IR 05000395/1982031
| ML20055A171 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 05/18/1982 |
| From: | Brownlee V, Skolds J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20055A165 | List: |
| References | |
| TASK-1.A.1.1, TASK-1.A.1.3, TASK-2.B.3, TASK-2.B.4, TASK-2.F.1, TASK-2.F.2, TASK-TM 50-395-82-31, NUDOCS 8207150544 | |
| Download: ML20055A171 (9) | |
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'q UNITED STATES
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NUCLEAR REGULATORY COMMISSION o
$
I REGION ll o,
101 MARIETTA ST., N.W., SUITE 3100
%
ATLANTA, GEORGIA 30303 Report No. 50-395/82-31 Licensee:
South Carolina Electric and Gas Company Columbia, SC 29218 Facility Name:
V. C. Summer Docket No. 50-395 License No. CPPR-94 Inspection at V. C. Summer Inspector: /
Y S//7df'Z
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pt.Skolds Vf/ 'g D~ ate Signed Approved by:,
dugn 7 d 5 -//$ /f L V
C. Brownlqf, 3ection Jfdef, Division of Date ' Signed roject and Resident #rograms
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SUMMARY Inspection on May 1-31, 1982 Areas Inspected This routine, unannounced inspection involved 144 inspector-hours on site in the areas of Open Item review, TMI Action Plan Item Review, 50.55(e) Review, Emergency Exercise Review, Independent Inspection Effort and Preoperational Test
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Results Review.-
Results Of the six areas inspected, no items of noncompliance or deviations were identified.
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8207150544 820618 PDR ADOCK 05000 O
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i DETAILS
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1.
Persons Contacted i
Licensee Employees
- 0. W. Dixon, Vice President Nuclear Operations
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- 0. S. Bradham, Station Manager J. G. Connelly, Deputy Plant Manager
B. G. Croley, Assistant Manager, Technical Support
- S. J. Smith, Assistant Manager, Maintenance L. F. Storz, Assistant Manager, Operations
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- A. R. Koon, Technical Services Coordinator S. S. Howze, Nuclear Licensing
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- W. A. Williams, Jr., General Manager, Nuclear Operations
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- H. C. Fields, Technical Services Engineer
- P. V. Fant, Director, Station QC
- C. Baily, Startup
- A. M. Paglia, Licensing Engineer i
Other licensee employees contacted included technicans, operators and
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office personnel.
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- Attended exit interview
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i 2.
Exit Interview i
i The inspection scope and findings were summarized on June 4, 1982, with j
those persons indicated in paragraph 1 above. The inspector attended the l
exit interview of J. Lenahan on May 7, 1982 and D. Andrews on May 6, 1982.
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3.
Licensee Action on Previous Inspection Findings Not inspected.
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Unresolved Items i
Unresolved items were not identified during this inspection.
5.
Emergency Exercise Review
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The inspector reviewed the emergency exercise conducted on May 5, 1982. The j
inspector assisted an inspection team from the NRC Regional Office. The
details of the inspection are included in Inspection Report 82-33.
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6.
TMI Action Item Followup
(Closed)
I.C.5 Feedback of Operating Experience to Plant Scaff.
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This item deals with the establishment of administrative procedures to I
assure that operating information, pertinent to plant safety, originating both within and outside the utility organization is continually supplied to operators and other personnel and is incorporated into training and
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retraining programs.
These procedures shall:
1.
Clearly identify organizational responsibilities for review of
operating experience, the feedback of pertinent information to operators and other personnel, and the incorporation of such infor-
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mation into training and retraining programs; 2.
Identify the administrative and technical review steps necessary in i
translating recommendations by the operating experience assessment group into plant actions (e.g., changes to procedures; operating
orders);
3.
Identify the recipients of various categories of information from
operating experience (i.e., supervisory personnel, shift technical
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advisors, operators maintenance personnel, health physics technicians)
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or otherwise provide means through which such information can be readily related to the job functions of the recipients;
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4.
Provide means to assure that affected personnel become aware of. and understand information of sufficient importance that should not wait for emphasis through routine training and retraining programs;
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Assure that plant personnel do not routinely receive extraneous and unimportant information on operating experience in such volume that it
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would obscure priority information or otherwise detract from overall job performance and proficiency; i
6.
Provide suitable checks to assure that conflicting or contradictory l
information is not conveyed to operators and other personnel until l
resolution is reached; and,
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Provide periodic internal audit to assure that the feedback program (
functions effectively at all levels.
The applicant's commitments concerning this item are described in the V. C.
Summer Safety Evaluation Report (SER) and Supplement No.1 to the SER.
Station Administrative Procedure (SAP)-147 " Operating Experience Reports"
Revision 0 was written to implement the commitments in the SER. The
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inspector reviewed SAP-147 and found the procedure acceptable with the
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following exception:
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When training is a recommended action after reviewing the operating experience report, the procedure allows the action to be reported I
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complete once the training is scheduled.
If no one had received any
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training the item could be reported complete. This did not seem to
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assure that affected personnel received the required training. The applicant agreed to change SAP-147 to indicate that the action would be reported complete only after the training was complete.
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(0 pen)
II.B.3.
Post accident Sampling Capability.
This item involved the '
requirement to install a post accident sampling system. The requirements of the system capabilities are described in NUREG 0737 and the V. C. Summer SER. The V. C. Summer SER, page 22-71 presently states that on-line analysis will be provided for chloride, hydrogen, pH, boron anti conductivity. The inspector reviewed the system description of the post accident sampling system and the presently installed system and could not find the on-line analysis capability described above. The SER indicates that a post-implementation review will be completed by NRR and reported in a supplement to the SER.
The inspector reviewed the procedures concerning the post accident sampling system.
Specifically, System Operation Procedure-107 and Chemistry Operating Procedure CHP-903 were reviewed. CHP-903 had a number of valves missing from the valve lineup and a number of valves listed on the valve lineup incorrectly. The procedures were reviewed specifically for how the procedures interfaced in setting the system up for normal operation. Until the NRR post-implementation review is complete and CHP-903 is revised, this item will remain open.
(Closed) II.F.1.2.E.
Reactor Building Water Level. This item deals with the requirements to have a continuous indication of containment water level
l in the control room. Also, a narrow range instrument shall be provided to l
cover the range from the bottom to the top of the containment sump. A wide range instrument shall be provided and shall cover the range from the bottom j-of the containment to the level equivalent to 600,000'. gallons. The
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inspector verified the following:
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Narrow range and wide range instruments are installed.
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Indication of wide range narrow range level exists in the control room on vertical indicator and strip charts.
(Closed) II.F.1.2.F.
Containment Hydrogen Indication. This item deals Yith a requirement to have a continuous indication of hydrogen concentration in the containment atmosphere provided in the control room. Measurement capability is required over the range 0 to 10 percent. The inspector verified that the hydrogen monitors are installed and that indication exists in the control room over the range of 0 to 10 percent.
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(0 pen) II.B.4 Training for Mitigating Core Damage.
This item, deals with the requirement to develop a training program to instruct all operating l
personnel in the use of installed systems, including systems that are not engineered safety features, and instrumentation to monitor and control
accidents in which the core may be severely damaged. The requirement for I
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fuel load is that the program be established. The inspector reviewed the training program for operators, shift technical advisors, I&C technicians, chemistry technicians and health physics technicians. The training program covers the following topics:
Incore Inst >umentation, Excore Nuclear Instru-mentation, Vital Instrumentation, Primary Chemistry, Radiation Monitoring and Gas Generation. Until all personnel required to receive this training receive such training, this item will remain open. This is a full power restraint.
(Closed) I. A.1.1 Shift Technical Adviscr. This item deals with the requirement to have a shift technical advisor (STA), available to the shift supervisor, present on all shifts (Modes 1-4) and available to the control
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room within ten minutes. The V. C. Summer Technical Specifications presently require one STA be present on shift (Modes 1-4).
At present, the
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applicant has hired and trained enough STAS' to cover Modes 1'4'.
The duties and the responsibilities of the STA's are described in Administrative
, Procedure (AP)-519, " Duties and Responsibilities of the Shift Technical
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Advisors". The V. C. Summer SER, Supplement No.1 indicates that all STA's will hold a bachelor's degree in engineering or related sciences. The inspector verified that all STA resumes indicate a bachelor's degree in engineering or related scie-he V. C. Summer SER, Supplement No.1 indicates that the STAS wi'
_,ve training in the response'and analysis of the plant for transients and accidents, reactor theory and plant specific systems, mitigating core damage, operating and administrative procedures, emergency plan drills, technical support center, problem solving an decision making as well as 50 classroom hours and 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> of simulator training.
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The inspector verified that the training records for all STAS., reflect that
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training in all the above mentioned areas has been completed.
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r (Closed) 1.A.1.3 Shift Manning. This item involves the requirements fer minimum shift crew composition and working hours for personnel performing safety-related functions. Both of these items are addressed in the Technical Specifications. Therefore the requirement to include these i.tems in administrative procedures is considered to be redundant to the Technical Specification requirement. Therefore this item is closed out based on the Technical Specification requirements.
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(0 pen)II.R.2InadequateCoreCooling. This item deals with the requirement to have instrumentation in order to detect inadequate core cooling.
In a
. letter to the NRC dated March 16, 1982 the applicant requested reliefzon the fuel load requirement of this item. This will be addressed in a Supplement to the SER. The inspector reviewed the instrumentation installed to verifyf it conforms with the information in the SER, FSAR and other correspondence between the applicant and the NRC.
Findings were acceptable with the following exception:
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1.
In a letter dated June 8, 1981 to the NRC, the applicant stated that the Core Subcooling Monitor consisted of a 15. lamp display. Also, the monitor divided the core into five regions, four peripheral and one central, with inputs from two thermocouple used 40 generate the disp, lay
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for each region. Th'einspcctorfoundthatthemonitordividesthec6rI'
up into four quardrants, not five. Also, because of th four quadriats the monitor consists of a 12 lamp display, not 15.
Since this itep *.-
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will be addressed in a Supplement to the SER, this' item will remain. /
open.
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50.55(c) Report /Part 21 Report Followup
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(Closed) 81-28-01. Diesel Generator Fuel Oil Lini.L Inalefr$rda+cd
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September 10, 1981 the applicant reported a proble.m with the. fuel oil return line developing a leak. This was due to the fact that the, fuel oil lile rubbed against a junction box during operation of the rs,it.
The cause of.
the vibration was due to imoroperly positioned pipe s'qports. 'in a lettr the applicant stated that the manufacvgrer sopplied' t -
dated March 29, 1982 drawings for the diesel that did not specifically address' pipe suoport locations on the fuel oil return lines.
The supports were installed to/the appro-ximate original locations and drawings were submitted to the malufacturer for verifications. The manufacturer then specified a spac h)' criteria for the supports. All supports have been installed in.tccordance with the manufacturer's spacing criteria and an as-built draw'ing submitted to the manufacturer.
The inspector reviewed thi( drawing qnd cc7;pareti the drawing to the as-built conditions. Ji/idings,were ' acceptable, i
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(Closed) 81-23-01.
Improper termination of solenoid valvi.
di a' letter
dated August 25, 1981 the applicant reported that a scienoid vqhe that controlled air to a containment isolation ulve w s fo'und to b. improperly (
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terminated.
The termination was not made in accordance with approved
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procedures and could have failed.
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The inspector reviewed the following documerits concerning the problem as
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well as the corrective action:
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Startup Field Report (SFR) - 4702
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Engineering Change Notice (ECN) - 1554 FM
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SFR - 4025 SFR - 3546 QA Type I Surveillance Report - 6811
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MWR - 54039 MWR - 54038 Findings were acceptable.
(0 pen)(80-05-02) Rod Drop Analysis.
In a letter dated March 6, 1980 the applicant reported a significant deficiency regarding rod drop analysis.
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The details of the deficiency and an interim solution was presented in a letter to SCE&G from Westinghouse dated January 16, 1980. The interim
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solution consists of placing the control rods in MANUAL when reactor power is greater than 90% and Bank D is not withdrawn greater than 215 steps.
If Bank D is withdrawn greater than 215' steps the rods may remain in automatic control.
For power levels between 0 and 904 there is no change to the mode of oneration. The inspector verified that this interim solution is reflected in the General Operating Procedures. A final solution to this problem will be developed at a later date.
Until a final solution is developed and a final report issued, this item
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will remain open.
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8.
Preoperational Test Results Review The inspector reviewed the results of the following preoperational tests:
DG-5 Diesel Generator A Test
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DG-7 Diesel Generator B Test-
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CC-1 Component Cooling Water SG-1 Safeguards Test with Blackout
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.SI-8 ECCS Check Valve Backleakage Test
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CR-3 Red Position Indication Initial Energization
'5 CR-10 CRDM Stepping Checks CS-2 Charging Pump Flow Test CS-4 Charging / Seal Injection and Letdown CS-9 BTR3 Demineralizer Thermal Performance The results were reviewed to verify the test was performed in accordance with the procedures and commitments in the FSAR. The results were also reviewed to ensure the indicated results were within the acceptance
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criteria.
Findings were acceptable.
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9.
Open Item Followup i
l (0 pen)(82-31-01) Earthquake Instrumentation.
In a letter dated May 25,
1982 the applicant indicated that a triaxial peak accelograph will be moved from the bottom of the reactor vessel to the surge line. Until the i
accelograph is installed, this item will remain open and is considered a fuel load restraint.
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(Closed)(82-29-05) Preoperational Test SG-7.
This item dealt with the fact that the SG-7 test package contained no objective evidence to prove that Acceptance criteria 8.4 was met. This acceptance criteria stated that the maximum safeguard loading does not exceed the 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> rating of 4250KW (4722 KVA). The inspector reviewed TCR-4 to SG-7 which listed the measured KVA loads for the major loads plus various motor control centers. Neither diesel generator maximum load exceeded the design KVA. Therefore Acceptance Criteria 8.4 was met.
(Closed) (80-13-01) Technical Specification Review. This item involved twenty comments on the proposed technical specifications. All comments have been resolved.
(0 pen)(82-31-02) Violation of Separation Criteria.
In a letter dated April 28, 1982 the applicant reported that the architech-engineer's specification of separation criteria for non-safety-related space heater circuits serving safety-related motors and safety-related switchgear. This item will be reviewed at a later date.
(0 pen) (82-31-03) Wide Range Reactor Coolant System Pressure Transmitters.
On May 12, 1982 the applicant reported that the post accident accuracy of the wide range reactor coolant system pressure transmitters was not correct.
The May 12, 1982 letter served as an interim report.
This item will be reviewed at a later dated when the final report is issued and reviewed.
(Closed) (81-31-04) RC-1 Hot Functional Testing. This item dealt with five comments concerning the performance of the cooldown from outside the control room. All comments were resolved.
(Closed)(82-26-01) Seismi-Support of RCP Oil Collection System.
This item involved the lack of data available on site to verify that the RCP Oil Collection System is seismically qualified.
In a letter dated April 12, 1982 from Gilbert Associates, Inc to SCE8G it was indicated that the RCP Oil j
Collection System is seismically qualified.
l (Closed)(81-08-07) Preoperation Test SP-2 Results.
This item involved the r
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calculation of piping friction loss for conditions different than the calculated loss. Specifically, the friction loss acceptance criteria for
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l the RB Spray Pump sumps were for conditions of 3000 gpm. The test, SP-2, measured the friction loss at 2950 gpm and 2975 gpm.
Further calculations estimated the friction loss at 3000 gpm. The estimations were within the acceptance criteria.
(Closed)(81-08-05) Preoperational Test SP-3 Results.
This item involved i
the failure to conduct a post test calibration check on level transmitters used to gather data for the acceptance criteria of the test. The applicant issued data from calibrated flow transmitters and the recorded times of stopping and starting various pumps to verify that the indicated level was in agreement with the actual level. Also, the curves in Section 6.2 of the FSAR were explained using the data obtained during the test.
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(Closed)(81-31-02) Coments on G0P-8. GOP-8 has been revised and reviewed.
All coments have been resolved.
(Closed)(81-23-09) This item dealt with a commitment to establish an emergency procedure on the administrative control and proper activation of the emergency siren system.
Emergency Plan Procedure (EPP)-21, " Actuation of EWSS, Early Warning Siren System" has been issued. The inspector reviewed the procedure and found it acceptable.
(Closed)(82-28-04) This item dealt with comments on the Integrated Leak Rate Test Report. All comments have been resolved.
10.
Independent Inspection Effort 1.
In reviewing TMI Action Plan item II.F.1, Attachment 6, System Operating Procedure (50P)-122 Post Accident Hydrogen Removal System was reviewed. The procedure was reviewed for technical adequacy and to verify the system is operated in accordance with the FSAR and other commitments made by the applicant.
Findings were acceptable with the following exceptions:
a.
Attachment 11 lists breakers to be positioned to line up the Post Accident Hydrogen Removal System but does not list in what position to put the breakers.
b.
The method described in the procedure for post accident hydrogen removal using the backup surge / reactor building pressure equali-zation system is not the same method described in the FSAR, Section 6.2.
(page6.2-98).
2.
The inspector discussed the applicability of the Technical Specifi-cations (TS) once the operating license is issued. The inspector informed the applicant that the Technical Specifications, "At all times' or "All modes", are applicable once the plant enters Mode 6.
Other TS that are not made dependent will be applicable once the license is issued.
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