined by a point by point review, that the STP Procedure OPGP03-ZO-0022, " Post-Trip Review,"
meets the requirements of 83-28 Item 1.1 in all area Item 1.1 of GL 83-28 is considered close (0 pen) TMI Item II.K.1.10 Operability Status of Safety-Related Systems Review and modify as necessary your maintenance and test procedures to ensure that they require:
o Verification, by test or inspection, of the operability of redundant safety-related systems prior to the removal of any safety-related system from service o Verification of the operability of all safety-related systems when they are returned to service following maintenance or testing o Explicit notification of involved reactor operational personnel whenever a safety-related system is removed from and returned to service The following procedures were examined to see if they fulfill the requirements stated in the NRC position:
o OPGP03-Z0-0004, Part 4.10. o OPGP03-Z0-0001, Part 5.2. o OPGP03-ZO-0004, Part 4.1 o OPGP03-Z0-0003, Parts 4.10.3 and 4.10.23
.
l The examined procedures meet the intent of the NRC position, with one
'
deficiency. The wording in OPGP03-ZO-0004 and OPGP03-ZO-0001 uses
"should" where a stronger "shall" would be more appropriate since the verification in question is mandatory, not optional. Until the wording l choice of "should" rather than "shall" is resolveu, this is an open item i (493/8708-09).
 
! (Closed) TMI Item I.C.6 Verification of Correct Performance of Operating Activities i
The applicant will have a procedure for verifying the correct performance of operating activities. The Shift Supervisor, or, in the absence of the Shift Supervisor, the Unit Supervisor (an SRO) should be responsible for releasing equipment for testing, maintenance, or modifications. Following
 
, .
 
such activities, a qualified person from the shift crew (who does not have to be a licensed operator) should be assigned to independently verify the proper positioning of valves, circuit breakers, and control switches of the systems that are important to safety, o STP Procedure OPGP03-ZA-0010, Revision 2, " Plant Procedure Compliance,' Implementation, and Review," sets forth the methods and requirements for ensuring proper procedure compliance and independent verification o STP Procedure OPGP03-ZA-0039, Revision 3, " Plant Procedures Writers Guide,"_ provides the guidelines for writing a procedure to include the proper precautions (e.g. permissions required and independent verifications).
 
o STP Procedure OPGP03-Z0-0005, Revision 0, " Reactor Operations Division Conduct of Operations," was reviewed. The review indicated the Shift Supervisor and Unit Supervisor have proper authority to release equipment for testing, maintenance, or modification Based on the above information it is concluded that the applicant has met the requirements for Item I. Therefore, Item I.C.6 is considered close (Closed) TMI Item I.C.5. Feedback of Operating Experiences Each applicant for an operating license shall prepare procedures to assure that operating information pertinent to plant safety originating both within and outside the utility organization is continually supplied to the operators and other personnel and incorporated into training and retraining program o An interview was conducted with the Lead Engineer of the STP Regulatory Compliance Group. During the interview, he displayed the methods used to track, classify, and route inside and outside events to the various facility departments according to STP Interdepartmental Procedure IP-2.2Q, Revision 2, " Operating Experience /In-House Experience Review." He also displayed, from the License Commitment Tracking System (LCTS), how the operations department responds to action items of the LCT o The Training Department procedures for LCTS action items was described by the Manager of Operator Training. The methods used by the Training Department cover both initial training and retraining of operator On the basis of the above information, it is concluded that the applicant has coa: plied with NUREG-0737 Item I. Therefore, Item I.C.S. is considered close . .
 
(Closed) TMI Item I.C.8 Pilot Monitoring of Selected Emergency Procedures for Near-Term Operating License Applicants Applicants will be required to correct any deficiencies identified by an NRC sample audit of selected emergency operating procedure (EOPs) before full power operatio o STP SER (NUREG-0781) Section 13.5.2.3 states, "This pilot monitoring program was used on an interim basis for evaluation of applicant's E0P's before staff approval of generic technical guidelines and staff development of the long-term program for the upgrading of E0P' This is no longer necessary because the NRC has approved the Westinghouse Emergency Response Guidelines (ERG's) and the applicant has committed to develop E0P's based on the ERG's."
 
o The applicant has in place all applicable E0Ps, as recommended by the Westinghouse ERGS. The effective dates for these procedures are 01-05-87 (42 procedures) and 02-23-87 (6 procedures).
 
o Four STP E0Ps were reviewed and found to be in compliance with the format, intent, and content of the Westinghouse ERGS. These procedures are:
o IPOP05-E0-E000, Revision 1, " Reactor Trip or Safety Injection" o IPOP05-E0-E010, Revision 1, " Loss of Reactor or Secondary Coolant" o 1 POP 05-E0-ES02, Revision 1, " Natural Circulation Cooldown" o 1 POP 05-E0-EC00, Revision 1, " Loss of All AC Power" Item I.C.8 is considered close [0 pen) TMI Item I.D.1 Control Room Design Review All licensees and applicants are required to conduct a detailed control room design review (DCRDR) to identify and correct design deficiencie o STP Safety Evaluation Report (SER), NUREG 0781, states in Section 18 that the applicant shows a commitment to comply with the requirements of this ite However, the SER further states that to complete the DCRDR activities, the following items must be resolved:
o Provide the results of the verification and validation program for the final E0Ps to confirm that the instrumentation and control needs have been adequately identified and satisfie o Provide the results of the investigation of the green Roto-tellite indicating lights in the control room under actual operating condition .- .. - - - .-. -
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ARLINGTON, TEXAS 76011
o Provide the results of the surveys of the lighting, sound, meter, and communication system, when planned work in the control room is complete o A letter from M. R. Wisenburg, STP Deputy Project Manager to
    - JUL k. 4 19ef In Reply Refer To:
. Vincent S. Noonan of the NRC Pressurized Water Reactor (PWR) Project Directorate No. 5, dated December 26, 1986, (ST-HL-AE-1864) contains Addendt.m 1 to the Human Engineering Deficiency (HED) Resolution Report and Addendum 2 to the Executive Summary. These documents include the results of the green Roto-tellite indicating light investigations in the control room under actual operating conditions (Pages A-7 and A-8 of the HED report and Page S-1 of the Executive Summary) and conclude that, after modifications, the visibility of these lights under actual operating conditions is acceptabl o A letter from M. R. Wisenburg, STP Deputy Project Manager to Vincent S. Noonan of the NRC PWR Project Directorate No. 5, dated December 23,1986, (ST-HL-AE-1860) contains the initial submittal of the E0P Validation Report. Control panel deficiencies and problems identified during the E0P validation each have an associated resolution identified with them. E0P problems encountered during the validation process are documented, and their resolutions are
Dockets: 50-498/87-08 50-499/87-08 Houston Lighting & Power Company ATTN: J. H. Goldberg, Group Vice President, Nuclear P. D. Box 1700 Houston, Texas 77001 Gentlemen:
  .
<
addressed in the STP Procedures Generation Package (PGP).
Thank you for your letter of June 25, 1987, in response to our letter and l
 
NoticeofViolation(498/8708-01) dated May 29, 1987. We have reviewed your reply and find it responsive to the concerns raised in our Notice of Violatio We will review the implementation of your corrective actions during a future inspection to determine that full compliance has been achieved and will be l maintaine Sincere f,  l l
Item I.D.1 remains open (498/8708-10) pending submittal of the results of the surveys of the lighting, sound, meter, and communication system
, referred to in the STP SER.
 
l (Closed) TMI Item I.A.1.1 Shif t Technical Advisor (STA)
Each licensee shall provide an on-shift technical advisor to the Shift
: Supervisor. The STA may serve more than one unit at a multiunit site if
'
qualified to perform the advisor function for the various units.
 
I o STP SER 13.2.2 provided the conclusion that the-STA Training program
', was acceptable.
 
l o A review of STP Station Procedure OPGP03-ZO-0008, Revision 0, " Shift
'
Technical Advisor," outlining the duties, responsibilities, training, experience, and retraining, was conducted and found to satisfy the
; requirements of NUREG 0737.
 
! o An interview was performed with an STP Reactor Engineer qualified as
: STA to determine the scope of STA training, present functions, and
'
the staffing levels of the STAS (present and near future). The interview provided information that the STA program is following the training program and station procedure Item I.A.11 is considered closed.
 
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.( Closed) TMI Item I.A.2.1.4 Immediate Upgrade of R0 and SRO Training and Qualifications - Modify Training Licensees and applicants shall review their training programs and upgrade them, as necessary, to include items relating to TMI-2 lessons learne Any necessary training program modification must be in place prior to fuel loa o STP SER (NUREG 0781) concludes in Section 13.2.1.3 that the STP training program meets the requirements of I.A. o STP TS 6.4.1 requires that the training program shall include the requirements of Sections A and C of Enclosure 1 of the March 28, 1980, NRC (H. R. Denton) letter to applicants and licensees concerning qualification of reactor operators, o STP Interdepartmental Procedure IP-8.8, Licensed Operator Training Program, stipulates operator and instructor selection criteria and licensed operator training requirement These were reviewed and found to be consistent with the requirements of STP TS 6. o STP Interdepartmental Procedure IP-8.9, Licensed Operator Requalification, provides instruction for the conduct of licensed SRO and RO requalification. This procedure was reviewed and found to be consistent with the requirements of TS 6. Item I.A.2.1.4 is considered close [Open) TMI Item I.C.4 Control Room Access Licensees are to assure instructions are in place which cover the authority and responsibilities of the person in charge of control room access, and establish clear lines of authority and responsibility in the control room during emergencies, o STP' SER (NUREG-0781) Section 13.5.1.2 states that the applicant has committed to limit access to the control room, o STP Procedure OPGP03-ZO-0005, Revision 0, Reactor Operations Division Conduct of Operations, Section 4.1, was reviewed with the following finding o Section 4.1.1 states that if the Unit or Shift Supervisor deems the number of people in the control room to be excessive, he has authority to direct excess people to leave and may require personnel who need further entry to obtain his prior approval, o Section 4.1.2 states that the individual screening for control room access shall be controlled by the security syste __ _-
. .
 
o Section 4.1.3 states that during a reactor trip or other plant abnormality "it is recommended that all personnel not directly involved with the recovery leave the control room. Assistance from other persons shall be determined and requested by the Shift Supervisor, Unit Supervisor, or the Reactor Operator."
 
The language used in Section 4.1.3 of OPGP03-Z0-0005 is not strong enough to " establish clear lines of authority and responsibility in the control room during emergencies." Conspicuously absent are directions which require all nonessential personnel to leave the control room during emergencies; instructions which stipulate who, specifically, is responsible and authorized to control access during emergencies; and instructions which specify who, specifically, may enter the control room during emergencies (e.g. licensed operators, the STA, the NRC resident inspector). Item I.C.4 remains open (498/8708-11) pending clarification of control room access requirements during emergencie (Closed) TMI Item I.C.2 Shift and Relief Turnover Procedures Licensees are to revise plant procedures for shift and relief turnover to ensure that each oncoming shift is made aware of critical plant status information and system availability, o STP procedure OPGP03-ZA-0063, " Reactor Operations Shift Turnover,"
was reviewed and found to meet the intent of NUREG 0694, Item I.C.2, as the applicant had committed to in the STP SER Item 13.5. o Direct observations were made of the shift relief process, including discussion with on-duty R0, as to how the turnover check lists were used and process of turnover accomplished. In addition, the on-coming crew briefing was observed. The observations indicated that the procedure noted above is being used to accomplish shift turnovers and personnel are being made aware of system availability and critical plant status information prior to assuming control of the plan .C.2 is considered close (Closed) TMI Item I.C.7 NSSS Vendor Review of Procedures Applicants are required to obtain reactor vendor review of their low power, power ascension, and emergency procedures as a further verification
 
of the adequacy of the procedure o STP SER (NUREG-0781) 13.5.2.3 was reviewed and states that because the applicant has committed to implement procedures based on the NRC approved ERGS, the staff does not consider an additional NSSS vendor review of the E0Ps necessary. Furthermore, this section states that STP committed to have the low power testing procedures and power ascension testing procedures reviewed by Westinghous _ _
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I o Four STP E0Ps were reviewed and found to be in compliance with the format, intent, and content of the Westinghouse ERGS. These procedures are:
i o IP0P05-E0-E000, Revision 1, Reactor Trip or Safety Injection o IPOP05-E0-E010, Revision 1, Loss of Reactor or Secondary Coolant
,
o 1 POP 05-E0-E502, Revision 1, Natural Circulation Cooldown
; o IPOP05-E0-EC00, Revision 1, Loss of All AC Power
,
o The Lead Engineer, STP Regulatory Compliance Group, indicated after
 
conversations with an STP Technical Support Supervisor that all IPEPO4-series procedures (Initial Startup Test Procedures) were reviewed by Westinghouse. This conforms to the requirement in STP Procedure OPGP03-ZA-0002 Plant Procedures, Addendum 2, Step 3, which states that Westinghouse shall be designated to review and comment on the initial issuance of the initial startup test procedure o An STP Reactor Engineer indicated in conversation with the Lead Engineer, STP Regulatory Compliance Group, that all, or parts of, the following procedures were submitted to Westinghouse for review and comment, based upon design features which are unique to STP:
o IPOP05-E0-FRC1 Response to Inadequate Core Cooling
,
o IPOP05-E0-FRC2 Response to Degraded Core Cooling l o IP0P05-E0-F002 Core Cooling Safety Function Status Tree a  o IP0P05-E0-E000 Reactor Trip or Safety Injection j  o IPOP05-E0-ES13 Transfer to Cold Leg Recirculation
'
Additionally, he indicated that informal constructive interchange between STP and Westinghouse was ongoing thrauchout the development of the E0Ps.
 
i TMI Item I.C 7 is considered close ,
(0 pen) TMI_ Item _II.E.1.2 Auxiliary Feedwater System Automatic Initiation
! and Flow Indications i
, The applicant will provide an auxiliary feedwater system (AFWS),
, initiation and flow monitoring capabilities.
 
.
. o A review was conducted of the STP SER. The staff concludes that the l  AFWS meets the guidelines of NUREG-0737 concerning reliability and
;
that the AFWS is compatible with staff guidance for unavailability per Standard Review Plan (SRP) Section 10.4.9.
 
.
,
o An interview was conducted with two STP startup engineers (SEs) to i determine the functional capabilities demonstrated by the AFWS system
! during the recent HFT. Both SEs stated, when questioned, that the AFWS logic, power supplies, flow indication, and valves functioned as j designed. However, the integrated safeguards test (IST) which will i
!,
 
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  ( 24 complete the automatic circuitry testing in regard to loss of power, auto sequenci19, and automatic safety injection actuation remains to be completed, o A review was conducted of the training department lesson plans and system descriptions involving the AFWS to determine if NUREG-0737 items were addressed. The training department's lesson plans covered all areas of the AFWS design, operation, and procedures. The depth of training was describe!d as systems training, classroom work, and simulatortrainingbythhManagerOperationsTrainingandSupervisor Operations Trainin ;
Item II.E.1.2 remains open (498/8708-12) pending the completion of the IST and submittal of results and acceptance by NR (Close_d) TMI Item II.E.3.1 Emergency Power for Pressurizer Heaters / Upgrade Power Supply Applicants shall provide the capability to supply, from either the offsite power source or the emergency power source (when off site power is not available), a predetermined number of pressurizer heaters and associated controls necessary to establish and maintain natural circulation at hot standby conditions, o STP Procedure IPOP05-E0-E000, Revision 1, states the conditions when the pressurizer heaters are to be loaded onto the emergency diesel o An interview was conducted with three STP SEs. They indicated that the pressurizer heater modifica tion was installed and functioning properly on the Engineered Safety Features (ESF) power supply, o STP Plant Electrical Drawing 9-E-PLAA-01, Revision 6, and 9-E-PLAA-01, No. 1, Revision 9, " Single Line Diagram 480V Class IE Load Center," both show heater groups power coming from ESF supplie:.,
Heater Group "A" from Bus EIA1, and Heater Group "B" from Bus E1C o A review of the training departnent lesson plans indicated that the training department has ensured operators have received proper training on the pressurizer heater On the basis of the above information Item II.E.3.1 is considered close (Closed) TMI Item I.A.1.2 Shift Supervisor Responsibilities - Delegate Non-Safety _ Duties Administrative functions that detract from, or are subordinate to, the management responsibility for assuring the safe operation of the plant are to be delegated to other operations personnel not on duty in the control roo _
. .-
 
o Section 13.1 of the STP FSAR, Organizational Structure of Applicant, was reviewed. It was found that provision is made in the STP organization for an on-shift administrative aide who reports functionally to the shift supervisor of the assigned work shif o The job description applying to the administrative aide position was reviewed and found to include administrative responsibilities which should significantly reduce the burden on the shift supervisor caused
.
by ancillary responsibilities.
 
!
o Four individuals are currently filling Administrative Aide positions on shift. The Manager of Administrative Services stated that these ( individuals are assigned to a three-shift rotation such that an Administrative Aide is assigned on-shift at all time Item I.A.1.2 is considered close (Closed) TMI Item II.K.1.17 Trip Per Pressurizer Low Level Bistable Facilities which use pressurizer pressure for automatic initiation of safety injection into the reactor coolant system must trip the low pressurizer level setpoints such that, when pressurizer pressure reaches the low setpoint, safety injection will be initiated regardless of the pressurizer leve o Drawing No. 5Z-10-9-Z-42112, Revision 7, "SSPS Logic Diagram," was reviewe It was ascertained from this review that STP does not utilize pressurizer water level coincident with pressurizer pressure for automatic initiation of safety injection. When any two of four channels of pressurizer pressure decrease to a predetermined value which corresponds to the low pressurizer pressure safety injection setpoint, automatic safety injection will be initiated. Therefore, TMI Item II.K.1.17 does not apply to ST TMI Item II.K.1.17 is considered close (Closed) TMI Item I.C.3 Shift Supervisor Responsibilities - Corporate Directive Applicants and licensees shall revise their procedures to assure that duties, responsibilities, and authority of the shift supervisor and control room operators are properly defined, o STP SER (NUREG-0781) concludes in Section 13.5.1.1 that the applicant's plans for organization and conduct of the operating shift crews meet the requirements of 10 CFR 50.54 and NUREG-0694 Item I.C.3. It is the position of the staff, however, that a letter, signed by the Vice President - Nuclear Plant Operations, be reissued on an annual basis to all station personnel directing that each shift
 
. .
 
supervisor has the responsibility of directing the licen ed activities of licensed operators on the supervisor's shift, pursuant to 10 CFR 50.54(1).
 
o STP Procedure OGP03-ZO-005, " Reactor Operations Division Conduct of Operations," Revision 0, was reviewed and found to include the duties, responsibilities, and authorities of all licensed on-shift personnel, o A letter from J. H. Goldberg to all Nuclear Group Personnel, dated February 18,1987, " Command Authority for Direction of Licensed Operations Activities South Texas Project Electric Generating Station," was reviewed. This letter was found to contain explicit and clearly defined direction concerning the command authority and responsibility with which the cognizant shift supervisor is charge TMI Item I.C.3 is considered close (Closed) TMI Action Item I.C.1 Guidance for the Evaluation and Development of Procedure for Transients and Accidents The licensee should reanalyze and propose guidelines and revise procedures for small break (SB) loss of coolant accident (LOCA), inadequate core cooling, and transients and accident o STP FSAR 13.5.2.1.4 indicates the Westinghouse Owner's Group (W0G)
and ERG would be the basis for STP E0P o SER 13.5.2.3 and NUREG-0737, Supplement 1, identify the need for and current staff review of PG The staff response will be in a subsequent supplement to the SER (NUREG-0781).
 
o SSER 1 and 2 indicate that the staff response to I.C.1 is still pendin o I.C.1.1.1, Small-break LOCA. The STP Station Procedure IPOP05-E0-E000, Revision 1, approved January 2, 1987, by the plant manager is " Reactor Trip or Safety injection," and is said to include the SB LOCA
'
response. Procedure IPOP05-E0-E000, Revision 1, does indeed include SB response such as verify AFW, leck PRZR PORV, check if steam generator (S/G) tubes are intact, verify RCS subcooling margin, verify secondary heat sink, and check for reactor coolant pump (RCP) trip.
 
; There are contingency actions identified of the check results in negative findings, for example, can't confirm secondary heat sin There are cautions to alert the operator to the need for designating the condition as a site emergency condition when certain criteria are met; and there are actions and contingency actions to support maintenance of critical safety functions.
 
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o I.C.1.2.B. Inadequate Core Cooling. The STP Station Procedures for inadequate core cooling are:
o IP0P05-E0-F002, Revision 1, " Core Cooling Safety Function Status Free," approved February 14, 1987, by the plant manager, o IPOP05-E0-FRC1, Revision ^1, " Response to Inadequate Core Cooling," approved January 2, 1987.
 
l o IPOP05-E0-FRC3, Revision 1, " Response to Saturated Core l  Cooling," approved January 2, 198 o 1 POP 05-E0-FRC2, Revision 1, " Core Exit TC's Less Than 755 degrees F."
 
The Core Cooling Safety Function Status Tree indicates which procedure to activate depending on core exit thermocouples (TCs) less than 1200 F, RCS subcooling, RVWL upper plenum, and core exit TCs less than 755 F. The " Response to Inadequate Core Cooling" procedure meets the requirements of GL 82-33, Supplement 1 to NUREG-0737,
" Requirements for Emergency Response Capabilit o Per GL 82-33, STP has developed " Plant Procedures Writer's Guides,"
for E0Ps as follows:
o OPGP03-ZA-0027, Revision 1, " Emergency Operating Procedures Preparation, Approval, and Implementation," approved May 15, 1986, by the plant manage o OPOP01-ZA-0006, Revision 2, " Emergency Procedures Writer Guide and Verification," approved August 14, 1986, by the plant manage o Reviewed ST-HL-AE-1848, " Response to NRC Comments on the Procedures Generation Package," in which an attachment demonstrated how OPGP03-ZA-0027, Revision 1, (see 6 above) was used to achieve a completed procedure for " Steam Generator Tube Rupture,"
OPOP09-E0-E03 o Per GL 82-33, NRC approval of the PGP is not necessary prior to upgrading and implementing the E0Ps (7.2.b.).
o HL&P All Sets and Volume Procedure Listing indicates that 48 individual E0Ps have been developed and approved to comply with I.C.1.3.B. " Transients and Accidents Procedure Revision."
 
Based on the results of the above review the Item I.C.1 is considered close l
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l (0 pen) TMI Item I.A.1.3 Shift Manning Applicants shall include in their administrative procedures provisions governing required shift staffing to assure that qualified personnel are readily available to man the operational shifts in the event of an abnormal or emergency situatio These procedures shall also set forth a policy, the objective of which is to operate the plant with the required ,
staff and develop working schedules such that the use of overtime is j avoided, to the extent practicable, for those persons who perform '
safety-related functions (e.g. , SR0s, R0s, health physicists (HPs),
auxiliary operators (A0s), and key maintenance personnel),
o STP Procedure OPGP03-ZO-000S, Revision 0, " Reactor Operations Division Conduct of Operations," was reviewed and found to be in compliance with the staffing and overtime requirements of NUREG-0737, Item I.A.1.3. Minimum staffing requirements are delineated for all modes of plant operation including refueling operations, and direction is stipulated which effectively limits the amount of overtime authorized to be worked by any individual. TMI Item I.A.1.3.2, Implement Minimum Shift Crew Requirements, is considered close o Interviewed the Manager of Management Services Business Support Group on March 17, 1987. He indicated that a system for tracking individual overtime to ensure limits are not exceeded is not yet in place. A procedure for which he is responsible, OPGP02-ZA-0060,
" Overtime Approval Program," is being typed and should be approved for issuance in the near future. This procedure, when implemented, will facilitate tracking of overtime for SR0s, R0s, HPs, A0s, and key maintenance personne TMI Item I.A.1.3.1, Limit Overtime, will remain open (498/8708-13) pending implementation of a verifiable system for supervisors and other responsible personnel to track individual overtime worked, and thereby limit overtime usage to within specified maximum requirement (0 pen) TMI Item I.D.2 Plant Safety Parameter Display Console Each applicant and licensee shall install a safety parameter display system (SPDS) that will display to operating personnel a minimum set of parameters which define the safety status of the plant. This can be attained through continuous indication of direct and derived variables as necessary to assess plant safety status, o Reviewed Letter ST-HL-HL-36428, dated February 12, 1987, concerning an audit by NRC of the STP SPDS. This audit lists five deficiencies (significant findings) in the SPDS. They are:
o A validation of the capability of the SPOS to rapidly assess the safety status of the plant should be performed, preferably before fuel loa O e
 
o Parameters to determine the status of the (NUREG-0737)
Radiological Control Critical Safety Function (CSF) should be included in the SPD o Administrative controls sitould be placed on modification to the SPDS software to insure that the system's capability to provide a rapid and reliable assessment of plant safety status is not jeopardize o The SPDS should provide continuous display of the status of the CSF o No formal review of system requirements versus NUREG-0737 requirements was performe o Reviewed Letter ST-HL-AE-1934, which is HL&P's response to NRC concerning noted deficiencies in the SPDS. All of the responses to audit concerns will require further review by HL& Based upon the two letters above, concerns raised from an audit of SPDS have not yet been resolved. This item will remain open (498/8708-14)
until the SPDS audit concerns are acceptably resolve (Closed) TMI Item II.K.3.12 Confirm Existence of Anticipatory Reactor Trip Upon Turbine Trip Licensees and applicants must confirm that their plants have an anticipatory reactor trip upon a turbine trip. The licensee for any plant where this trip is not present should provide a conceptional design and evaluation for the installation of this tri o STP SER (NUREG-0781) concludes in Section 7.2.2.4 that the STP design, which includes an anticipatory reactor trip on a turbine trip above 50 percent of rated thermal power (P-9 interlock) is in compliance with the TMI Action Plan Guideline o Drawing No. 5Z-10-9Z-4211 " Reactor Trip Signals Logic Diagram" was reviewed and found to include logic which feeds the automatic reactor trip circuitry whenever turbine EHC fluid pressure drops below a predetermined setpoint (2/3 coincidence) or all four turbine throttle stop valves are closed (2/4 coincidence), and reactor power is above a predetermined setpoint (P-9) as sensed by Nuclear Instrument Channels N41, N42, N43, and N44 (2/4 coincidence).
 
TMI Item II.K.3.12 is considered close (0 pen) TMI Item I.B.1.2 Evaluation of Organization and Management Applicants shall establish an on-site ISEG to perform independent reviews of plant operation ._ _ _ _ _ _ . .- _ . _  _ .. _-
I
, .
 
o STP SER (NUREG-0781) concludes in Section 13.4.4 that the HL&P organizations which perform review and audit functions for STP (including ISEG) are in conformance with applicable guidelines and standards (including NUREG-0737).
 
o STP TS (proof and review copy) Section 6.2.3 regarding ISEG was reviewed and found to be in conformance with the requirements of NUREG-0737, Item I.B.1.2.
 
l o STP ISEG Operating Philosophy Document (OPD), dated February 9,1987, f was reviewed. This document, prepared by the Director - ISEG, and
'
approved by the Chairman - Nuclear Safety Review Board (NSRB),
 
delineates how the ISEG plans to carry out its function It does not, however, present an item-by-item discussion of how the group satisfies every stated requiremen The following ISEG OPD inclusions, relevant to NUREG-0737, Item I.B.I.2, were found to be in conformance with the requirements of Item I.B.1.2:
o Staffing and Reporting Relationship - The ISEG will have a multi-disciplimary, five-member technical staff located at the STP site. The ISEG Director reports to the chairman of the NSRB, located in Houston, Texas. This reporting relationship
;  provides the ISEG with a high-level forum for the review of its i
'
determination o ISEG Functions - The ISEG will review the operations at STP with
;  particular emphasis on assessing the activities which impact i  nuclear safety. These assessments will carry recommendations for NSRB consideration which will focus on the root causes of
'
events, problems, or undesirable trend The ISEG will maintain appropriate relationships with other similar operating plants and will participate in industry-sponsored groups which bring together utility personnel performing functions similar to the STP ISE o On March 16, 1987, the Director - ISEG, was interviewed. He provided information indicating that the specific procedures relative to actual performance of ISEG duties have yet to be written, and that four full-time engineers have yet to be integrated into the ISEG to satisfy the minimum staffing requirement of NUREG-0737 Item I.B. At that time, the ISEG will be considered fully operationa TMI Item I.B.1.2 remains open (498/8708-15) pending completion of the following two items which will make the ISEG operational:
o Attaining a minimum staff of five dedicated, full-time engineers (including the Director - ISEG), located onsite at ST l o The issuance of approved procedures which specifically address the
: responsibilities and duties of the ISE {
!
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      -
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(Closed) TMI Item II.B.1 Reactor Coolant System Vents Each applicant shall install RCS and Reactor Vessel Head High point vents remotely operated from the control room, o STP FSAR Appendix 7A addresses NUREG-0737, Item II.B.1, in regard to safety analysis of design for seismic and LOCA events, with 10 CFR 50.44 and 10 CFR 50.46 as the acceptance criteri o STP SER Section 5.4.12 concluded that the STP design of the RCS High Point Vents was satisfactory and meets the requirements of NUREG-073 o STP procedure IPOP05-E0-FRI3, Revision 1 " Response to Voids in Reactor Vessel," was reviewed to determine the degree of guidance provided to the operator. This procedure does provide necessary information to the operator for initiating and terminating vent usage in emergency condition o STP Procedure 10P02-RC-0003, Revision 1. " Filling and Venting the Reactor Coolant System," and 10P03-ZG-0001, Revision 2, " Plant Heatup," provide procedural steps for vent valve lineup and operations when cold, o STP TS 3.4.11 provides the operability requirements of the Reactor Vessel Head Vent System (RVHVS).
 
o STP Orawing SR149F05001, Revision 6, shows the piping and valve arrangement of the RVHVS. This arrangement meets the design requirements as stated in the STP FSAR and was accepted by the staff in the STP SER, o The following electrical one-line diagrams were reviewed; 9-E-DJAA-01, Revision 8; 9-E-DJAC-01, Revision 8; 9-E-DJAE-01, Revision 10; and 9-E-RC19-01, Revision 4. From these drawings it was determined that the RVHVS is supplied power from a Class IE power suppl o The HL&P Company " Pump and Valve Inservice Test Plan" lists the RVHVS valves with information provided as to class of valves, category of system, fati position, normal position, and test requirement o A tour of the Unit 1 control room was made to determine if the RVHVS could be operated from the control panels. Indication and control is available from the control panel Item II.B.1 is considered close _
. .
 
(0penj TMI Item _II.B.4 Training for Mitigating Core Damage
_
_
Applicants are required to develop a training program to teach the use of installed equipment and systems to control or mitigate accidents in which the core is severely damaged. They must then implement the training progra o STP FSAR Appendix 7A states that training is provided as described in FSAR Section 13.2. o The staff concludes in the STP SER that the applicant has met item II.B.4, which includes all subjects in Enclosure 3 of H. R. Denton's letter of March 28, 1980. Also, the STAS and all operating personnel including licensed operators, appropriate managers, instrument and control (I&C) technicians, HP technicians, and chemistry technicians shall receive training commensurate with their responsibilities, o A review of two lesson plans (CLT 006.03, Revision 0, Small Break LOCA and CLT 006.02, Revision 0, Post Accident Cooling) used to teach Mitigating Core Damage to licensed operators (plant manager through the operations chain) was conducted. The lesson plans met the requirements of NUREG-073 o A telephone interview was conducted with the Staff Training Manager to determine the training provided to the I&C technicians on Mitigating Core Damage. He stated that the I&C technicians receive training from Lesson Plan ICT9.21. This lesson plan adequately covers the requirements for the !&C technician o An interview was conducted with the Radiation Protection Supervisor concerning training of HP personnel. He provided a completed, but unapproved, lesson plan, RPT001.61, " Radiological Aspects of a Core Damage Accident." The lesson plan coverage is adequate, but the actual implementation time was not available. In addition, he provided lesson plan CATTP phase 3, " Radiological Aspects of a Core Damage Accident," for the chemistry technician, which is implemented for all chemistry technicians and is adequate in coverag Item !!.B.4 is to remain open (498/8708-16) pending verification that the HP training has been implemente f(Closed)T$ystemTMI soIFio  Item !!.K.3.1.8 Testing /_ Installation __of Automatic PORV The applicant must provide a system which will automatically cause the power operated relief valve (PORV) block valve to close when the RCS pressure decays after the PORV has opened, o FSAR Appendix 7A, Section II.K.3.1, states that automatic PORV block valve closure is not required in the STP desig The basis for this
 
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I    33 i
is found in a study performed by Westinghouse for the WOG in response i to TMI Item II.K.3.2 - Report On Overall Safety Effect of Power-Operated Relief Valve Isolation System. The results of this study, WCAP-9804, concluded that with the incorporation of specific post-TMI modifications, which have been or will be impleme.ited on STP, the reduction in PORV of LOCA frequency is such that an automatic PORV block valve closure system is not require o STP SER (NUREG-0781) concludes in Section 15.6.1 that WCAP-9804 is acceptable on a generic basis, and that the STP design, hardware, and j
PORV setpoint is acceptabl '
TMI Item II.K.3.1.B is considered close (Closed) TMI Item II.E.4.1 Dedicated Hydrogen Penetrations Plants using external hydrogen recombiners or purge systems for post-accident combustible gas control must have containment penetrations dedicated to that service only, o STP design includes two redundant, 100 percent capacity electric hydrogen recombiners located inside the containment. Therefore, the requirement for dedicated penetrations for this system does not apply to STP.
 
; o Procedure IPOP-CG-001, Electric Hydrogen Recombiners is in place, which addresses post-LOCA startup, verification of operation, placing i in standby, and returning the recombiners to pre-LOCA conditio ,
      ,
o The following procedures, which contain steps directing operation of I the hydrogen recombiners when containment hydrogen concentrations j reach specified levels, are in place:
o 1 POP 05-E0-FRI3, Revision 1 - Response to Voids in Reactor Vessel
      '
o 1 POP 05-E0-FRC1, Revision 1 - Response to Inadequate Core Cooling o 1 POP 05-E0-FRZ1, Revision 1 - Response to High Containment Pressure o In Procedure 1 POP 05-EO-FRZ1 Steps 7.3 and 10.2 both erroneously
.
'
refer to Procedure 1 POP 02-CM-0001. This procedure does not exist; the references should be to IPOP02-CG-000 o In Procedure 1 POP 05-E0-FRC1, Step 8.3 erroneously refers to Procedure 1 POP 02-CM-0001. Again, this procedure does not exist; the reference should be to Procedure 1 POP 02-CG-0001.
 
*
TMI Item II.E.4.1 is considered closed. Open item (498/8708-17) will i track the procedural errors noted above, it should be noted that the j
I l
      ,
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_ _ _ _ _ _ _ _ - .- - _ . - ._
  . .
 
.
described error exists in all three copies of each procedure; i.e.,
supervisor's copy, primary operator's copy, and secondary operator's cop [
(0 pen]TMIItemII.E.4.2ContainmentIsolationDependability Applicants must comply with the Containment Isolation Dependability by providing improved diverse isolation, minimum containment pressure setpoints for isolation, containment purge valve changes, and closure of purge valves on high radiation, o STP SER was reviewed with the following acceptable conclusions of the i
staff: "the applicant has complied with the provisions regarding i diversity in parameters sensed for initiation of containment i isolation, identification of essential and nonessential systems,
, automatic isolation of nonessential systems, and closure of
{ containment purge and vent isolation valves on a high radiation
! signal."
 
. o By a letter dated October 30, 1985, the applicant committed to equip
!' the outboard isolation valves for supplementary purge system supply and exhaust lines with pneumatic operators. Configuration Control l Package (CCP) No. IMST0115 was reviewed to determine the status of
; valve installation. The CCP shows a Field Notification of Change Completion dated March 4, 198 o A visual check of the plant showed that the valves stated above are installed in the system indicated.
 
,
,
o STP TS were reviewed to verify the required position of the 48-inch i
    > . MM K E. G gliardo, Chief
'
  ..//V Reacto Projects Branch l
containment purge valves during power operation, startup, hot standby, and hot shutdown. TS 3.6.1.7 requires that these valves be sealed closed and TS 4.6.1.7 provides the surveillance requirements that the valve positions be verified periodicall Procedure IPSP03-ZQ-0002, Revision 0, " Routine Passive Instrument Surveillances for Modes 1, 2, 3, and 4," covers the requirement of l TS 4.6. ,
l cc:
t o Containment pressure setpoint that initiates containment isolation is ;
Houston Lighting & Power Company ATTN: M. Wisenberg, Manager, Nuclear Licensing P. O. Box 1700 Houston, Texas 77001 Houston Lighting & Power Company ATTN: Gerald E. Vaughn, Vice President Nuclear Operations P. O. Box 1700    ;
i to be reduced to the minimum value compatible with normal operating l l conditions. The setpoint value and justification is to be provided
Houston, Texas 77001    p[
: by the applicant to the NRC staff in conjunction with the development of the plant TS per STP SER Section 6.2.4 page 6-13. The setpoint
      '
'
Texas Radiation Control Program Director   i }
value currently included in the proof and review TS is 5.8 psig. The NRC staff is reviewing this value for acceptabilit ,
l Q731003aa70724 i
l o Procedure IPOP02-HC-0001, " Containment HVAC " was reviewed and found I to be lacking a General Precaution on maintaining containment
o ^ DUCK 05000498 PDR
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I ressure between the TS limits. Procedure IPOP02-HC-0003, p' Supplementary Containment Purge," was reviewed and found to have incorrect values for containment pressure in the General Precaution sectio I
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The following subitems of II.E.4.2 are closed:
II.E.4.2.1-4 11.3.4. II.E.4. II.E.4. The following are to remain open pending resolution as follows:
o II.E.4.2.5B: All drawings identified in CCP 1-M-ST-0115 are issued as revised drawing which incorporate the modification change o Procedures IPOP02-HC-0001 and IP0P02-HC-0003 are revised to incorporate statement and correct values, respectively, as abov These items will be tracked collectively as Open Item 498/8708-18.
 
,
(0 pen) TMI Item I.G.1.3 Training Requirements During Low Power Testing Training will be provided during low power test programs to provide " hands on" experience for plant evaluation and off-normal events for each operating shif It is not expected that all tests will be required to be conducted by each operating shift. Observation by one shift of training of another shift may be acceptabl The results of this training will be documente This item requires training during low power testing, and the reporting of l the results. This item will remain open (498/8708-19) since it cannot be
! completed until low power testing is performe (0 pen) SALEM ATWS 2.2 Equipment Classifications and Vendor Interfac_e (Programs For AITSafe_ty-Related Components), Generic Letter (GL) 83-28 l Licensees and applicants shall submit a description of their programs for
, safety-related equipment classification and vendor interface including I criteria for identifying components, a description of the information handling system, station personnel use of the handling system, management controls, and a demonstration of design verification and qualification for procuremen o ST-HL-AE-1274, June 28, 1985, " Response to NRC Generic Letter 83-28,
! Required Actions Based on Generic Implications of Salem ATWS Events,"
2.2, commits to the requirements of GL 83-28, o The Project Q-List identifies systems, structures, and components that are safety-relate The Project Q-List, 5A479NQ1000, is consistent with FSAR Table 3.2.A-1, Balance of Plant Quality Classification of Structures, systems and Component o The Project Q-List, Table 1, Item 1.20.0, does not identify any Q-List items for the Gaseous Waste Processing System (GWPS).
 
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However, FSAR Table 15.7-1, GWPS Failure Analysis (P15.7-6),
identifies that the whole body gamma dose would exceed 0.5 Rem at both the Exclusion Zone Boundary and the Low Population Zone Boundary, and therefore, at the site boundary. FSAR Section 3.2 as
  ,
  ' well as the Project Q-List identifies that failure of systems,  i
        '
<
structures, or components that would result in greater than 0.5 Rem
  '
a whole body dose at the site boundary should be Safety Class 3 if not already Safety Class 1 or The Q-List designation of non-nuclear
  !
'
safety (NNS) for the GWPS appears incorrectly designated since Safety
,  Class 3 would make it a Q-List item. The HL&P Licensing Group stated
  '
that the GWPS was designed NNS per RG 1.26, Revision 3, C.2.d, which t exempts radioactive waste management systems from the 0.5 Rem criteria. HL&P will pursue a change to the FSAR Section 3.2, and the Q-List to correct the statement that all systems are inclusive in the
,
0.5 Rem whole body dose criteri ,  o The licensee identified in ST-HL-AE-1274 that it intends to implement
!j  the recommendations of the INPO Nuclear Utility Task Action r
Committee (NUTAC) program for the handling of vendor technical information, Vendor Equipment Technical Information Program (VETIP).
 
The NRC inspector interviewed a consultant for HL&P, Operations l  Support Engineering, on the vendor interface. The " Design Finalization Program Executive Summary" statuses the startup field validation of safety-related vendor manuals and transmittal of vendor bulletins and advisories. The consultant stated that replacement
,
vendors and equipment are sought when an existing vendor ceases l  business. Interdepartmental Procedure, Control of Vendor Documents, STPEGS IP-1.80, Revision 2, approved August 8, 1986, establishes a
'.
single program for the receipt, review, statusing, and distribution
,
of vendor supplied technical informatio Interdepartmental
   '
Procedure, Nonconformance Control, IP-4.10, approved September 2, 1986, describes resolving nonconformance t o STP Station Procedure OPGP03-ZN-0003, Revision 8, approved March 17,
,
1987, by the Plant Manager, establishes guidance for assigning a safety classification to maintenance. This procedure is also the mechanism for corrective action for safety-related equipment that has defects, deficiencies, deviations, or malfunction o The applicant did not provide any response for the following:
o Plant and corporate managements' oversight activities over safety-related structures, systems, or components, o Planned and periodic audits over activities impacting
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safety-related equipmen o Verification that vendor-recommended modifications were implemented on the Reactor Trip System (RTS) breaker \,
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o Implementation of a preventive maintenance program for components where there is inadequate traceability of component performance to the vendor (vendor can't be identified, went out of business, or won't supply required information),
i o Procedures to provide instructions for assigning a safety l classification to operating and surveillance procedure o GL 83-28, Item 2.2, will remain open (498/8708-20) pending resolution of designation of GWPS components in the Q-List and responses to the items immediately abov (0 pen) SALEM ATWS, 3.1, Post-Maintenance Testing (Reactor Trip System Components)
Applicants shall submit their review of test and maintenance procedures and TS to assure post maintenance operability testing and submit their check of vendor and engineering recommendations to ensure that any appropriate guidance is included in test and maintenance procedure Applicant shall identify any TS requirements for post-maintenance testing that degrade rather than enhance safet o STP's response to this position is in ST-HL-AE-1274 and commits to implementing the requirement o EGG-NTA-7159, February 1986, was reviewed and it was found that the applicant met the requirements of 3.1.3 -- that no TS for post-maintenance testing degrade safety for the RT o STP Station Procedure OPGP03-ZE-0020, Revision 0, approved January 7, 1987, by the plant manager, describes the post-maintenance testing program, including initiating requests, criteria and responsibilities for review and approval, criteria and responsibility for designating the activity as safety-related, and guidance for determining the testing to be performed. Section 5.3.1 of OPGP03-ZE-0020, Determination of Applicable Components, gives guidance on_ identifying components that should have post-maintenance testing. It was noted that the master document for identifying safety-related components, the Project Q-List, is not referenced in Section 5.3.1 of OPGP03-ZE-002 o STP's, " Post-Maintenance Testing Reference Manual," Revision 0, approved February 28, 1987, describes 17 types of post-maintenance tests and lists their requirements (e.g., Auto Start Test, Pump Operability Test, Valve Leakage Test, etc.).
o STP Station Procedure OPGP03-ZM-0003, Revision 8, approved March 17, 1987, by the Plant Manager, establishes criteria and responsibilities for review and approval of maintenance. Section V of the Maintenance Work Request Form requires a yes or no indication for  i post-maintenance testin ,
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GL 83-28, Item 3.1, will remain open (498/8708-21) pending completion of a change to Procedure OPGP03-ZE-0020 to include reference to the Q-List as discussed abov (0 pen) SALEM ATWS GL 83-28/4.5 Reactor Trip System Reliability (Safety Functional Testing)
On-line functional testing of the reactor trip system, including independent testing of the diverse trip features, shall be performed on all plant (On-line is Modes 1 through 6).
 
o The NRC inspector reviewed Station Procedure " SSPS Logic Train R Functional Test," 1 PSP 02-SP-0001R, Revision 0, which was approved January 25, 1987, by the plant manage o The Train S Procedure, IPSP02-SP-00015, Revision 0, was on the plant procedure computer printout list and was approve o FSAR Section 7.2.2.2.3.10.4, Testing of Reactor Trip Breakers, and Figure 7.1-2 are not consistent with IPSP02-SP-0001R. The bypass breaker designations in the FSAR, 52/BYR and 52/BYS, have been interchanged in the procedure, 1 PSP 02-SP-0001R. That is, in the FSAR, Reactor Trip Breaker 52/RTR is bypassed by Bypass Breaker 52/BYR and Breaker 52/RTS is bypassed by 52/BYS. But in IPSP02-SP-0001R Reactor Breaker 52/RTR is bypassed by 52/BYS and Reactor Breaker 52/RTS is bypassed by 52/BYR. The licensee indicated that 1PSPS02-SP-0001R is written to agree with wording in the
,
Westinghouse Maintenance Manual and not wording in the FSAR. The
'
FSAR wording, 1 PSP 02-SP-0001R, and Westinghouse Maintenance Manual should be made consistent to avoid misinterpretation by the operating staff. On page 39 of 48 of IPSP02-SP-0001R, Step 7.9.12.b should read " Train R Reactor Trip Breaker-0 pen / Tripped position" not
" Train S Reactor Trip Breaker-0 pen / Tripped position." This requires a change to Procedure 1 PSP 02-SP-0001R and appropriate review and approval per TS 6.5.1.6a before this item can be closed ou o Preoperational Test Procedure 1-RS-P-03 " Reactor Trip Switchgear" was reviewed. Section 7.6 of this procedure, performed December 12, 1986, satisfactorily demonstrated that opening a Reactor Trip Breaker with its respective Bypass Breaker closed does not cause a loss of voltage to the Rod Control Power Cabinet GL 83-28. Item 4.5, will remain open (498/8708-22) pending resolution of inconsistent wording between FSAR and IPSP02-SP-0001R and completion of IPS02-SP-001R procedure chang (Closed) SALEM ATWS Item GL 83-28/2.1 Equipment Classification and Vendor Interface (Reactor Trip System Components)
,
Licensees and applicants shall confirm that all components whose j functioning is required to trip the reactor are identified as l
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safety-related on documents, procedures, and information handling systems used in the plant to control safety-related activities, including maintenance, work orders, and parts replacement. In addition, for these components, licensees and applicants shall establish, implement, and maintain a continuing program to ensure that vendor information is complete, current, and controlled throughout the life of the plant, and appropriately referenced or incorporated in plant instructions and procedures. Vendors of these components should be contacted and an interface established. Where vendors can not be identified, have gone out of business, or will not supply the information, the licensee or applicant shall assure that sufficient attention is paid to equipment maintenance, replacement, and repair, to compensate for the lack of vendor backup and to assure RTS reliability. The vendor interface program shall include periodic communication with vendors to assure that all applicable information has been receive The program should use a system of positive feedback with vendors for mailings containing technical information. This could be accomplished by licensee acknowledgement for receipt of technical mailings. The program shall also define the interface and division of responsibilities among the licensees and tFe nuclear and non-nuclear divisions of their vendors that provide service on RTS components to assure that requisite control of and applicable instructions for maintenance work are provide o The NRC inspectors examined Letter ST-HL-AE-1274, which states that HL&P has reviewed all components whose functioning is required to trip the reactor and these components have been properly classified in the design document o A technical evaluation report (Supplemental SER (SSER) 1, Appendix N)
states that applicable RTS components were verified to be properly classifie SSER 1 states that the program meets the requirements of GL 83-28/2.1 (Part 1) and is acceptabl o A technical evaluation report (SSER 2, Appendix T) states that STP is a participant in the Westinghouse interface program for the RT SSER 2 states that the program meets the requirements of GL 83-28/ (Part 2) and is acceptabl Based upon the above information, this item is considered close (Closed) Salem ATWS Item GL 83-28/1.2 Post-Trip Review - Data and Information Capability Licensees and applicants, shall have or have planned a capability to record, recall, and display data and information to permit diagnosing the causes of unscheduled reactor shutdowns prior to restart and for ascertaining the proper functioning of safety-related equipmen _ _ _ _ _ _ _ _ _
 
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o The NRC inspectors reviewed Letter ST-HL-AE-1274 dated June 28, 198 This letter contains a description of the capabilities of the Proteus Computer System and The Emergency Response Facilities Data Acquisition and Display System (ERFDADS). STP SSER 1 includes a conclusion that the applicant's post-trip _  1 review data and information capabilities are acceptabl ,
o The CRTs and typers associated with Proteus Computer System and ERFDADS were inspected in the Unit 1 control room. The status of the systems was discussed with a control room operator. Although testing is continuing on the systems, the CRTs and typers are operating and it appears the systems will function as described in Letter ST-HL-AE-127 Documentation and inspection indicate the facility has planned the required capability for post-trip data and information. This item is considered close { Closed) Salem ATWS GL-83-28 Item 4.3 Reactor Trip System Reliability (Automatic Actuation of Shunt Trip Attachment for Westinghouse and B&W Plants)
Westinghouse and B&W reactors shall be modified by providing automatic RTS actuation of the breaker shunt trip attachments. The shunt trip attachments shall be considered safety-related (Class 1E).
 
o Procedure IPSP02-SP-00012, Revision 0, was reviewe STP SER Item 5 of Page 15-22, requires that quality assurance criteria set forth in Appendix B to 10 CFR 50 be met. During a meeting with the QC General Supervisor and the Audit / Surveillance Supervisor the QA program was described concerning operational testing of equipment. The program meets the requirements as indicated in GL 83-28.
 
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o Procedure IPSP03-RS-0002, Revision 0, " Manual Reactor Trip TAD 0T,"
was completed. It was determined that this procedure meets the operability test requirements to verify contacts and wiring of the manual trip circuit before startup after each refueling outag Item 4.3 is considered closed.
 
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  (0 pen) Salem ATWS Item GL 83-28/4.1 Reactor Trip System Reliability (Vendor-Related Modifications)
All vendor-recommended reactor trip breaker modifications shall be reviewed to verify that either: (1) each modification has, in fact, been implemented, or (2) a written evaluation of the technical reasons for not
,
implementing a modification exist <
!
o STP has Westinghouse type DS-416 reactor trip breakers installe STP SSER 1 (NUREG-0781 Supplement No. 1) states in Section 15.8.2
:
:
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under Action Item 4.1 that the applicant has committed to implement all vendor-related modifications before. fuel loading, and that the applicant's position on Item 4.1 is acceptabl o Letter ST-HL-AE-1274 was reviewed. It indicated on Page 14 of 18 in i
Section 4.1 that "HL&P has been informed by Westinghouse that a
~
design discrepancy had been identified in the undervoltage attachment
-
and that Westinghouse intended to replace the undervoltage attachments on DS-416 reactor trip switchgear. Field change notices (FCNs) have been issued by Westinghouse for installation and adjustment of the replacements."
o STP FCN TGXM-10563, Shop Order No. 386, indicates that the undervoltage trip assemblies in the raattor trip switchgear have been replaced. This FCN was closed on January 28, 198 o Westinghouse Nuclear Service Division Technical Bulletin 83-03 was reviewed and found to provide the Westinghouse recommendation for a single method of independently verifying the function of the shunt
^
trip and the undervoltage trip mechanisms of the Reactor Trip Breaker It was not intended that a utility would follow this general procedure verbatim, but would first parform a thorough review of the general procedure against the plant specific system.
. o STP Procedure ISP02-SP-0001R, Revision 0, " Solid State Protection System (SSPS) Logic Train R Functional Test," date January 29, 1987, was reviewed and found to include appropriate procedure steps fo" testing the Undervoltage Trip Attachment and the Auto Shunt Trip Function independently of one anothe o Westinghouse Nuclear Service Division Technical Bulletin 84-02 was reviewed and found to advise all affected plants of a condition which may exist on DS or DSL Class 1E circuit breakers used for Reactor Trip or Safeguards Power Breakers. This condition, which should be
  . investigated, involves rotential wire damage on the left side,
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particularly ,in the vicinity of the wire retainer which forms the extreme lef t coundary of the breaker, f acing the breaker fron Instruction for dealing with this conditicn (replacement of damaged wires and providing additional rigidity and mechanical support) are i'  also included in this fechnical Bulleti STP Procedure OPMP05-NA-0008, Revision 1, Westinghouse 480 Volt Breaker Test, dated February 24,
,-  1987, was reviewed and found to include a procedural step to insoect breaker wiring, witn a precautionary note dealing with the left side, facing the front of-the breaker. Westinghouse Nuclear Service Division Technical Bulletin No. NSD-TB-84-02, Revision 1, "DS/DSL Breakers - Potential Wire Damage For Reactor Trip or Safeguards Power Breaker," is referenced in this procedure. An attempt was made to ascertain that inspection of Reactor Trip Breakers had actually been performed and the inspection results documented, but this information was not made availabl . _, __, ,_. -
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Salem ATWS Item GL 83-28/4.1 will remain open (498/8708-23) pending STP submittal of verifiable evidence that inspection of the Reactor Trip Breakers for potential wire damage has been completed and any deficiencies noted have been correcte . Licensee Action on Previous Inspection Findings (Closed) 498/8620-01 - This item concerned the use of ambiguous terms such as " extent necessary" for criteria in certain preoperational procedures without offering guidance as to what constitutes " extent necessary." The specific procedures have been changed to delete such phases. The procedure changes were reviewed by the NRC inspectors and deemed sati sfactory. This item is considered close (Closed) 498/8620-02 - This item concerned using acceptance criteria in
,  procedures without guidance as to what is acceptable. The specific case involved fuel handling trolleys and hoists which were required to be operated without excessive vibration. The criteria in this instance was found to be superfluous as such abnormalities are routinely watched for and covered by SAI 18, paragraph 5.3.4. The procedure was changed and deemed satisfactory by the NRC inspector . Site Tours The NRC inspectors conducted site tours independently. These tours were made to assess the protection on inplace safety-related equipment, plant status, and to observe testing. The areas toured included:  Unit 1 -
Mechanical and Electrical Auxiliary Building (MEAB), Reactor Containment Building (RCB), Fuel Handling Building (FHB), and the Emergency Diesel Generator Building. With the exception of the violation noted in paragraph 11, the NRC inspectors noted a marked improvement in the maintenance of areas and equipment turned over to the Nuclear Plant Operations Department (NPOD).
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!  No additional violations or deviations were identifie . Review of The Manual Trip Circuit
,
Because of a drawing error relative to the electrical location of the manual trip circuit in relation to the output transistors Q3 and Q4 in a
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l  similar Westinghouse designed plant with a SSPS, the NRC inspector inspected
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facility Drawings 14926-0387(2)00171-BWN and 14926-0387(2)00172-BWN.
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These drawings correctly depict the manual trip circuit downstream of the output transistors Q3 and Q .
16. Procedures Review The NRC procedure review team by review of selected procedures, personnel interviews, and system walkdowns, assessed the applicant's procedures for adequate administrative controls, technical accuracy, and compliance with
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, , ,--r--. , - . , , , , , , , -,- - - . , --,n- , - , - - - . ,,, , , , - , --n- - ,--, . .  . ___ _  __ _ _ _ _ _ _ _ _ _ _ _ _ _ __
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regulatory requirements. A selected number of the procedures reviewed were field verified by walking down the applicable system and verifying that the procedure reflected as-built condition Procedures in~ the following areas were selected for review:
i
' Plant Operating Maintenance Emergency Operating Off-Normal Operating and Alarm Response Surveillance Following is a discussion of the NRC inspection effort in each of'these
; areas including a list of the specific procedures reviewed and NRC i inspector observations and findings. (The above letters identify the discussion for the corresponding procedure category below.):
'~ Plant Procedures (1) Review Of Program For 10 CFR Part 50.59 Safety Evaluations
:  The NRC inspector reviewed administrative procedures to ascertain whether responsibilities have been assigned to i  appropriate personnel to ensure that plant procedures will be reviewed, updated, and approved as required and to ensure that
;  the revision process includes provisions for 10 CFR 50.59 consideration Procedures reviewed were as follows:
o OPGP03-ZA-0002, Revision 8, " Plant Procedures"
,
l  o OPGP03-ZA-0003, Revision 4, " License Compliance Review" I  .o OPGP03-ZA-0010, Revision 2, " Plant Procedure Compliance, Implementation and Review" o OPGP03-ZA-0033, Revision 0, "10 CFR 50.59 Safety
;    Evaluations"
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These procedures define the procedure revision process for all plant procedures with the exception of plant organization, plant
;  policies, and emergency operating procedures. They define the
'
responsibilities of personnel to ensure that each procedure is reviewed within a 24-month interval and is revised, if required, 4  and that revisions are approved by the plant manage NRC inspector observations / concerns are discussed below:
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(a) OPGP02-ZA-0010, Revision 2, " Plant Procedure Compliance, Implementation and Review." This procedure requires that plant procedures be reviewed within 24-month interval This review of plant procedures determines whether a
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revision needs to be generated. However, once a procedure revision has been determined to be necessary, there are no guidelines to ensure that the procedure will be revised in a timely manne These guidelines should be written to ensure that a procedure that is determined, by the review, to need revising is not used while it is being revise This is an open item (498/8708-24).
F (b) OPGP03-ZA-0033, Revision 0, "10 CFR 50.59 Safety Evaluations." This procedure is included in the procedure review process to ensure that procedure reviews include 10 CFR 50.59 considerations. The inspector noted that the technical support superintendent is responsible for the review of 10 CFR 50.59 Safety Evaluations. However, there is no technical support superintendent within the operations department. Pending correction, this is an onen item (498/8708-25).
(c) Procedures OPMP08-ZI-0065, Revision 0, " Field Testing .of Power Supplies and Over Voltage Protectors," and OPMP07-SP-0001, " Revision 0, "SSPS Decoder Printed Circuit Board Test and Rework." The NRC inspector determined that these procedures had been reviewed on July 29, 1986, but they had not been revised at the time of the NRC inspection 8 months later, even though the applicable procedure biennial review form, OPGP03-ZA-0012-2, stated that these procedures required revision to correct inadequacies. This L  is contrary to Step 3.3.2.3 of Procedure OPGP03-ZA-0002, Revision 3, " Plant Procedure Compliance, Implementation, and Review" which states, "The cognizant DM shall ensure that identified procedural inadequacies are corrected in accordance with OPGP03-ZA-0002 (Plant Procedures)." This failure will be tracked as an open item pending resolution by the licensee (498/8708-26).
(2) Review Of Standing Order and Short Term Instruction Programs The NRC inspector verified that administrative controls were established for STP standing orders and short term instructions (STIs).
The administrative controls for standing orders were defined in PRO-1, Revision 2, " Standing Orders," which provided a mechanism for their issue and distribution. The standing orders were required to be reviewed and updated quarterly; however, there was no mechanism to document the review. Pending the applicants establishing a method to document the quarterly review of
_ _ _ . _ _ _ _ . - _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . . _. ,_ _ _ _    .  .- __
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standing orders, this is an open item (498/8708-27).
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Responsibility was assigned to the Unit 1 operations supervisor to issue, distribute, review, and update standing orders. PRO-1 established a limitation on the type of instructions that could be issued as a standing orde The administrative controls for STIs were defined in PRO-25, Revision 0, "Short Term Instructions," dated March 31, 198 The standing order provided a mechanism for the issue, distribution, review, and updating of STIs. The shift
  ,
  ,
supervisor was assigned responsibility for control of STIs including reviewing and maintaining them. The standing order provided limitations on the type of activities that could be accomplished by STI Issuance and cancellation of STIs may be accomplished by signature of the shift supervisor, operations supervisor, or the reactor operations manager. The NRC inspector reviewed selected standing orders and STIs to verify they met the guidance specified and did not perform activities that should be covered by procedures. The applicant's Standing Order PRO-1 states that standing orders shall not be used in
,
lieu of procedures. The below listed standing orders were reviewed:
''  Number Revision  Subject
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PRO-1 2  Standing Orders PRO-2 1  Required Reading Program PRO-18 0  Event Reports PRO-23 0  Work Control Guidelines PRO-25 0  Short Term Instructions From discussions with a shift supervisor and after comparison of the guidance contained in Procedures OPGP03-ZA-0002, Revision 8,
  " Plant Procedures" and OPGP03-ZA-0003, Revision 4, " License Compliance Review," the NRC inspector determined that the shift supervisor was knowledgeable of the steps necessary to make temporary changes to procedures.
!
  (3) Control Room Logbooks
,  The NRC inspector verified that OPGP03-ZQ-0001, Revision 0, i  " Maintenance of Reactor Operations Logbooks," provided guidance for preparation and review of logbooks. The procedure described the usage, control, and type of logbooks maintained in the control room. The requirements for retention in the quality assurance vault and maintenance and storage in the control room were specifie NRC inspector observations made during the review of OPGP03-ZQ-0001 are discussed below:
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o There was a typographical error (typo) in Step 6.5.6 in that the statement "only to reference the procedure" was not needed. The step should match Step 6. o The steps in Section 6.5 were apparently misnumbered since Step 6.5.5 was missin Plant Procedure OPGP03-ZA-0063, Revision 0, " Reactor Operations Shift Turnover," described how the shift relief and turnover was
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to be conducted and what documentation was to be generated and
In Reply Refer To:  dRil 2 4 Md7 Dockets: 50-498/87-08 50-499/87-08 Houston Lighting & Power Company ATTN: J. H. Goldberg, Group Vice President, Nuclear P. 0. Box 1700 Houston, Texas 77001 Gentlemen:
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Thank you for your letter of June 25, 1987, in response to our letter and Notice of Violation (498/8708-01) dated May 29, 1987. We have reviewed your reply and find it responsive to the concerns raised in our Notice of Violatio We will review the implementation of your corrective actions during a future inspection to determine that full compliance has been achieved and will be maintaine
retained. This procedure detailed the requirement for a Shift Relief and Turnover Log. Observation on this procedure is:
Typo in Step 4.1.3.6.3, " Shift turnover of prior to end of shift," should not include the first "of!"
b. Operating Procedures This area of inspection was conducted in order to confirm that plant operating procedures are prepared and approved to control safety-related operational activities. An index of all current procedures dated March 10, 1987, was reviewed for plant operating procedures identified in Regulatory Guide 1.33, Revision 2, in the following categories:
. o Administrative Procedures o General Plant Operating Procedures
,  o Procedures for Startup, Operation, and Shutdown of Safety-Related Systems Also, a sample: review of plant operating procedures in the above categories was conducted to verify that the appropriate format was used and that each procedure was technically adequate-to accomplish the stated purpos The results of the reviews in each category are documented below:
  (1) Administrative Procedures
;  (a) The following procedures were not issued as of March'13, 1987:
o OPGP03-ZA-0064, "NPOD Preshift Briefing" o OPGP03-ZQ-0001, " Maintenance of Reactor Operations
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Logbooks" o OPGP03-CN-0001, " Radio Communication" l
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  -_______________   _
Sincerely, Cricinal Signal aj
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    'd+P..Jaudon J. E. Gagliardo, Chief Reactor Projects Branch cc:
Houston Lighting & Power Company ATTN: M. Wisenberg, Manager, Nuclear Licensing P. O. Box 1700 Houston, Texas 77001 Houston Lighting & Power Company ATTN: Gerald E. Vaughn, Vice President Nuclear Operations P. O. Box 1700 Houston, Texas 77001 Texas Radiation Control Program Director f,/
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bcc: (seenextpage)
  , , A PI 7%f!3 C: PPB /C e Q:RF B \
HFBundy:cs ' GLConstable A/JEGkglidrdo 7/zt/87 7/n-/87 7g4f87
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o 0PGP03-CN-0002, " Telephone Communication" o OPGP03-CN-0003, " Plant Public Address and Alarm System" o OPGP03-CN-0004, " Pocket Pager System" o OPGP03-CN-0005, " Communications Console System" o OPGP03-CN-0006, " Communications System Test Program" Pending issuance of the above procedures, this is an open item (498/8708-28).
'
  (b) Procedures Reviewed Procedure N Revision Title OPGP03-ZA-0002 8 " Plant Procedures" 0PGP03-ZA-0010 2 " Plant Procedure Compliance.
,
Implementation and Review" l
OPGP03-ZO-0001 3 " Equipment Clearance" i
OPGP03-ZO-0003 3 " Temporary Modifications and Alterations OPGP03-ZA-0055 0 " Plant Surveillance Scheduling" 0PGP03-ZE-0004 3 " Plant Surveillance Program" OPGP03-Z0-0004 0 "Flant Conduct of Operations" 0PGP03-ZA-0033 0 "10 CFR 50.59 Safety Evaluations" (c) NRC Inspector Observations / Concerns 1) OPGP03-ZA-0002, Revision 8, " Plant Procedures" o There is no limit on the number of " field change requests" (FCRs) that can be issued / approved for a procedure before the procedure must be revise o There is no requirement to incorporate permanent FCRs into procedure' revision .- ,
o The procedure does not describe a program for controlling the expiration dates on FCRs (i.e.,
is the FCR deleted or must it be evaluated for procedure incorporation).
o There are no administrative controls for the use of FCRs (i.e., mark up controlled procedure to reference FCRs, page for page replacement, sign off on FCR or controlled procedure, etc.).
o The approved FCRs are not listed in the master procedure listing /inde o There is no required time limit for FCRs to receive final approval. The administrative limit of 14 days (should procedure) is not being followed as evidenced by FCN 87-21 (initial approval 1/21/87), FCN 87-055 (initial approval 2/13/87), and FCN 87-064 (initial approval 2/19/87) not receiving final approval as of March 13, 198 Pending resolution of the above concerns related to the control of temporary procedure changes, this is an open item (498/8708-29).
2) OPGP03-Z0-0003, Revision 3, " Temporary Modifications and Alterations"
        ,
o The procedure makes statements in Steps 1.2 and 1.3 which appear to allow other programs to be used for temporary modifications. This appears to allow circumventing of the required controls for temporary modification o The procedure steps cannot be followed in sequence as required by administrative procedur o There is no requirement for the shift supervisor to acknowledge that the temporary modification has been installe o The procedure allows the use of one tag on the
:  outside of a panel door for several temporary jumpers or lifted leads.
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;  o The inclusion of the startup temporary alteration
!  program in the plant operations program procedure causes unnecessary complications of this procedure. A simple reference to the startup procedure and requirement for shift supervisor l
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I approval of temporary alterations and for all temporary alterations to be converted to temporary modifications would be sufficient. All required temporary alterations should be converted to temporary modifications prior to operating license issuance. See paragraph ~16.b.(2)(b)2) for further concerns regarding modifications and alteration o The procedure allows the use of plastic screws
,
and washers for lifted leads with no inventory requirements to assure that they are remove Pending resolution of the above concerns, this is an open item (498/8708-30).
3) OPGP03-Z0-0004, Revision 0, " Plant Conduct of Operations" o Step 4.2.3 of this procedure indicated that only 4  an RO is required in the control room during operating Modes 1 through 4 while TS Section 6.2.2.b requires an RO in the control room during all modes (with fuel in reactor) and an R0 and SRO in the control room during Modes 1 through o Step 4.10.3 states that entry into limiting conditions of operations (LCOs) as required by TS be controlled in accordance with approved procedures; however, there is no procedure for tracking and controlling of LCOs by plant operations.
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Pending resolution of the above concerns, this is an open item (498/8708-31).
4) A general concern identified during the review of administrative procedures was the apparent overuse of the word "should" which denotes a recommendation rather than a strict requirement. The NRC inspector is concerned that this will lead to procedure abuse
,
and a loss of contro This was discussed with licensee management and discussed again during the NRC exit meeting. The plant manager stated that
!
individual managers are expected to enforce the
"should" statements in procedures and that the licensee QA organization would issue QA findings in areas where lack of control is evidenced. The NRC inspector cited the specific example of the final
, approval of FCNs to procedures exceeding the 14-day
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i time limit as a lack of control. Also, the following examples of inappropriate "shoulds" in procedures were noted by the NRC inspector:
t o Step 4.10.3.1 of OPGP03-ZO-0004 states, "The Shift Supervisor should be informed of TS surveillance tests that have failed to meet their
,    acceptance criteria . . . ."      ,
o Step 4.10.1.2 states, "The operator who completed the checklist should sign and date the checklist and present the checklist to the supervisor who directed the checklist be complete For components that are safety-related, supervisory
;    personnel should direct that an independent verification be performed. When the independent verification is completed, the operator who performed the independent verification should sign and date the checklist."
,
o Step 4.10.1.3 states, "The completed checklist should be reviewed by the shift supervisor, unit supervisor, or chemical operations foreman, as appropriate, to verify that the directed actions were completed and to note any exceptions or unusual conditions. The completed checklist should be inserted in the control room system      *
4    status file or the watchstation system status
'
file, as appropriate. The superseded checklist
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should be forwarded for record retention or i    discarded as appropriate."
i Pending resolution of this concern related to the use
~
of "shall" and "should", this is an open item
,
  (498/8708-32).
I  No additional concerns or comments were identified during the review of administrative procedure (2) General Plant Operating Procedures (a) Procedures Reviewed i
Procedure N Revision  Title
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OPGP03Z0-0022 0 " Post Trip Review" 1 POP 03-ZG-0001 2 " Plant Heatup" 1 POP 03-ZG-0004 1 " Reactor Startup"
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, , . , - - - - ~ - ,- - - - r--r.--w, w # w , w way -, - - - -,_-..--.mi,, - , ,
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4 j    IP0P03-ZG-0005 0  " Power Operations" IP0P03-ZG-0006 0  " Plant Shutdown To
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Hot-Standby" l  (b) NRC Inspector Observations / Concerns 1) Generally, there were several inconsistencies noted with signoffs of steps and sections (i.e., _some action
;    steps not initialed, some prerequisites not signed
_
off, and the location of signoff blocks varied which i
creates confusion as to what is being signed off).
;    Pending resolution, this is an open item (498/8708-33).
2) During the review of Procedure 1 POP 02-EW-0001, Revision 2, " Essential Cooling Water Operation," it was noted that FC 87-092 to the procedure had been
. issued because of " leads landed on plastic during S/U." The NRC inspector asked for the temporary
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modification documentation to support these lifted leads. Temporary Modification Nos. T1-EW-8710, -8711, and -8712 were provided to the NRC inspector. A review of these modifications revealed that the j    emergency load sequencing capability of the essential cooling water system components had been disabled by the installed modifications. The " License Compliance Review Forms" completed for the modifications incorrectly stated these modifications did not involve j
'
a change to the facility as described in_the FSAR and do not require a change to the TS when, in fact, the installed modifications do change the facility as
!    described in the FSAR and would require a change to
*
the TS. These installed modifications would also
;
require a 10 CFR 50.59 Safety Evaluation to be performed which was not accomplished. This failure to
;    identify a facility modification which involved an
,    unreviewed safety question was discussed with licensee
,
personnel and they stated that all temporary i    modifications existing at fuel load would have been reevaluated per 10 CFR Part 50.59 for unreviewed
;    safety question determinations. This will be
;    evaluated by the NRC during closure of the Open Item 498/8708-30 on conversion of temporary
,
alterations to temporary modification Pending the 4    licensee's response which should address the steps i    taken to assure that all previous License Compliance j    Reviews have been performed properly or that controls
;    are in place to assure that needed
!    reevaluations / reviews are performed, this is an open item (498/8708-34).
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Houston Lighting & Power Company -2-bcc to DM8 (IE01)
 
bec distrib. by RIV:
3) IPOP03-ZG-0001, Revision 2, " Plant Heatup" The prestartup checklist does not include related activities such as emergency shutdown system readiness or the performance of Procedure 1 POP 02-SI-0001,
RPB   DRSP RRI-0PS R. D. Martin, RA RRI-CONS Section Chiaf (RPB/C)
  " Safety Injection Accumulators," for filling and venting the safety injection accumulators. This is an open item (498/8708-35).
RPSB   MIS System j RIV File D. Weiss, RM/ALF RSTS Operator  R. Pirfo, OGC i R. G. Taylor, RPB/C RSB   ]
 
Project Inspector, RP8 R. Hall  1 P. Kadambi, NRR Project Manager   i l
4) IPOP03-ZG-0004, Revision 1, " Reactor Startup" There is no requirement to verify that the neutron count rate on the source range instruments is above a set minimum. This is an open item (498/8708-36).
 
5) 1 POP 03-ZG-0008, Revision 0, " Power Operations" The purpose and scope section is incomplete. There.is nothing listed after . . . "following:". This is an open item (498/8708-37).
 
No other concerns were identified during the review of general plant operating procedure (3) Procedures For Startup, Operation, and Shutdown Of Safety-Related Systems (a) The following procedures were not issued as of March 13, 1987:
o IP0P02-CG-0001, " Electric Hydrogen Recombiners" o 1 POP 02-CZ-0001, " Electric Hydrogen Recombiners" o IPOP02-DB-0005, " Technical Support Center Diesel Generator" o 1 POP 02-II-0001, "Moveaule Incore Detector System Operation" o 1 POP 02-SB-0001, " Steam Generator Blowdown System" Pending issuance of these procedures, this is an open item (498/8708-38).
 
(b) Procedures Reviewed Procedure N Revision Title 1 POP 02-SI-0002 2 " Safety Injection System Normal Lineup"
 
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      .1 POP 02-AF-0001 1 " Auxiliary Feedwater" 1 POP 02-CS-0001 1 " Containment Spray Standby Line-up"
-      IPOP02-CV-0001  1 " Makeup To the Reactor Coolant System" 1 POP 02-DG-0001 1 " Emergency Diesel -
'        Generator No. 11" 1 POP 02-EW-0001 2 " Essential Cooling Water j
Operation" i
IPOP02-RH-0001  1 " Residual Heat Removal System Operation" 1 POP 02-SI-0001 1 " Safety Injection Accumulators"
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    . (c) Generic NRC Inspector Observations / Concerns l      o There are no procedural instructions provided for j      filling and venting certain systems (i.e., auxiliary
,      feedwater and safety injection systems). This is an l      open item (498/8708-39).
 
l-      o There are no procedural requirements provided for disposition of exceptions identified during system
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lineups. There is a page for listing exceptions but no requirement to evaluate each identified exception.
 
.
This is an open item (498/8708-40).
 
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!      o There were inconsistencies between the procedures l-      valve lineups for identifying that pipe caps were l      installed on vents, drains, and test connections. At
!      least one procedure identified pipe caps while others did not. This is an open item (498/8708-41).
 
l      0 It is recommended that operating procedures for TS    1 i
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systems reference the applicable TS sections. It was noted that some procedures did reference TS sections I,
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and others did not. This is an open item (498/8708-42).
 
!      o There were inconsistencies with required signoffs on
!      different procedures (i.e., IPOP02-CS-0001 did not require signoffs for control board lineups while other procedures did). This is an open item (498/8708-43).
 
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,, e. , - ~ - ---- --w . , _ . .n, ,~,,,,. - _ . - , , - . . , . . . ,,,_._y. ,m .-,,.mn -,  m ,w _ ww,,. ,-,._.,,-.,..,,---,,--,n, -,-
 
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, Maintenance Procedures The NRC inspectors reviewed selected applicant maintenance and administrative procedures to verify that the procedures were in the appropriate format, that they were technically adequate, and that maintenance activities would be controlled in accordance with regulatory requirements. The procedures were divided into the areas of maintenance and measuring and test equipment (M&TE) for this review. A number of Unit 1/ common maintenance procedures (as indicated in the table below) had not been issued at the time of the i  NRC procedure revie Prior to licensee issuance, the NRC will review the number and types of procedures not issued to determine if this would impact plant operations. The numbers in the table below were taken from a computer maintenance procedure listing dated April 7, 1987. Pending a subsequent NRC review, this is an open item (498/8708-44).
 
Total N Total No. of In Review Total N Procedure Volume    Procedures Process Apprcved
"
PMP01-Maintenance Adminis-trative Procedures    5 1  3
!  PMP02-Maintenance General Procedures    8 0  7 PMP05-Electrical Maintenance i  Procedures    438 21  384 PMP06-Metrology Lab Calibration Procedures    321 46  236 i
PMP07-I&C Maintenance Procedures 155    1  51
;  PMP08-I&C Calibration Procedures 613    27  328
 
PSP 02-I&C Functional Test
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Surveillance Procedures    129 17  119 PSPO4-Mechanical Surveillance Procedures    3 1  1
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PSP 05-I&C Calibration Surveil-lance Procedures    160 8  142
!  PSP 06-Electrical Surveillance
: Procedures    22 1  20
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PSP 13-Response Time Surveil-
;  lance Procedures    20 0  0
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(1) Maintenance Procedures (a) Procedures Reviewed Procedure N Revision . Title OPGP03-Z0-0007 2 " Conduct of Maintenance" OPMP01-ZA-0004 4 " Maintenance Procedures" OPGP03-ZM-0003 7 " Maintenance Work Request Program" OPGP03-ZA-0010 2 " Plant Procedure Compliance, Implementation, and Review" 1 PSP 06-DJ-0001 0 "125 Volt Class IE Battery 7-Day Surveillance Test" l
1 PSP 06-DJ-0002 0 "125 Volt Class 1E Battery Quarterly Surveillance-Test" OPGP03-ZO-0004 0 " Plant Conduct of Operations" OPMP04-AF-0001 3 " Auxiliary Feedwater Pump Maintenance" OPMP04-CC-0001 3 " Component Cooling Water Pump Maintenance" 0PMP04-DG-0004 0 " Standby Diesel Generator Starting Air Compressor Maintenance" OPMP04-DG-0005 0 " Standby Diesel Generator Maintenance" OPMP04-FC-0001 1 " Spent Fuel Pit Cooling Pump Maintenance" OPMP04-FW-0003 1 "Atwood-Morrill Air Assist 20-Inch Feedwater Check Valve Maintenance" OPMP04-J F-0001 0 " Fuel Handling Machine Inspection" OPMP04-MS-0001 1 " Main Steam Safety Valve Removal and Installation" l
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i
 
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l OPMP04-MS-0002 2 " Main Steam Dump Valve Actuator Maintenance" 0PMP04-MS-0003 1 " Main Steam Dump Valve Actuator Removal and Reinstallation" 1PMP05-DJ-006 0 " Battery Charger Maintenance-Class 1E 125 VDC Distribution Panels" 1PMP05-VA-002 0 " Inverter /Reclifier Mainten-ance Westinghouse 7.5 KVA" 1PMP05-PM-1101 0 "Switchgear Maintenance-MCCEIAl" 1PMP05-PK-1014 0 "Switchgear Maintenance-Bus E1A Cubicle 14" 1PMP05-PK-2014 0 "Switchgear Maintenance-Bus EIB Cubicle 14" 1PMP05-PK-3014 0 "Switchgear Maintenance-Bus E1C Cubicle 14" OPMP04-SN-0002 1 " Hydraulic Snubber Removal and Installation, Fluid Addition and Sampling" 0PMP04-SN-0006 0 " Anchor Darling Model 151 Mechanical Snubber Maintenance" OPMP04-RC-0003 1 " Reactor Coolant Pump Maintenance" OPMP04-RC-0007 0 " Pressurizer Spray Valve Maintenance" OPMP04-RX-0001 1 " Reactor Vessel Head Removal For Non-Rapid Refueling" OPMP04-SI-0002 1 "High Head Safety Injection Pump Maintenance" OPMPO4-ZG-0006 3 "Limitorque Operator Removal and Installation" OPMPO4-ZG-0017 0 " Pacific Gate Valve Mainten-ance (Bolted Bonnet)"
 
  .
. .
 
(b) NRC Inspector Observations / Concerns 1) OPGP03-ZM-0003, Revision 7, " Maintenance Work Request Program" o Step 2.5 rather than stating "see addenda" could more appropriately specify the applicable addenda for this ste o Step 3.1.1 is worded in such a way that a number of people other than the shift supervisor (control room) would have the authority to release installed systems for maintenance. The work start approval authority needs to be more clearly defined so that the shift supervisor is clearly the control point for releasing installed components / equipment. Pending clarification of the approval authority for release of systems that could affect the plant, this is an open item (498/8708-45).
 
o Step 4.1.7 should include words to the effect that deletions should be lined once through, initialed, and date ) OPGP03-ZO-0004, Revision 0, " Plant Conduct of Operations" Step 4.10.6 states, "The shift supervisor or Chemical Operations Foreman, as applicable shall authorize work start approval for all maintenance, test, and other activities that may affect the operation of the plant or the status of plant structures, systems, and components." The "as applicable" should be more definitive to insure the shift supervisor controls installed components / equipment. Pending clarification of the "as applicable," this is an open
'
item (498/8708-46).
 
3) OPMP04-AF-0001, Revision 3, " Auxiliary Feedwater Pump Maintenance"
  " Documentation" section required by Addendum 3 of Procedure OPMP01-ZA-0004 was missin ) OPMP04-CC-0001, Revision 3, " Component Cooling Water Pump Maintenance" o There was a possibility of placing wrong data on data sheet due to dats Steps 5.12.5 and 5.1 being out of sequenc . - --
    . - ,
 
. . _ . . .. ~   _ . _ _ _ - .-. . _ _ - ..  . _ ._
. .
 
'
              !
,
o One fastener torque value was different from that recommended in vendor manual. The procedure required torquing of the motor mounting bolts to 324 ft-lbs while the vendor manual required torquing to 70 ft-lbs.
 
.i      Pending correction of the above items, this .is an open
;      item (498/8708-47).
 
5) OPMP04-FC-0001, Revision 1, " Spent Fuel Pit Cooling Pump Maintenance" o " Documentation" section required by Addendum 3 of Procedure OPMP01-ZA-0004 was missing.
 
4      o Part numbers in parenthesis after the part name
'
refer to items on the figure in Addendum 1; I      however, there was no reference to the addendum at the start of the procedure (Step 5.7).
 
l      o Steps 5.13.24 and 5.13.26 are incorrectly i      identified as 5.13.25 and 5.13.27 on the data
,
shee ,
o Step 5.12.9 should note the total indicated    ,
runout to be less than 0.002-inch.
 
!      Pending resolution of the above items, this is an open    !
!      item (498/8708-48).
 
!    6) OPMPO4-FW-0003, Revision 1, "Atwood-Morrill Air Assist 20-Inch Feedwater Check Valve Maintenance"
      " Documentation" section as required by Addendum 3 of Procedure OPMP01-ZA-0004 was missin ) OPMP04-JF-0001, Revision 0, " Fuel Handling Machine
:
Inspections" o Items located on the data sheet did not have an
,.
asterisk and corresponding note nor any notation
'
within the main body to identify information required to be recorded, o Step 3.1 was apparently misnumbered, since on the data sheet it was a chart for making not n      applicable undesired sections while in main body of the report, it was a prerequisite for obtaining a cleaning solvent.
 
.
i w.-.m,.,----.-m--,-,%-+,.-,,.e,cy----,&  --,,---.---,--------ewy -m y-----w-- , - - , -v-yy, --, , - - -r,,,,+w,-
.--n - ,, -e , - - - . tm-
 
,
. .
 
o Step 6.1 is not numbered on the data sheet as it should be to be consistent with other procedure Pending resolution of the above items, this is an open item (498/8708-19).
 
8) IPMP05-DJ-006, ~ Revision 0, " Battery Charger Maintenance-Class IE 125 V DC Distribution Panels" Step 4.4 should specify minimum accuracy requirements for M&T ) IPMP05-JA-0002, Revision 0, " Inverter / Rectifier Maintenance Westinghouse 7.5 KVA" o Data Sheet -1, Step 6.9 should read "3CB Input DC Breaker in lieu of 3CB AC Input DC Breaker."
 
o Step 4.3.1 M&TE requirements should include a second torque wrench to be used when low torque values are require o Step 6.17 should include instructions to include data sheets from referenced procedures as part of the required documentatio Pending resolution of the above items, this is an open item (498/8708-50).
 
10) IPMP05-PM-1101, Revision 0, "Switchgear Maintenance -
MCCEIAl" o Step 4.3.2 should specify accuracy requirements for M&T o Torque wrench should be of 15-80 foot pound range to correspond to Addendum o Quality Control should be notified prior to performing any torquing on bolted connections.
 
; Pending resolution of the above items, this is an open
,
item (498/8708-51).
 
i 11) IPMP05-PK-2014, Revision 0, "Switchgear Maintenance -
Bus ElB Cubicle 14."
 
i- .
i Step 4.5.3 should include opening of breakers for containment spray and component cooling water pumps.
 
i
!
i i
 
- _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ ___ . _ _ _    .
a  .
 
12) IPMP05-PK-3014, Revision 0, "Switchgear Maintenance -
Bus E1C Cubicle 14."
 
o  Step 4.5.3 should place 4.16 KV Bus E1C Cubicle I supply breaker in "Open" position and overcurrent lockout relay in " Reset" positio Procedure states them in revers o  Overcurrent lockout relay should be identified as
      "86B" and not as " Breaker" under the " Comp. Name" column in Step 4. o  "86A" relay should be identified as " Generator Differential Lockout Relay" in Step 4. This same comment applies to Procedures 1PMP05-PK-1014 and 2PMP05-PK-201 Pending resolution of the above items, this is an open item (498/8708-52).
 
,    13) OPMP04-DG-0004, Revision 0, " Standby Diesel Generator Starting Air Compressor Maintenance" o  " Documentation" section as required by Addendum 3 of Procedure OPMP01-ZA-0004 is missin o  It was unclear as to what " GAP (New)" and
      " Installed" meant in the table on the data sheet for Steps 5.1.3.4 through 5.1.3.7 when the procedure called for compression ring " gap" and
      " location".
 
Pending resolution of the above items, this is an open item (498/8708-53).
 
14) OPMP04-DG-0005, Revision 0, " Standby Diesel Generator Engine Maintenance" o  " Documentation" section as required by Addendum 3 of Procedure OPMP01-ZA-0004 was missin o  On Page 82, Step 5.7.5 currently reads 5. o  QIP designations were missing to left of QA/QC Rep line in Steps 5.16.16.1, 5.24.12, 5.29.13, 5.34.7.1, 5.34.12, and 5.38 on data shee o  Step 5.16.5 (on page 26) would be clearer with addition of "by tightening air starting valve nut."
 
- - _ _ _ _ _ _ _ _ _ _ _ _ _ _
 
. o
 
o QIP designations were missing to left of Step 5.33.5 in main body of procedure and on data shee o Step 5.29.8 regarding QIP should read 5.29.8 in lieu of 5.8.2 o Step 5.13.3 would be clarified it if referenced figure o Section 5.16 should have QC verify cleanliness if item would be enclosed after assembly, o Step 5.16.5 was unclear as to the desired torque valu o Clarity would be increased if during reassembly references to figures were include The applicant stated that this procedure was going to ,
be deleted and activities accomplished using other procedure Pending its deletion, this is an open item (498/8708-54).
 
15) OPMP04-ZG-0017, Revision 0, " Pacific Gate Valve Maintenance (Bolted Bonnet)"
o " Documentation" section as required by Addendum 3 of OPMP01-ZA-0004 was missin o It was unclear regarding which column was to be used on the torque chart in Addendum Pending correction of the above items, this is an open item (498/8708-55).
 
16) OPMP04-SN-0002, Revision 1, " Hydraulic Snubber Removal and Installation, Fluid Addition and Sampling" o Step 5.1.18 in the main body appeared as Step 5.1.17 on the data shee o " Documentation" section as required by Addendum 3 of OPMP01-ZA-0004 was missin ) OPMP04-SN-0006, Revision 0, " Anchor Darling Model 151 Mechanical Snubber Maintenance" The note and corresponding asterisks to the left of steps with data sheet entries was not used for all but one other of the procedures in the OPMP04 grou .
 
_
. .
 
Pending correction of the above item, this is an open item (498/8708-56).
 
18) OPMP04-SI-0002, Revision 1, "High Head Safety Injection Pump Maintenance" o After reviewing the corresponding vendor manual, there appeared to be a discrepancy in the acceptance criteria for Step 5.12.3.5 The criteria given was .008 to .012 inches and that stated in the vendor manual for the " Head and bowl bearing running clearance" was .013 to
  .020 inche o Step 5.13.6 in the procedure should be asterisked and the data sheet should contain entries pertaining to torquing of the fourth stage through seventeenth stage bowl fastener o After comparing the vendor manual torque chart to the procedure required torque values, the inspector identified discrepancies in the following steps:
  -
5.13.5.7 and 5.1 .13. .13. .13.10.10 o For the smaller torquing values called for in the procedure there was not a prerequisite for a torque wrench that has that value fall into the upper half of the scal Pending resolution of the above items, this is an open item (498/8707-57).
 
19) OPMP04-ZG-0006, Revision 3, "Limitorque Operator Removal and Installation"
" Documentation" section required by 0PMP01-ZA-0004 was missin ) OPMP04-RC-0007, Revision 0, " Pressurizer Spray l Valve Maintenance"   l
  " Documentation" section required by OPMP01-ZA-0004 was missin _
 
. . . _ _ _ _ . _ _ - .__ _ - ___ _ _ . _ _ _. _ ____ _ __
i .- .
. 63
.
;  21) Generic Comments i  o Some of the procedures appeared to have too many references. Elimination of references that are not necessary for the performance of the procedure would simplify the  '
procedure and make it more_ user oriente '
Examples of this were Procedure OPMP03-ZO-0004, i'  Revision 0 (40 references), .
Procedure OPGP03-ZM-0003, Revision 7 (40
,  references), and Procedure OPMP01-ZA-0004,
{    Revision 4 (11 references).
 
I o The NRC inspector investigated the j   licensee's controls over torquing of l-  safety-related fasteners. Torquing of
 
fasteners at STP followed the information
  ,
located in each component's respective
!  vendor manual. If there was no information I  given, but the determination that torquing of fasteners would be required, then Plant Procedure OPMP02-ZG-0004, " Fastener Torquing-l  and Detensioning," was required to be
;  utilized.
 
I l  The NRC inspector compared 10 vendor manuals
;    to their respective procedures out of a l    sample of 18 procedures. It was determined I    that the corrective maintenance procedures I    satisfactorily reflected the information contained in the manuals.
:  o The NRC inspector reviewed the licensee's l
process for adding quality control hold
;  points to corrective maintenance procedures.
 
;    In the procedures reviewed, the licensee had
"
hold points to verify cleanliness of the component and parts before enclosing the part and/or system. QC was in the review cycle, and had the option during maintenance work i    request review to add additional hold points
!  as required.
 
5  o Referencing other procedures and documents
. by revision number could be confusing, t  Administrative controls could be used to I    require use of the latest revision, except
'
in the cases where a specific revision to a document is applicable (i.e., a commitment).
 
i l
.
,
'.----...      *
 
_ - - _ _ . . . _ _ ''.,.
i i
    . ,
&
 
  (2) Procedures For Control of M&TE The NRC' inspector reviewed plant maintenance procedures to ascertain whether-the licensee's measuring and test control program adequately provides. controls for the calibration, testing, and checking of instrumentation and equipmen Calibrating and testing procedures were reviewed to verify that each was in the appropriate format as defined in administrative control procedures and was technically adequate to accomplish the stated purpos The NRC inspector reviewed Procedure OPGP03-ZM-0001, Revision 9, a  " Measuring and Test Equipment Control Program," which provides
,  program guidelines for the control of M&TE. For the purposes of
' J'  this procedure, M&TE does not include permanently installed
<
 
plant instrumentation. Calibration and testing of plant r instruments is addressed by " Plant Instrumentation Sealing Program," Revision 2. This procedure is applicable to devices which require calibration or testing by the instrumentation and i  controls section of the maintenance division. This procedure delineates controls for calibration and-status
;
verification / notification of installed permanent plant instruments. Thirteen calibration and testing procedures were i  reviewed to verify conformance to administrative guidelines and
  <
technical adequacy. The procedures reviewed were found to be
  '
technically adequate. They defined step-by-step instructions to accomplish their stated purpose. The NRC inspector did note some procedures needed to be reviewed and revised to conform to administrative guidelines. The NRC inspector held discussions with various personnel during the review of the procedures.
 
4c  These discussions indicated that they were familiar with program guidelines and requirements. NRC inspector observations were discussed with appropriate licensee personne (a) Procedures Reviewed
,
Procedure N Revision  Title IPMP05-YA-0001  1 " Vital Distribution Panel Tests" i
OPMP05-ZE-0104  0 " Frequency Transducer Calibration" OPMP05-ZE-0034  1 " Calibration of ITE-27 Relays"
{-
      " Insulation Resistance
'
OPMP05-ZE-0203  2
!
Testing-4.16K and 13.8K
'
Volt Motors"
  . ._ - - .._-_ ___  .. - _ _ . _ ~ _ - - - - _ _ _ _ _ - _ . - . . . - - ~-- _ _ . . _ -
 
_ _ _ _ - _ _ _ __-___-____ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
 
   ..    .
 
OPMP05-ZE-0206 0 " Potential Transformer Tests" OPMP06-ZT-0186' 2 " Calibration of the Westinghouse Relay Test Set" 0PMP06-ZT-0279 0 " Calibration of General Resistance RTD-100 RTD Simulator"
      .0PMP06-ZT-0293 0 " Calibration of Fluke 80I-600 Clamp-On Current Transformer"
'
OPMP07-SP-0001 0 "SSPS Decoder Printed Circuit Board Test and Rework" OPMP08-ZI-0011 2 " Generic Temperature Switch l        Calibration (Filled
      -
Element)"
OPMP08-ZI-0065 0 " Field Testing Of Power Supplies and Over Voltage Protectors" 0PMP08ZI-0203' 2 " Pressure or Differential Pressure Indicator
  '
Calibration"
, s  -n  IPMP08-SI-0861 0 "RHR/LHSI Pump.1.A Discharge
,
c,  -
Pressure Calibration lo.lJ-
'
A?
        (P-0861)"
  ', N f; OPGP03-ZM-0011 2 " Plant Instrumentation Scaling" 0PGP03-ZM-0016 0 " Installed Plant Instrumentation Calibration
      '
L g. ]gf      '
Verification Program" l      ,
      (b) NRC Inspector Observations / Concern _s
"
1) OPMP05-ZE-0104, Revisfeb u, " Frequency Transducer Calibration" o Step 6.6.5 data sheet should read " Transducer removed from service."
 
o Precautions should include instructions to verify equipment clearance, if applicabl .
 
_
. .
 
2) OPMP05-ZE-0034, Revision 1, " Calibration of ITE-27 Relays" Precautions should include instructions to verify equipment clearance,'if applicabl ) OPMP05-ZE-0203, Revision 2, " Insulation Resistance Testing-4.16 K and 13.8 K Volt Motors" o Steps 6.12 through 6.15 should be incorporated into Restoration / Documentation section o Procedure should specify actions to take if acceptance criteria is not me ) OPMP07-SP-0001, Revision 0, "SSPS Decoder Printed Circuit Board Test and Rework" o Procedure should include instructions on how to complete and process test documentatio o Precautions should include instructions to verify equipment clearance, if applicable, o Though this procedure is technically adequate to accomplish its stated purpose, it should not be used until it is revised due to inadequate procedure guidelines and instruction Procedure was in the review and revision process at.the time of the NRC inspectio Pending completion of this procedure revision, this is an open item (498/8708-58).
 
5) OPMP08-ZI-0011, Revision 2, " Generic Temperature Switch Calibration (Filled Element)"
o Precautions should include instructions to verify equipment clearance, if applicable, o Step 7.4.3 states " proceed to Step 7.3," which is incorrec ) IPMP08-SI-0861, Revision 0, "RHR/LHSI Pump 1A Discharge Pressure Calibration (P-0861)."
 
Addendum 1 does not identify V2 per Step 7.3. ) OPMP08-ZI-0065, Revision 0, " Field Testing of Power Supplies and,0vervoltage Protectors"
 
_ - .  .  ._    _
'
. .
4~
67-    l
 
l This procedure should be clarified in the following-areas:
o Procedure should define responsibilities of personne o Procedure should define actions to be taken if-acceptance criteria is not me o Precautions should include restoration / documentation instruction o Procedure should specify M&TE accuracy requirements.
 
'
o Procedure should identify references where
*
acceptance criteria can-be foun '
.'
o Procedure should not be used until a revision is issued do to inadequate procedure guidelines and instruction Revision 0 is in the review and revision proces Pending resolution of the above items, this is an open item (498/8708-59). Emergency Operating Procedures
;  (1)' Purpose
!    The purpose of this inspection was to determine whether E0Ps had been prepared in accordance with the PGP and whether they were technically adequate to control safety-related functions in the
;    event of system or component malfunction. At the time of this inspection, the PGP had been submitted to the NRC Office of NRR, but the NRC staff review of PGP was not yet complete.
:
i (2) Procedures Reviewed i    The following procedures were reviewed during this inspection:
o  OPGP03-ZA-0027, Revision 1, " Emergency Operating Procedures Preparation, Approval, and Implementation" o  OPOP01-ZA-0006, Revision 2, " Emergency Procedures Writers
,    Guide and Verification" i
o  1 POP 05-E0-E000, Revision 1, " Reactor Trip or Safety Injection"
,
'.
!
  , - , . . - , . . . _ . . . . _, , _ . _ _ ~ _ - . _ , - . . - -  _ _ . _ . . _ _ - - . . . . - . _ , . , _ . - . . . .
 
    . _
      ._-  _   . . . _
,
c.^ ,
-
S 68 s
o IPOP05-E0-E010, Revision 1, " Loss of Reactor or Secondary Coolant" o 1 POP 05-E0-EC00, Revision 1, " Loss of All AC Power" o IPOP05-E0-FRH1, Revision 1, " Response to Loss of Secondary Heat Sink" (3) Status Of Completi2n The NRC inspector compared the index of the applicant's E0Ps to the index of the 40G ERGS. This index indicated that E0Ps had been prepared ano approved for all WOG ERGS. One additional E0P was in preparati>n in response to the fire hazards analysi This E0P was scheduled to be approved prior to loading fuel.
 
i   ,
  (4) Technical Adequicy The NRC inspector assessed E0P technical adequacy by comparing the E0Ps to the WOG ERGS and plant piping and instrumentation drawings. While'no major technical inadequacies were
<
identified, several errors and discrepancies indicating the    ,
  -applicant's failure to develop E0Ps appropriate to the
. circumstances as required by 10 CFR 50 Appendix B, Criterion V, I  were identified. These are listed below:
  (a) Statior, Procedure 1 POP 05-E0-E000, " Reactor Trip or Safety Injection" o Step 10 did not state how many essential cooling water pumps should be runnin o Step 11.1 required ' transfer' of reactor containment fan cooling to component cooling water. This step should have verified that automatic transfer had
,
occurred.
 
l j    o Step 22.2 listed incorrect equipment designations for the pressurizer spray valves,
#
o Step 31.2 incorrectly directed the secondary operator to sample the steam generators. This function should be performed by plant chemists, as directed by the Unit Superviso (b) Station Procedure IPOP05-E0-E010, " Loss of Reactor or
.
Secondary Coolant," had an incorrect reactor coolant system pressure referenced in Entry Condition 3.
 
.
;
e e v -1* y,.-.~.-, - - .--- ,. .-,-y ,~. _.. - .,____, e--,-,..,,,, ,-,..r,,,,_,2 . - _ . _ _.---._,.-,.r._..-2.-_,,, c.,.v. - - r m.m.-e p r
; - .
 
  (c) Station Procedure IPOP05-E0-EC00, " Loss of All AC Power,"
failed to include the desired entry step of the referenced procedure in the contingency action statement of Step 2 (d) Station Procedure IP0F05-E0-FRH1, "Resperse to Loss of Secondary Heat Sink."
 
o The step 9 caution statement stated, ". . . establish RCS heat by RCS bleed and feed." The word ' Removal'
should be inserted after ' heat'.
o The Step 19 caution statement contained an incorrect procedure number referenc o The note before Step 22 was repeated before Step 23, deleting the proper note which should have been placed before Step 23, "After closing a PORV, it may be
  .necessary to wait for RCS pressure to increase to permit stopping high-head SI pumps in Step 22."
 
Pending licensee response, the above deficiencies will be tracked as an open item (498/8708-60).
 
(5) Deviations from WOG ERGS Section II of the STP PGP dated May 16, 1985, stated, "As the procedure writer prepared the E0P, deviation between the WOG ERGS and the plant specific procedure caused by plant design or preferred due to control board layout were documented on form
! (4) of the writer's guid Reasons for deviations were also documented on the same form." Section 2.3.1 of Station Procedure OPOP01-ZA-0006, " Emergency Procedure Writers Guide and Verification," states, "When preparing the E0Ps, situations may arise where the intent specified by the guidelines may have to be altered . . . due to the STP design. When this happens, the
, writer shall complete the documentation for changes to the WOG
'
Guidelines or 'EOP Step Justification / Verification Form' (-4)."
 
; The NRC inspector reviewed the E0P verification packages for Revisions 0 and 1 of the E0Ps listed above. The Step l'
Justification / Verification (SJ/V) forms completed during procedure preparation and revision were included in these package For a small number of deviations from the WOG ERGS, the NRC inspector found that the SJ/V forms provided a very good
. justification for the deviation. However, many deviations from
! the WOG ERGS were not documented and justified on SJ/V forms and
'
there were inadequate basis or justification for scme of the deviations which were documented on SJ/V form Pending
; licensee resolution, the lack of documentation o' deviations from the WOG ERGS will be tracked as an open item (498/8708-61).
 
.
 
a
. ..
 
The NRC inspector found that SJ/V forms were generally not used to document and provide a basis for plant specific information
  ~
inserted into the E0Ps where the WOG ERGS used a notation such as, " Establish main feedwater flow [-Enter plant specific means]."
It should be noted that the lack of SJ/V forms was frequently
, documented by. reviewers in the procedure verification process using E0P Discrepancy / Comment forms from Station Procedure OPOP01-ZA-0006. In these cases the resolution of the comment provided some justification of the deviation from the WOG ERG (6)' Plant Specific Values One plant specific value from each of the four E0Ps was selected for verification. The reference for plant specific values.was the HL&P Emergency Operation Procedure Setpoint Document, Revision 1, dated November 10, 1986. No problems were identified in this verification. Applicant representatives informed the NRC inspector that Revision 2 of the Setpoint Document has been issued but not yet incorporated into the E0P They plan to incorporate the latest setpoints into-the E0Ps prior to loading fue (7) Compliance With Writers Guide The NRC inspector reviewed the four E0Ps listed above to determine whether they had been written in accordance with the guidance provided in Station Procedure OPOP01-ZA-0006,
" Emergency Procedure Writers Guide and Verification." General conformance was noted with the exceptions listed below:
(a) Section 3.1.2.1 of the writers guide required that each operator copy of an E0P present the user information and steps on the left page and the non-user information on the right page when opened. The NRC inspector noted that action steps identified as being performed by the Unit Supervisor were not included with the non-user information on the right page of the operator's copies of the EOP (b) Section 5.4 of the writers guide required maintenance of a direct horizontal relationship between the related action steps in the left column and the contingency action steps in the right column. The NRC inspector found that the contingency action step associated with action Step 5.3 of Station Procedure IPOP05-EO-E010 was aligned horizontally with action Step (c) Section 17.4 of the writers guide required that missing information shall be listed on a separate punchlist at the
 
. ,
 
end of the written procedure bod The NRC inspector found that no punchlist was attached to Station Procedure 1 POP 05-E0-EC00 although Steps 15 and 16 of this procedure were missing information which should have been identified on a punchlist. The missing information related to contingency actions for filling the auxiliary feedwater storage tank using the fire water system and local operation of steam dump valve (d) Section 18.2 of the writers guide required that the appropriate emergency action level be entered into the E0P at the earliest possible poin The NRC inspector found that Station Procedure IPOP05-EC-FHR1 contained no reference to emergency action levels which had been reached. Station Procedure OEPP01-ZA-0001, " Emergency Classification," Addendum 3, indicated that reaching Step 9.0 of 1 POP 05-E0-FRH1 was the emergency action level for declaration of a General Emergenc Pending licensee response, the above exceptions to the writers guide will be tracked as an open item (498/8708-62).
 
(8) Verification and Validation The NRC inspector reviewed the verification packages for Revisions 0 and 1 of the E0Ps listed above, the Emergency Operating Procedure Validation Report dated December 22, 1986, and the checklists and deficiency sheets associated with the validation progra An applicant representative stated that the final validation report was in preparation at the time of this inspectio It appeared that the verification and validation program was conducted in accordance with the PGP and associated plant procedures. However, some weakness in this program was indicated by the dis ~crepancies discussed abov .
(9) Other Comments (a) During the review of E0P verification packages, the NRC inspector noted that one individual signed a License Compliance Review Form (OPGP03-ZA-0003-1) as both preparer and reviewer. While this action was not prohibited by Plant Procedure OPGP03-ZA-0003, the NRC inspector stated that this was not a generally accepted practice. Applicant representatives stated that they had recognized this as a problem and that corrective action was underwa (b) Step 6.3.1.6 of Station Procedure OPGP03-ZA-0027 appeared to be incomplet (c) The E0Ps included no statement of purpose or scope.
 
(  NUREG-0899, Section 5.4.3 states, "Each E0P should contain l
l
. .. , _
_ _ _ --
 
  . . . -  . . -- - . ~ . .  .- ~ .  -
l'
. .
 
.
;
a brief statement that describes what it is intended to accomplis In many cases it may be possible to include the scope in the title of the E0P without making the title too long." While the PGP indicated that the writers guide
;    was based on NUREG-0899 (and other references), the writers guide did not require E0Ps to have a statement of purpose
;    or scope. The NRC inspector noted that the WOG ERGS each i    begin with a statement of purpose, some of which provide considerably more information about the purpose of th procedure than does the procedure titl This comment is
,
expected to be resolved during the process of NRC review and approval of the PGP.
 
i Off-Normal Operating and Alarm Response' Procedures The NRC inspector reviewed selected applicant off-normal. operating j-  procedures and annunciator response procedures to verify they were in
'
the required format and that they were technically adequate to perform the designated function. The NRC inspector walked down selected procedures to verify that the 'as-built' conditions were compatible with the plant procedures.
 
.
  (1) Procedures Reviewed (a) Off-Normal Procedures
              ,
t              ~
Procedure N Revision  Title
              *
1 POP 04-RC-0001  2  "High Reactor Coolant
,
System Activity" i
1 POP 04-FW-0001  1  " Loss of Feedwater Flow or Control" 1 POP 04-CR-0001  1  " Main Condenser-Loss of Vacuum Off-Normal'
Procedures" I    IPOP04-RC-0002  1  " Loss of Reactor Coolant Pump"
!
    (b) Operating Procedures a
 
Procedure N Revision  Title 4    1 POP 02-CV-0004  1  " Chemical & Volume i          Control System"
;
 
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_ _ _
    . ~ _ - - . . . . .  .  . . - . --. .-- -. .-  .  .- -.
 
  . .
 
    - (c) Annunciator Response Instructions IPOP09-AN-04M8-D-4  0 "LETDN HX OUTLET FLOW HI/LO" 1 POP 09-AN-04M8-D-3  0 "LETDN HX TEMP HI DEMIN DVRT" 1 POP 09-AN-04M8-D-1  0 " SEAL WTR INJ FILTER DR HI" 1 POP 09-AN-04M8-C-4  0 "LETDN HX OUTLET PRESS HI" 1 POP 09-AN-04M8-C-3  0 "LETDN HX OUTLET TEMP    ,
HI"
 
1 POP 09-AN-04M8-C-2  0 "BTS DEMIN DP HI" 1 POP 09-AN-05M2-E-1  0 "RCP CCW FLOW LO" 1 POP 09-AN-05M2-E-2  0 "RCP TRIP" 1 POP 09-AN-05M2-E-3  0 "Rx VSL FLNGE LEAK TEMP HI" 1 POP 09-AN-05M2-C-1 thru 0  "RCP UPPR OIL RSVR LVL IPOP09-AN-05M2-C-4    HI/LO" 1 POP 09-AN-05M2-D-1 thru 0  "RCP LOWR OIL RSVR LVL-1P0P09-AN-05M2-D-4    HI/LO" 1P0P09-AN-07M3-E-7  0 " MAIN COND VACUUM LO" 1 POP 09-AN-07M3-F-6  0 "F.W. S/V PMP L. O. AUX PMP TRBL"
,
IPOP09-AM-07M3-F-8  0 " SEAL LEAK OFF TNK LVL l          HI" l
l These annunciator response instructions were walked down in j-    the plant.
 
'  (2) NRC Inspector Observations / Concerns i
.    (a) IPOP02-CV-0004, Revision 1, " Chemical and Volume Control      l
:    System"
!
i
!
!
i
:
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-. ..
 
o In Step 11.3 there is an incorrect valve designation in the valve lineup in that in Cation Bed Demin IA valve line up should read "1*CV-129A" in lieu of
  "1*CV-129B".
 
o In Step 11.5 'MAB' should be 'MEAB'.
(b) IPOP04-FW-0001, Revision 1, " Loss of Feedwater Flow or Control" According to the instructions in Steps 4.3 and 4.4 of
' Procedure OPOP01-ZA-0007, Revision 1, "Off-Normal Procedures Writer's Guide," there is a conflicting use of
" Note" versus " Caution" statements. An example of this is the following " Note" located after Step 4.3 which should be a " Caution" statement:
" Reducing turbine load too rapidly may cause an unnecessary reactor trip due to the effects of SG shrink."
 
Pending resolution, this is an open item (498/8708-63).
 
(c) IPOP04-RC-0002, Revision 1, " Loss of Reactor Coolant Pump" The following " Note" which occurs at Step 4.0 in the procedures is an instruction statement (i.e., action step).
 
This is contrary to the instructions in Step 4.4 of OPOP01-ZA-0007, Revision 1, "Off-Normal Procedures Writer's Guide," which states that notes should not be used as instruction statement " Note: If Rx pwr is above P-8 and conditions exist calling for an immediate RCP trip then trip the reactor first then the RCP to ensure heat removal. I_f Rx power is less than P-8, trip the RCP only."
 
Pending resolution, this is an open item (498/8708-64).
 
(d) OPOP09-AN-05M2-E-1, Revision 0, "RCP CCW Flow LO" The immediate actions section (below) of the procedure should also verify the remainder of the CCW system is in service (i.e., per a POP or by looking at a control panel).
 
" Verify 1*CC-MOV-318, 1*CC-MOV-029, inlet isol. are open and 1*CC-MOV-403, 1*CC-FV-4493, 1*CC-MOV-404, 1*CC-MOV-542, outlet isolation valves are open."
 
(e) 1 POP 39-AN-05M2-E-3, Revision 0, "Rx VSL FLNGE LEAK TEMP HI"
 
_ - ____ - _ _____ -_____________-___ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _
o  o
 
3rocedure states " Place HS-3400 to 'Close' position as firected by Unit Supervisor," however, the switch is not libeled as HS-3400 on the control board. Applicant has aided this item to the labels correction lo (f) IPOP09-AM-07M3-F-8, Revision 0, " SEAL LEAKOFF TANK LVL HI" Supplementary actions states to " Manually control level using Bypass Valve 1-FW-0483;" however, Valve 1-FW-0483 is installed under floor grating that makes it not readily accessibl Pending resolution, this is an open item (498/8708-65).
 
(g) In general, action statements are inconsistent in the amount of detail provided for performing the action. Some of the action statements only give general guidance as in Step 5.1 of IPOPO4-FW-0001, Revision 1, " Loss of Feedwater Flow or Control" which states:
      " Ensure steam dumps are operating properly and verify that T
AVG is being matched with T Ref' "
This action statement should be more specific by referencing instruments to be monitored. Examples of action statements which provide specific instructions are Steps 4.1 and 5.1 of Procedure IPOP04-CR-0001, Revision 1,
      " Main Condenser-Loss of Vacuum Off-Normal Procedures,"
which state respectively:
o  "4.1, Verify all vacuum pumps operating (hogging)
        (CD-009)."
 
o  "5.1, Verify steam supply to turbine seals (CP-008)."
 
Pending resolution, this is an open item (498/8708-66). Surveillance Procedures The NRC inspectors performed the following activities to verify that the applicant had established adequate procedures to perform required TS surveillances:
o Compared the proof-and-review version of STP Unit 1 TSs to the STP index of surveillance procedures to verify that the applicant had established or was establishing a procedure to accomplish each surveillance required by the TS o Performed in-depth review of selected effective / approved surveillance procedures to verify that TS surveillance requirements were satisfie _ - _ _ _ _ _ _
 
r a .
 
o Performed in plant walkdowns of selected surveillance procedures to verify that as-built equipment and indicators reflected the TS requirement (1) Procedures Reviewed (a) Administrative Procedure N Revision Title OPGP03-ZE-0005 3 " Plant Surveillance Procedure Preparation" (b) Instrumentation and Control (I&C) Functional Procedure N Revision Title
  *1 PSP 02-EH-6328, 0 " Turbine Thrott!e Valve TA00T" 1 PSP 02-FW-0574 0 "SG-ID Narrow Range Level Set 1 ACOT" 1 PSP 02-HC-0935 0 " Containment Pressure Set 3 ACOT" 1 PSP 02-MS-0506 1 " Turbine Impulse Chamber Pressure Set 2 ACOT"
  *1 PSP 02-NI-0042 0 " Power Range Neutron Flux Channel II ACOT" 1 PSP 02-RC-0427 0 "RCS Flow Loop 2 Set 1 ACOT" 1 PSP 02-SI-0952 0 " Accumulator 18 Level Group IV ACOT" 1 PSP 02-SP-0001R 0 "SSPS Logic Train R Functional Test" 1 PSP 02-RC-0452 0 "RCS Temperature Loop II Set 1 ACOT" 1 PSP 02-RC-0466 0 " Pressurizer Level Set II ACOT" (c) System and Component IPSP03-AF-0004 1 "AFW Pump II Reference Value Measure"
 
r a o-
 
1 PSP 03-SI-0017 0 " Containment Spray Valve Checklist" 1 PSP 03-CV-0007 0 " Boric Acid Transfer Pump 1A Reference Valves Measures" 1 PSP 03-CV-0010 0 "Boration Flow Verification" 1 PSP 03-0G-0011 1 " Standby Diesel IZ Auto Start on ESF Actuation Test Signal" 1 PSP 03-EA-0002 0 "ESF Power Availability" 1 PSP 03-FP-0001 0 " Fire Protection System Valve Operability Test" 1 PSP 03-RC-0006 0 " Reactor Coolant Inventory"
*1 PSP 03-RH-0005 0 " Residual Heat Removal Pump 1B Reference Valve I Measurement"
*1 PSP 03-RS-0002 0 " Manual Reactor Trip TADOT" 1 PSP 03-SI-0013 0 " Accumulator Isolation Valve Verification"
*1 PSP 03-SP-0001 0 " Remote Shutdown Monitoring Instrumentation Channel Check" 1 PSP 03-0G-0003 1 " Standby Diesel B Operability Test" 1 PSP 03-RM-0001 1 "Raactor Makeup Water System Valve Operability Test" (d) I&C Calibration
*1 PSP 05-AF-7524 0 "AFW Flow Loop II Channel B Calibration" 1 PSP 05-FW-0503 0 "SG-IC Wide Range Level Channel B Calibration" 1 PSP 05-AC-0936 0 " Feed Flow Loop I Set 3 Calibration" 1 PSP 05-NI-0031 0 " Source Range Channel I Calibration"
  . _ _ _ _ - - __
    .- . - . _
 
o    .
 
1 PSP 05-NI-0044 0 " Power Range Channel IV Calibration" 1 PSP 05-RC-0417 0 "RCS Flow Loop I Set I Calibration" 1 PSP 05-RC-0451 0 "RCS Temperature Loop I Set I Calibration" 1 PSP 05-RC-0458 0 " Pressurizer Pressure Set IV Calibration" 1 PSP 05-RC-0466 0 " Pressurizer Level Set II Calibration" 1 PSP 05-SI-0954 1 " Accumulator 1C Level Group III Calibration" 1 PSP 05-WL-0478 0 " Plant Liquid Waste Discharge Flow Calibration" 1 PSP 05-CC-4503 b 0 "CCS Surge Tank Compartment A Level Switch Calibration" (e) Electrical IPSP06-DG-0001 0 "Undervoltage Loss of Relay Voltage Channel Calibration" 1 PSP 06-DJ-0004 0 "125 V Class 1E Battery Service Surveillance Test"
      * Procedures selected for in plant walkdow (2) NRC Inspector Observations / Concerns (a) 1 PSP 03-SP-0001, Revision 1, " Remote Shutdown Monitoring Instrumentation Channel Check" The acceptance criteria in Step 7-1 on channel checks needs to be more specifi Tolerances should be included so the test parformer knows when an indicator is not functioning properly. Presently the only criteria is that an indication exist The NRC inspector determined from discussions with applicant personnel that the applicant feels that since a quantitative assessment of the channel behavior is not required by the TS that a tolerance is not desired. The procedure does include a step that states that if in the operators judgement the indications are in error he is to report that information to the shift supervisor.
 
_ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
 
r m
 
    (b) IPSP03-RH-0005, Revision 0, " Residual Heat Removal Pump 18 Reference Values Measurement" This procedure lines up the system to run the RHR pump with the suction valve closed. This did not appear to be a good practice since the small volume of water within that closed system could heat up rapidly and could cause cavitation and possible water hammer problems. The NRC inspector determined from discussion with applicant personnel that the system design allows for cooling by the CCW system of the water being pumpe The utility feels that this will be sufficient to prevent cavitation problems due to heatu (c) OPGP03-ZE-0005, Revision 3, " Plant Surveillance Procedure Preparation" Step 3.2.4.b specifies that any LC0 which may be entered during the performance of a surveillance test be inserted in the pretest verification section of the procedure. This was omitted from many of the procedures that were reviewe Applicant personnel stated that during the next revision to PGP03-ZE-0005, " Plant Surveillance Procedure Preparation,"
the instructions for the contents of the pre-test verification section will be modifie Pending this revision, this is an open item (498/8708-67).
 
(d) 1 PSP 03-RC-0006, Revision 0, " Reactor Coolant Inventory,"
Procedure does not specifically address " leakage to RCP seals." This leakage to the seals is stated in TS Surveillance 4.4.6.2.1.c. Applicant states that TS Surveillance 4.4.6.2.1.c is being delete Pending this deletion and correcting the procedure, if required, this is an open item (498/8708-68).
 
(e) IPSP03-SI-0013, Revision 0, " Accumulation Isolation Valve Verification" Step 5.3 states that an indicating light on the main control board is used to verify power removed to the valve operato This verification should be done by breaker positio The NRC inspector determined from discussion with applicant personnel that the design of the controls on the main control board provides for removal of control power to the isolation valve and indication of such removal and valve position. This is sufficient to ensure that power has been removed to the isolation valv (f) IPSP03-SP-0001, Revision 0, " Remote Shutdown Monitoring Instrumentation Channel Check"
_ _ - _ _ _ _ _ _ _ _ - _ _ _ - _ - _ _ _ _ _
 
r-w .
 
Performance frequency for this procedure was listed as quarterly in the surveillance procedure index computer printout; however, TS 4.3.3.5, Table 4.3-6, requires the surveillance to be performed monthly. Applicant revised the surveillance testing master index so that the frequency for this procedure was changed to monthly to comply with TS requirement (g) The NRC inspector observed that QC should be involved in performance of surveillance procedures. It was noted that some procedures have QC involvement but most do not. The NRC inspector determined from discussions with applicant personnel that it is the applicants philosophy to have QC be on the worker level. The applicant feels that the training given the worker is sufficient to ensure quality control and it is not necessary to include QC on all procedure The quality assurance department spot checks the performance of procedure No violations or deviations were identifie . Exit Interview The NRC resident inspectors met with -licensee representatives (denoted in paragraph 1) on April 10, 1987, and summarized the scope and findings of the inspection. Other meetings between NRC inspectors and licensee management were held periodically during the inspection to discuss identified concerns.


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Revision as of 14:00, 22 February 2021

Ack Receipt of 870625 Ltr Informing NRC of Steps Taken to Correct Violations Noted in Insp Repts 50-498/87-08 & 50-499/87-08
ML20236D718
Person / Time
Site: South Texas  STP Nuclear Operating Company icon.png
Issue date: 07/24/1987
From: Gagliardo J
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To: Goldberg J
HOUSTON LIGHTING & POWER CO.
References
NUDOCS 8707310038
Download: ML20236D718 (3)
v • d • e


Text

_

.. _

jd#"% UNITED STATES y" - ~, NUCLEAR REGULATORY COMMISSION b REGloN IV

8 611 RYAN PLAZA DRIVE. SUITE 1000

, ,

ARLINGTON, TEXAS 76011

- JUL k. 4 19ef In Reply Refer To:

Dockets: 50-498/87-08 50-499/87-08 Houston Lighting & Power Company ATTN: J. H. Goldberg, Group Vice President, Nuclear P. D. Box 1700 Houston, Texas 77001 Gentlemen:

<

Thank you for your letter of June 25, 1987, in response to our letter and l

NoticeofViolation(498/8708-01) dated May 29, 1987. We have reviewed your reply and find it responsive to the concerns raised in our Notice of Violatio We will review the implementation of your corrective actions during a future inspection to determine that full compliance has been achieved and will be l maintaine Sincere f, l l

,

> . MM K E. G gliardo, Chief

..//V Reacto Projects Branch l

l cc:

Houston Lighting & Power Company ATTN: M. Wisenberg, Manager, Nuclear Licensing P. O. Box 1700 Houston, Texas 77001 Houston Lighting & Power Company ATTN: Gerald E. Vaughn, Vice President Nuclear Operations P. O. Box 1700  ;

Houston, Texas 77001 p[

'

Texas Radiation Control Program Director i }

l Q731003aa70724 i

o ^ DUCK 05000498 PDR

'

,

'

In Reply Refer To: dRil 2 4 Md7 Dockets: 50-498/87-08 50-499/87-08 Houston Lighting & Power Company ATTN: J. H. Goldberg, Group Vice President, Nuclear P. 0. Box 1700 Houston, Texas 77001 Gentlemen:

Thank you for your letter of June 25, 1987, in response to our letter and Notice of Violation (498/8708-01) dated May 29, 1987. We have reviewed your reply and find it responsive to the concerns raised in our Notice of Violatio We will review the implementation of your corrective actions during a future inspection to determine that full compliance has been achieved and will be maintaine

Sincerely, Cricinal Signal aj

.

'd+P..Jaudon J. E. Gagliardo, Chief Reactor Projects Branch cc:

Houston Lighting & Power Company ATTN: M. Wisenberg, Manager, Nuclear Licensing P. O. Box 1700 Houston, Texas 77001 Houston Lighting & Power Company ATTN: Gerald E. Vaughn, Vice President Nuclear Operations P. O. Box 1700 Houston, Texas 77001 Texas Radiation Control Program Director f,/

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bcc: (seenextpage)

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HFBundy:cs ' GLConstable A/JEGkglidrdo 7/zt/87 7/n-/87 7g4f87

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Houston Lighting & Power Company -2-bcc to DM8 (IE01)

bec distrib. by RIV:

RPB DRSP RRI-0PS R. D. Martin, RA RRI-CONS Section Chiaf (RPB/C)

RPSB MIS System j RIV File D. Weiss, RM/ALF RSTS Operator R. Pirfo, OGC i R. G. Taylor, RPB/C RSB ]

Project Inspector, RP8 R. Hall 1 P. Kadambi, NRR Project Manager i l

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