IR 05000498/1987023
| ML20234E745 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 06/29/1987 |
| From: | Bundy H, Carpenter D, Constable G, Hildebrand E, Reis T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20234E715 | List: |
| References | |
| TASK-1.A.1.3, TASK-1.C.4, TASK-2.B.4, TASK-2.D.3, TASK-2.E.1.1, TASK-2.G.1, TASK-2.K.1, TASK-2.K.3.05, TASK-2.K.3.09, TASK-2.K.3.10, TASK-2.K.3.25, TASK-TM 50-498-87-23, 50-499-87-23, GL-83-28, IEB-82-02, IEB-85-001, IEB-85-002, IEC-77-15, IEIN-85-093, NUDOCS 8707070668 | |
| Download: ML20234E745 (19) | |
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APPENDIX U.S. NUCLEAR REGULATORY COMMISSION l
REGION IV
NRC Inspection Report:
50-498/87-23 Construction Permits: CPPR 128 50-498/87-23 CPPR 129 j
Dockets:
50-498 Expiration Date: December 1987 and l
50-499 December 1989 Licensee:
Houston Lighting & Power Company (HL&P)
l P. O. Box 1700 Houston, Texas 77001 Facility Name:
South Texhs Project, Units 1 and 2 (STP)
l Inspection At: STP, Matagorda County, Texas
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Inspection Conducted: April 11 through May 4, 1987
Inspectors:
V g/25/17 D. R. Carpenter, 5fnior Resident Inspector Date Project Section C, Reactor Projects Branch
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T. Reis, Resident Inspector, Project Date Section C, Reactor Projects Brar.:h j
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C/27/ft7 H. F. Bundy, Project Inspector, Project Date
Section C, Reactor Projects Branch j
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f.. H. Hildebrand, R'eactor Inspector, Reactor Date Operations Section, Reactor Safety Branch 8707070666, 870630 HDDCK 050 g 8 PDR i
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i Other NRC l
Personnel:
J. F. Lara, C0-0P Student J. P. Clausner, Engineer, CEA France Consultants:
N. Jensen, M. Bishop; EG&G Idaho, Inc.
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Approved:
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/'/'4. L. Constable Chief, Project' Section C Dafe /
Reactor Projects Branch e
Inspection Summary i
Inspection Conducted April 11 through May 4,1987 (Report 50-498/87-23; 50-499/87-23)
i Areas Inspected:
Routine, unannounced inspection including licensee action on
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previous inspection findings; licensee action on previously reported deficiencies;theinservicetesting) program;theplantsafetyreviewand Generic Letter (GL) 83
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l committees; Three Mile Island (TMI followup; preoperational test procedure review; preoperational test witnessing; l
IE Bulletins, Circulars, and Notices; and site tours.
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Results: Within the areas inspected, no violations or deviations were l
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identified.
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DETAILS l
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Persons Contacted Principal Licensee Employees l
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- W. H. Kinsey, Plant Manager j
- W. G. Isereau, Operations Quality Assurance Supervisor
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- S. M. Head, Lead Engineer, Licensing
- K. L. Trippel, Lead Engineer, Technical Support l
- H. S. Blinka, Supervisor Engineer, Technical Support
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- B. R. Waahlheim, Lead Engineer, Technical Support
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- R. Daly, Startup Manager
- M. R. Rejcek, Chemical Operations
- E. L. Brown, Licensing Engineer
- G. Ondriska, Lead Engineer, Startup
- T. Underwood, Chemical Operations Manager In addition to the above, the NRC inspector also held discussions with l
various licensee, architect engineer (AE), constructor, and other contractor personnel during this inspection.
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- Denotes those individuals attending the exit interview conducted on May 4, 1987 2.
Licensee Action on Previous Inspection Findings (Closed) Open Item 498;499/8708-04, Conflicting Verification Signatures l
This open item concerned preoperational test procedures where documentation showed that quality control (QC) and test' engineering personnel verified specific action statements on differing dates.
Statements were obtained by the QC inspectors that the contradicting dates were an administrative error and the inspections were performed at the proper time.
Personnel training will be conducted by startup and QC organizations addressing " attention to detail." This item is closed.
(Closed) Violation 498/8626-01;499/8624-01, Failure to Follow Procedures for Dealing with Nonconforming Items (IRC 339)
This violation dealt with a failure to follow procedures by the licensee in dealing with nonconforming items.
Several examples were' identified by NRC inspectors where the licensee had not followed the detail of Standard Site Procedure (SSP) 8, "Nonconformance Reporting." The licensee performed a detailed evaluation of all site activities that dealt with nonconformance repwting and hold tag application. This evaluation identified seven areas where inadequate attention to the requirements of SSP 8 were noted.
The NRC inspector concluded that this did not involve a breakdown of the nonconformance report (NCR) system but rather identified
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isolated cases where the system had failed or the opportunity for failure existed.
For this violation, the root causes were failure to implement
the Transfer and Replacement Request.(1RR) system where NCRs were involved and inadequacies in SSP 8 for dealing with nonconforming items still in the warehouse after the system was released for test to the startup organization. The licensee has. revised and reissued SSP 8 to insure'all requirements are identified, including special cases; performed a detailed plant and warehouse audit of NCRs and hold tags; and has provided additional training to those involved in the NCR process. Also SSP 48,
" Equipment and Component Interchange," was revised to ensure the J
requirements of SSP 8 were considered.
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The NRC inspector has reviewed the revisions of SSP 8 and 48 and randomly checked the warehouse and plant areas for compliances with acceptable results. The licensee has satisfactorily resolved the issues of this violation and it is considered closed.
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Licensee Action on Previously Reported Deficiencies (Closed) Incident Review Committee (IRC) 198 Deficiency in Comsip Incorporated Standard Bed Catalyst This IRC package originated from a 10 CFR Part 21 report generated by Comsip, Incorporated, dated April 26, 1984, IE Notice (IEN) 84-22 and a 10 CFR 50.55(e) of May 16, 1984. The licensee's investigation of this issue indicated that STP did in fact have the Comsip K-III standard catalyst beds.. These catalyst beds could have a significantly shorter useful life under worst case loss of coolant accident conditions. The licensee ordered, received, installed and tested a modification kit provided by Comsip Incorporated which modified the catalyst bed assembly and provided other system enhancements. The hydrogen analyzers now function as required by specification. The NRC inspector has reviewed the documentation and the modified system and considers this item satisfactorily resolved.
(Closed) IRC 190 - Main Control Room Habitability During a Bechtel design review of the control room, heating, ventilation and air conditioning (HVAC), it was noted that if a toxic chemical release occurred concurrent with a loss of offsite power (LOOP) the control room habitability requirements (GDC 19) were not satisfied. This was due to a system alignment which did not stop the makeup air filtration trains.
A design change has been implemented which deletes the start of makeup air i
unit fans by load sequence mode II (LOOP). The result is that no makeup air is provided. The habitability requirements of 10 CFR 50, Appendix A, GDC 19 are satisfied.
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(Closed) IRC 136 - Westinghouse EMD Gate Valve Limit Switch A design deficiency was discovered which concerned the design of internal limit switches used for gate valves supplied by Westinghouse for the high head and low head safety injection system valves, reactor compartment fan cooler (RCFC) chilled water valves, and component cooling water (CCW) heat exchanger bypass valves.
Because of the internal 1 4it switch design, the potential existed for the control room to receive a fully closed position indication when the valve (s) were still partially open.
This situation was corrected by moving the "still open" limit switch to a different rotor whose setpoint could be established at the " fully closed" position.
j This action mitigates any potential safety hazard which might exist from an unanalyzed condition caused by improper valve. position indication, l
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Inservice Testing (IST) Program The STP IST Program was reviewed to ensure compliance with 10 CFR 50.55.a(g). The following documents were among those reviewed to verify compliance with regulatory requirements and licensee commitments:
NUREG 0781, STP Safety Evaluation Report together with Supplement 2.
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STP Pump and Valve Inservice Testing Plan, Revision 1
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OPGP03-ZA-002, Revision 8, Plant Procedures
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OPGP03-ZA-0020, Revision 1, Plant Surveillance Scheduling
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OPGP03-ZE-0022, Revision 1, Inservice Testing Program for Pumps
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OPGP03-ZE-0021, Revision 0, Inservice Testing Program for Valves
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IPSP03-CS-0006, Revision 0, Containment Spray Pump IC Reference
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Valves Measurement IPSP03-SI-0004, High Head Safety Injection Pump 1A Inservice Test
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IPSP03-SI-0003, Revision 0, Low Head Safety Injection Pump IC
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Inservice Test IPSP03-AF-0007, Revision 1, Auxiliary Feedwater Pump 14 Inner ice
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Test OPSP11-ZA-0005, Revision 1, Local Leakage Rate Test' Calculations and j
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Guidelines
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6 OPGP03-ZE-0011, Revision 2, Containment Leakage Rate Testing' Program
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IPSP03-CC-0008, Revision 1, Component Cooling Water System Train 1B
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Valve Operability Test The program appeared to meet the stated requirements. However, it had not been fully implemented. The following items will remain open pending availability for inspection:
j Test acce ?,ance criteria must be added to procedures, e.g.
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IPSP03-afb 010.
(498/8723-01)
Procedures for testing safety relief valve set points must be
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approved.
(498/8723-02)
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Procedures for checking pump and valve indication at the remote I
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l shutdown panel must be prepared.
(498/8723-03)
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IST for designated pumps and valves must be performed prior to fuel
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l load per licensee's schedule.
(498/8723-04)
No violations or deviations were identifie 5.
Plant Safety Review Committee The NRC inspectors reviewed the activities of two safety groups for STP.
l These were the Nuclear Safety Review Board (NSRB) and the Plant Operations Review Committee (PORC). The inspector attended two PORC meetings and one
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l NSRB meeting. Additionally, the meeting minutes of the preceeding
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meetings of each group were reviewed. Conmiittee charters and procedures were examined for completeness and compliance. These committees are functioning effectively during this phase of plant activities and both are filling their required functions. Below, however, are listed the results
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of the documentation review. The open items need to be resolved before the inspectors can conclude that these groups are fully satisfactory.
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Procedure NGP820, NSRB Policy
Step 3.1.2 requires appointment of only four members, but five
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members are required for a quorum in Step 9.2.
Members are not specified in Step 3.1.2 as reflected in Technical
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Specification (TS) 6.5.2.2.
Procedure does not address whether or not the chairman can vote and
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splits him out from " members."
Resolution of the above items remains an open item.
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Procedure NSRB01, NSRB Rules of Conduct Procedure does not require " Unscheduled Meeting" minutes to be
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provided to Group Vice President - Nuclear within 14 days following meeting in Step 6.6.4 as required by TS 6.5.2.9.a.
Items 6.5.2.f and 6.5.2.h do not appear on Attachment I as required
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by Step 6.5.2.
There are no provisions for appointment' of NSRB Administrator or no
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job qualifications for position in procedure. He is assigned
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significant duties by this procedure and will need some
qualifications for the position.
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Step 6.6.4 does not reflect the TS requirements for " unscheduled
meetings" which are essentially the same as for " scheduled meetings."
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meetings would be in Step 6.5.2.3.
i Resolution of the above items remain an open item.
(498/8723-06)
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Procedure OPGP03-ZA0004, PORC Step 2.10 allows a quorum with the Operations QA Manager as l
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" Alternate." This is not allowed by TS 6.5.1.5.
Procedure does not require review of Fire Protection Program as
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specified by TS 6.5.1.6.p.
There is no method of initiating all required PORC reviews.
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secretary can be changed daily per Step 2.9 and is not required at every meeting; so it would be easy to miss a required review. The role and responsibilities of the PORC secretary need to be better defined.
Resolution of the above items will be tracked as Open Item 498/8723-07.
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NSRB Meeting Minutes for Meeting 87-04 The title page was not included as required by Procedure NSRB01,
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Step 6.5.2.
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TMI and GL 83-28 Action Items Followup
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(0 pen) TMI Item I.A.1.3, Shift Manning Applicants shall include, in their administrative procedures, provisions l
governing required shift staffing to assure that qualified personnel are readily available to man the operational shifts in the event of an
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dbnormal or emergency situation. These procedures shall also set forth a policy, the cbjective of which is to operate the plant with the required staff, and develop working schedules such that the use of overtime is avoided, to the extent practicable, for those persons who perfonn safety-related functions (e.g., senior reactor operators, reactor operators, health physicists, auxiliary operators and key maintenance personnel).
Procedure OPGP02-ZA-0060, " Overtime Approval Program," was reviewed and found to be incomplete in that it fails to include shift chemical operations and analysis personnel.
The above discrepancy was discussed with the manager of Management
' Services Business Support Group on April 21, 1987. He stated that a revision will be issued the week of May 1,1987, to incorporate the shift chemical personnel.
A shift supervisor was interviewed on April 21,,1987.
He was able to demonstrate the inplace tracking system and n,ethod of compliance with Procedure OPGP02-ZA-0060.
A leading reactor plant health physics technician was also interviewed.
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He was able to demonstrate acceptable knowledge of the overtime approval programs.
TMI Item 1. A.1.3.1, Limit Overtime, will remain open (498/8723-08) pending
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issuance of the revision to OPGP02-ZA-0060, covering shift. chemistry personnel, discussed above.
Open Item 498/8708-13 is considered closed.
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l (0 pen) TMI Item II.B.4, Training for Mitigating Core Damage i
Applicants are required to develop a training program for the use of installed equipment and systems in cases where the reactor core is severely damaged. They must then implement the training program.
The NRC inspector verified that the health physics personnel training lesson plan, RPT001.61, " Radiological Aspects of a Core Damage Accident,"
as discussed in NRC Inspection Report 50-498/87-08, has been approved.
Open Item 498/8708-16 is considered closed.
The applicant informed the NRC inspector that training of health physics personnel on mitigation of core damage would be completed prior to full.
power.
Item II.B.4 is to remain open for Unit 1 until completion of the above item.
(498/8723-09)
(0 pen) TMI Item II.K.3.9, Proportional Integral Derivative Controller Modification The Westinghouse recommended modification to the proportional integral derivative (PID) controller should be implemented by affected licensees.
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The NRC inspector reviewed Startup Work Request (SWR) 02394 This SWR was performed to remove the jumpers form the derivative circuit portion of the power-operated relief valves (PORV) controller.
The STP Final Safety Analysis Report (FSAR) has been updated to show the derivative controller modification.
No documentation was available stating that the staff has reviewed this modification, therefore, TMI Item II.K.3.9 remains open pending completion of the NRC review.
(0 pen) n termination of Compliance with ATWS Rule 10 CFR 50.62
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e Each applicant must have equipment from sensor output to final actuation device, which is diverse from the reactor trip system, to automatically initiate the auxiliary feedwater system and initiate a turbine trip under conditions indicative of an ATWS.
The Westinghouse Owners Group (WOG) developed three generic designs for l
ATWS Mitigation System Actuation Circuitry (AMSAC) to meet the
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requirements of the rule.
These were submitted to the NRC staff for j
review and acceptance by Letter OG156 dated July 25, 1985, as Topical J
Report WCAP10858.
In a letter to the NRC dated October 8,1985, (ST-HL-AE-1371), the applicant stated that the equipment required by the ATWS rule will be installed and tested in Unit I no later than the first refueling outage j
(this is taken to mean prior to startup follow %g the first refueling outage). The design information required by the rule would be provided to the NRC by June 1986.
In a letter to the NRC dated May 30, 1986, the applicant stated that l
submittal of the AMSAC design information would 3e delayed pending release of the NRC Safety Evaluation Peport (SER) of the WOS report on this issue (WCAP10858). The applicant still committed, however, to have the AMSAC installed and operable by the first refueling outage.
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I In paragraph 15.8.1 of SSER-1, the staff concluded that the generic designs presented in WCAP-10858 are acceptable.
I STP FSAR, Section 7.8 (Amendment 57) describes the AMSAC design chosen by i
STP (one of the three generic designs of WCAP-10858). The ANSAC will.
initiate the required actions when flow drops below a predetermined l
setpoint, and remains below this value for a preset time delay, in three
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of the four main feedwater lir.es, when power (as sensed by turbine impulse l
pressure) is above 70 percent (C20 permissive).
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In Supplemental SER (SSER) 2, paragraph 15.8.1, the staff recognized that the applicant submitted the required AMSAC design information for staff review on October 20, 1986.
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1 The STP AMSAC design continues to be reviewed for acceptance by the staff.
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Their conclusion will be provided in a future supplement to the SER.
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This item will remain open and be traced as (498/8723-10) pending completion of the following:
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i Completion of NRC staff review of the STP plant-specific AMSAC design.
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Proper procurement, installation, and satisfactory testing of the AMSAC
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system by HL&P.
i Incorporation of the installed AMSAC system into operating procedures,
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training programs, and maintenance and surveillance programs by HL&P.
(Closed)TMI Item II.D.3, Direct Indication of Relief and Safety Valve Position
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Reactor coolant system relief valves shall be provided with a positive indication in the control room derived from a-reliable valve position detection device or a reliable indication of flow in the discharge pipe.
The staff concluded in the STP SER, that the PORV and safety valve
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position indication meets the guidelines of NUREG0737 and is acceptable.
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The NRC inspector reviewed Drawings D453, Revision 5, TD1768, Revision 8, I
and 9-E-MS22-01, Revision 2, associated with safety relief valve positions indication. The results of Test Item TRC0Z032771 of Procedure SMP08-ZIO061, l
Revision 4, " Functional Check of Instruments and Components," was also
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reviewed.
l The NRC inspector reviewed Drawing 9-G-RC13-01 associated with the PORV
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position indication. The applicable portion of Procedure I-RC-P02,
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Revision 1, " Hot Functional Test," was also reviewed.
The above documentation is considered acceptable and TMI Item II.D.3 is
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considered closed.
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(Closed) TMI Item II.E.1.1, Auxiliary Feedwater System (AFWS) Evaluation The Office of Nuclear Reactor Regulation has required reevaluation of the AFWS for all pressurized water reactor (PWR) operating plant licensees and operating licensee applications. This action includes:
i Perform a simplified AFWS reliability analysis that uses event-tree and
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fault-tree logic techniques to determine the pot?ntial for AFWS failure under various loss-of-main-feedwater-transient conditions.
Particular emphasis is given to determining potential failures that could result from human errors, common causes, single-point vulnerabilities, and test and i
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Perform a deterministic review of the AFWS using the acceptance criteria
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of Standard Review Plan Section 10.4.9 and associated Branch Technical Position ASB 10-1 as principal guidance.
Reevaluate the AFWS flowrate design bases and criteria.
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STP SER (NUREG0781) concluded in Section 10.4.9 that the design-basis auxiliary feedwater (AFW) system flow requirements is acceptable.
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STP SSER 1 concludes that the applicant has complied with the' guidelines of NUREG-0737 Action Item II.E.1.1 concerning the AFWS reliability
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l analysis, and that AFWS unavailability per demand for the South Texas Project is compatible with staff guidance in SER Section 10.4.9.
TMI Item II.E.1.1 is considered closed.
(Closed) TMI Item II.G.1, Emergency Power for Pressurizer Equipment Power supply for pressurizer relief and block valve and pressurizer level indicators shall have the following characteristics:
1 Motive and contral components of the PORV shall be capable of being
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supplied from either the offsite power source or the emergency power source when the offsite power is not available.
Motive and control components associated with the PORV block valves shall
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be capable of being supplied form either the offsite power source or the emergency power source when the offsite power is not available.
Motive and control power connection to the emergency buses for the PORVs
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and their associeted block valves shall be through devices that have been l
qualified in accordance with safety-related requirements.
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The pressurizer level indication instrument channels shall be powered from
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the vital instrument busses. The busses shall be capable of being supplied from either the offsite or the emergency power source.
The NRC inspector reviewed the applicable STP electrical dic cams te u
verify the following:
The PORVs and PORV block valves are Class 1E qualified and capabie of
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being supplied power from either the offsite power source or the emergency onsite power source.
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The pressurizer level indication instrumentation and their associated
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busses are Class 1E qualified and powered from Class 1E vital instrument buses.
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The PORV and PORV block valve motive and control power are supplied from
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different emergency power buses.
Therefore, the installation of equipment meets the licensee commitments and NRC requirements.
l TMI Item II.G 1 is considered closed.
(Closed) TMI Item II.K.3.10,' Proposed Anticipatory Trips Modification i
The anticipatory trip modification proposed by some licensees to confine the range of use to high-power levels should not be made until it has been shown on a plant-by-plant basis that the probability of a small-break-loss-of-coolant accident resulting from a stuck open, power operated relief valve is substantially unaffected by the modification.
STP SER concludes in Section 7.2.2.4 that the STP design, which includes reactor trip on a turbine trip above 50 percent of rated an anticipatory (P9 interlock), is in compliance with the TMI Action Plan thermal power Guidelines.
Drawing No. 52-10-92-4211, " Reactor Trip Signals Logic Diagram," was reviewed and found to include logic which needs the automatic reactor trip circuitry whenever turbine electric hydraulic control (EHC) fluid pressure drops below a predetermined setpoint or all four turbine throttle stop valves are closed and reactor power is above a predetermined setpoint (P9)
as sensed by Nuclear Instrument Channels N41, N42, N43, and N44 Preoperational testing has been completed.
TMI Item II.K.3.10 is considered closed.
(Closed) TMI Item II.K.3.25', Effect of Loss of Alternating Current Power on Pump Seals Applicants should determine on a plant specific basis, by analysis or experiment, the consequences of a loss of cooling water to the reactor recirculation pump seal coolers. The pump seals should be designed to withstand a complete loss of alternating current (AC) power for at least i
2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Adequacy of the seal design should be demonstrated.
In the STP SER, the staff concludes that the design is acceptable.
l The NRC inspector reviewed the results of the' A, B, and C engineered
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safety feature (ESF) diesel sequencer functional test (SMP03-SF0001, SMP03-SF0002, and SMP05-SF0003) and verified that the seauencer delay i
times for the CCW and charging pumps were within the acceptance criteria.
TMI Item II.K.3.25 is considered closed.
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(Closed) 498/8708-23, Salem ATWS Item GL 83-28/4.1, Reactor Trip System Reliability (Vendor-Related Modifications)
All vendor recommended reactor trip) breaker modifications shall be (1 each modification has, in fact, been i
reviewedtoverify)thateither:
implemented,or(2 a written evaluation of the technical. reasons for not
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implementing a modification exists.
The NRC inspector reviewed HL&P memorandum dated April 2, 1987, from R. A. Johnson to J. C. Leonard.
The memorandum discusses completion of the reactor trip breaker inspections of Westinghouse Technical Bulletin NSP-TB84-02 as discussed in NRC Inspection Report 50-498/87-08 and states that the deficiencies described in the bulletin do not exist.
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Salem ATWS Item GL 83-28/4.1 is considered closed.
(Closed) 498/8708-11 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 the control room access and establish clear lines of authority and responsibilities in the control room during emergencies.
STP SER (NUREG-0782) Section 13.5.1.2 states that the applicant has committed to limit access to the control room to meet the NUREG-0694 Item I.C.4.
STP Procedure OPGP03-ZQ0004, Revision 1, " Plant Conduct of Operations,"
Section 4.5 was reviewed with the following findings:
Section 4.5.1 and 4.5.7 established a clear line of authority,
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responsibility, and succession in the control room during normal and emergency situations.
Section 4.5.3 limits access to the control room to those persons who have
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official business or are required to be in the control room (e.g., those required for direct operations of the plant, technical advisors, and e
assigned resident NRC personnel).
Section 4.5.4 allows the shift supervisor or unit supervisor to direct
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nonessential personnel to leave the control room as deemed necessary, i
Nuclear Group Policies 945 covers the requirement that the shift supervisor shall be a licensed senior reactor operator (SR0).
The STP Emergency Plan, Section C, specifies the shift supervisor actions in the event of an emergency.
i Based on the above, TMI Item I.C.4, control room access is considered i
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(Closed) TMI Item II.K.3.5', Automatic Trip of Reactor Coolant Pumps During Loss of Coolant Accident (LOCA)
GL 85-12. " Implementation of TMI Action Item II.K.3.5, Automatic Trip of Reactor Coolant Pumps," required owners of Westinghouse Nuclear Steam Supply Systems (NSSS) to evaluate their plants with respect to reactor coolant pump (RCP) trips. The objective was to demonstrate that their proposed RCP trip setpoints ensure pump trip for small break LOCAs and, in addition, to provide reasonable assurance that RCPs are not tripped unnecessarily during non-LOCA events.
Several plant-specific items were identified which were to be considered by applicants, including the selected RCP trip parameter, instrumentation redundancy, instrumentation uncertainty, potential RCP and RCP-associated problems, operator training, and operating procedures.
l Letter ST-HL-AE-1753 from HL&P to the NRC, dated September 30, 1986, was reviewed.
It was determined by this review that STP responded to several of the above listed considerations as follows:
Selected RCP Trip Parameter - STP stated that Reactor Coolant
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Systems (RCS) wide range pressure will be used to determine whether or not RCP trip is required.
Instrumentation Redundancy - Three separate channels are available, two l
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with a range of 0 to 3500 psig and the third with a range of 0 to 3000 psig.
Instrumentation Uncertainty - STP stated that the wide range pressure
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instrument channel uncertainty corresponds to 81.2 psig, considering all error components from the sensor through the RCS wide range pressure display.
It was further stated that since these instrument transmitters are located outside the containment, they are not subject to adverse containment conditions.
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'lation Uncertainty - STP concluded that the calculations uncertainty
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is -5 to +25 psig for the RCS pressure RCP trip setpoint.
Letter ST-HL-AE-1433 from HL&P to the NRC, dated November 6,1985, was reviewed.
This review showed that STP responded to the other listed considerations as follows:
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RCP and RCP-Associated Problems - STP stated that containment
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isolation (CI) will not cause RCP problems if it occurs for non-LOCA transients and accidents, because the CI signal by itself does not terminate either of the two water supplies to the RCP seals.
Furthermore, STP stated that RCP operation can continue, without seal damage, following CI.
STP also identified components required to trip the RCPs, and concluded that the RCP trip function will be operable during any postulated LOCA.
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Operator Training - STP states that the. training program will include
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instruction to operators in the use. of procedures concerning RCP trip in the esent of a small break LOCA.
Operating Procedures - STP provided a list of 22 procedures which include
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RCP trip related operations, and stated that~ they are being developed using the guidelines established in the Westinghouse Owners Group Emergency Response Guidelines, Revision 1.
Letter ST-H1 -AE-1825 from HL&P to the NRC, dated December 5.1986, was reviewed.
as it STP stated that the RCP trip setpoint.. based on the RCS pressure criteria selected, is 1477 psig for both normal and adverse containment conditions.
d The following procedures were reviewed, and found to have RCP trip-related operations in them, either as specific procedural steps, or as part of a conditional information page, which must be monitored throughout
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performance of the procedure:
1 POP 05-E0-E000 Reactor Trip or Safety Injection
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1 POP 05-E0-ES01 Reactor Trip Response
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1 P0P05-E0-ES02 Natural Circulation Cooldown
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1 P0P05-E0-E010 Loss of Reactor or Secondary Coolant
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In each case, the operator is instructed to stop the RCPs if at least one l
high head safety injection pump is running and RCS pressure is less than i
1477 psig.
TMI Item II.K.3.5 is considered closed.
(Closed) TMI Item II.K.1.5, Review of Safety-Related Valve Position Applicants are required to review all safety-related valve positions, positioning requirements, and positive controls to assure that valves i
remain positioned (open or closed) in a manner to assure the proper operation of ESF. Also review of related procedures (e.g., maintenance, surveillance, etc.) to ensure such valves are returned to their correct positions following their ESF positions is required.
A review was conducted of Procedure OPGP03-Z0004, " Plant Conduct of l
Operations," Revision 1, dated April 9,1987.
Section 4.10 of this procedure, " Control of Systems and Equipment," addresses the use of checklists for aligning system valves, independent verification for
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safety-related components, and supervisory input for component alignment i
l in cases where a written checklist or procedure does not exist. Also addressed in Section 4.10.is the control of designated components, which are required to remain in preset configuration, through the use of locking devices.
The NRC inspector interviewed the Shift Supervisor of Operations Division, Crew E, during which the following information was provided:
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STP Security has responsibility and control for all facility locks and
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l locking devices.
A craftsman or other person desiring to perform j
maintenance on, or otherwise manipulate a normally locked valve must l
produce an equipment clearance, approved by the Shif t Su >ervisor or Unit Supervisor to Security, before being allowed to obtain tie required key.
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Operating personnel are issued a " master" key, which fits the series of l
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locks installed on ESF valves, for their use in required plant operations.
At tb time of shift turnover, the oncoming operating crew walks down the i
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boards to ensure no ESF components are wrongly positioned. The oncoming l
shift also reviews all required operating logs since last being on duty,
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or for a predetermined period of time. A review is made of the equipment
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clearance bock, and the status of any ongoing surveillance, preventive l
maintenance, and corrective maintenance activities is ascertained.
Key ESF components, including valves, are monitored by the ESF status
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monitoring panel in the control room.
If a monitored ESF valve's position changes to an unacceptable position, this occurrence is attended by an alarm (ERFDADS Trouble Alarm) on the Supervisor's Panel.
Equipment Clearance Procedure OPGP03-200001, Revision 8, dated February 3,
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1987, was discussed and reviewed.
It was found to contain appropriate and adequate direction to ensure safety-related ESF valves are correctly positioned and those positions independently verified after release of a clearance. Also included in Procedure OPGP03-Z0-0001 is a list of systems requiring independent verification upon release, so there can be no
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question as to this requirement. This list contains all ESF systems.
The
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issuing authority, (person directing the tagging operations), refers to
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this list prior to issuance, and then stamps the clearance " Independent Verification Required" as appropriate.
Normal surveillance activities and plant operations which involve
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repositioning ESF valves are governed by their own specific procedural steps and/or valve alignment checklists which, if properly followed,
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ensure that at the conclusion of the operation or surveillance test, all ESF valves are again positioned such that they can perform their intended ESF function.
TMI Item II.K.1.5 is considered closed.
l 7.
Preoperational Test Procedure Review The following preoperational test procedures were reviewed for compliance i
with the FSAR and Regulatory Guide 1.68 and found acceptable:
1-SF-P-01 Safeguards Systems Response No Blackout
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1-SF-P-02 Safeguards Systems Response With Blackout
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1-SF-P-03 Safeguard Test Cabinet Train A
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1-SF-P-04 Safeguard Test Cabinet Train B
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1-SF-P-05 Safeguard Test Cabinet Train C
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1-SP-P-03 Reactor Trip and ESF Time Response J
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Within the scope of the inspection, no violations or deviations were found.
8.
Preoperational Test Witnessing
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l The NRC inspectors witnessed the performance of various Integrated
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Safeguards System Response Tests (i.e., Train A with blackout, Train C no blackout) and found the tests to be properly executed with the use of I
I procedures. However, due to unacceptable test results, the retesting will be scheduled for June 1987.
No violations or deviations were found.
9.
IE Bulletins (IEB), IE Circulars (IEC), and IE Notices (IEN)
l (0 pen) IEB 85-01, " Steam Binding of AFW Pumps" The applicant satisfied the requirement for monitoring conditions at the AFW pump discharge by providing continuous monitoring of discharge pipe temperatu re. The temperature monitor will alarm the Emergency Response l
Facility Data Acquisition and Display System (ERFDADS) at 200 F.
Plant Procedure 1PMP04-AF0001, Revision 0, " Auxiliary Feedwater Discharge Header High Temperature," was developed to provide guidance for actions to be taken in response to an ERFDADS alarm on the AFW discharge header temperature monitor.
The NRC inspector verified the procedure provided a
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method for determining the extent of the AFW piping high temperature i
condition as well as verifying the referenced procedures provided for venting the AFW pump and for determining pump operability. This bulletin is open until approval of 1PMP04-AF-0001, Revision 0, and verification of the alarm function in the ERFDADS computer.
(Closed) IEB 85-02, "Undervoltage Trip Attachments of Westinghouse DB50 Type Reactor Trip Breakers" The applicant's reactor trip breakers (RTB) are Westinghouse DS416 which required actuation of the RTB shunt trip coil on all automatic reactor trip signals.
South Texas Project Electric Generating Station (STPEGS)
procedure OPMP05-NA0008, Westinghouse 480 Volt Breaker Test," required performance of specific tests which included performance of " Shunt Trip Attachment Test" and "Undervoltage Trip Attachment Force and Load Check."
This bulletin is closed.
(Closed) IEB 82-02, " Degradation of Threaded Fasteners in the Reactor Coolant Pressure Boundary of-PWR Plants" The NRC inspector reviewed the below listed procedures and determined that in the areas reviewed the lubricant used and torque values / bolt tensioning method utilized in the maintenance procedures met the requirements in the respective vendor manual. The procedures reviewed were:
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1 OPMP04-CS0001, Revision 1, " Containment Spray Pump Maintenance"
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OPMP04-MS0001, Revision 1, " Main Steam Safety Valve Removal and i
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Reinstallation"
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OPMP04-RC0003, Revision 1, " Reactor Coolant Pump Maintenance"
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OPMP04-RC0004, Revision 0, " Steam Generator Primary Manway Cover
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Removal and Reinstallation" OPMP04-SI0001, Revision 1, " Low Head Safety Injection Pump
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Maintenance" OPMP04-SI0002, Revision 1, "High Head Safety Injection Pump
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Maintenance" The NRC inspector reviewed Interdepartmental Procedure IP3.2Q, Revision 2,
" Expendable Material Control Program," which established a program to control expendable material used so as to prevent adverse effects on permanent plant components and system.
Procedure IP.3.2Q also established the STPEGS Expendable Materials Manual. The Expendable Materials Manual provided information relative to material specification use and restrictions.
The other interdepartmental procedure reviewed was IPS.1Q,
" Procurement of Items," Revision 1, which provided guidance on departmental responsibilities in the procurement of items at STP. This procedure met the requirements, as applicable, of ANSI.N18.7-1976, Section 5.2.13,
" Procurement and Material Control." Procedure OPGP03-ZP0001, " Nuclear Plant Operations Procurement of items and Services," establishes the method for procurement of items and services by and for the Nuclear Plant OperationsDepartment(NP0D).
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The NRC inspector reviewed three purchase orders (P0) related to fastener lubricants, one of which included the procurement recommendation form.
Each P0 met the required documentation and specifications as stated on their Expendable Materials Specification. This bulletin is closed.
(CLOSED) IE Information Notice (IEN)85-093, " Westinghouse Type DS Circuit Breakers, Potential Failure of Electric Closing Feature Because of Breaker Spring Release Latch Lever."
The applicant stated that, "the capability to manually operate breakers is common operator knowledge...." This situation was evaluated for safety. significance by Bechtel Energy Corporation (BEC) and determined to have no safety significance.
STP Procedure OPMP05-NA0008, Revision 1, " Westinghouse 480 Volt. Breaker Test," dated February 24, 1987, incorporates inspection of the spring
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release latch lever during routine maintenance. The safety evaluation conducted by BEC and the inspection of the lever satisfies the IEN.
This
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notice is closed.
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(Closed) IEC 77-15 Degradation of Fuel Oil Flow to the Emergency Diesel Generator l
This issue concerned inadequate fuel flow to Emergency Diesel Generators partly because of a clogged system strainers. Station personnel were not
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aware the strainer existed due to inadequate documentation. The STP
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design minimizes the potential for flow degradation and the as-built plant l
and design drawings have been reconciled.
This circular is closed.
10. Site Tours The NRC inspector conducted site tours both independently and accompanied by licensee personnel.
These tours were made primarily to assess the condition of inplace safety-related equipment, plant status, and to observe ongoing preoperational testing and work activities.
The areas toured included:
Unit 1 - mechanical and electrical auxiliary building (MEAB), reactor
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containment building (RCB), FHB, emergency diesel
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building (EDGB), and the isolation valve cubicle (generator
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IVC).
Unit 2 - MEAB, RCB, FHB, EDGB, and IVC.
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11.
Exit Interview The NRC senior resident inspector met with licensee representatives (denoted in paragraph 1) on May 4,1987, and summarized the scope and j
findings of the inspection.
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