ML19263D885
| ML19263D885 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 03/21/1979 |
| From: | Tibbitts D Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7904170016 | |
| Download: ML19263D885 (16) | |
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NUCLEAR REGULATORY cC"EsslC'.
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v;p 31.g79 Docket No.
50-395 APPLICANTS: South Carolina Electric ano Gas Company and South Carolina Public Service Authority FACILITY:
Virgil C. Sumer Nuclear Station, Unit No.1
SUBJECT:
SUMMARY
OF MEETING HELD ON FEBRUARY 6 AND 7,1979 WITH CAROLINA ELECTRIC AND GAS COMPANY On February 6 and 7,1979, we met with repre.sentatives of South Carolina Electric and as Ccapany (SCE&G), Gilbert Associates Inc. (GAI) and Westinghouse Electric Corporation (Westinghouse). The purpose of the meeting was to discuss the responses to many first and second round requests for information on the Final Safety Analysis Report (FSAR) for the Virgil C. Summer Nuclear Station, Unit 1.
The review areas that were covered included reactor systems, transient analysis, mechanical engineering, materials engineering and initial tests and operation. The meeting was held in Bethesda, Maryland and the persons attending all or part of the meeting are listed in Enclosure No.1.
The discussion of material under reactor systems and initial startup and tests involved many items. The sumaries of these discussions and the action to be taken by SCE&G or the staff are presented in Enclosure No. 2 (reactor systems) and No. 3 (initial startup and tests). The summary of the other items discussed is presented below.
Transient Analysis The items discussed relative tc transient analysis were requests for additional information (RAI) 222.3, 222.4, and 222.5. These i tems covered the analysis of a steam generator tube rupture, the experimental verifi-cation of the LOFTRAN computer code and a request for detailed information needed by the staff to perform audit calculations.
Regarding the first item, we stated that the information provided in the response was not as detailed as what we needed. We stated that we were interested in having Westinghouse describe how the FLASH code was set up to calculate the cressure distributions in the primary and secondary systems. SCE3G stated that they would tell Westinghouse to set up a meeting with us to present this information.
Regarding the second item, RAI 222.4, we had requested that SCE&G perform several tests on the nuclear steam supply system to provide additional experimental verification of the LOFTRAN computer code.
In their response 79041700
.,..g SCE&G stated they did not feel such tests were warranted on Summer.
In the meeting, we stated that we intend to discuss this experimental justification directly with Westinghouse.
Regarding the third item, RAI 222.5, we requested detailed information on parameters used in the analysis of the nuclear steam supply system.
In their response, SCE&G, stated that they had no objection to us visiting Monroeville, Pa. and obtain the information from Westinghouse's files, but they were unwilling to reproduce it in the FSAR. We stated that the reason for requesting it on the b docket was because when we went directly to Westinghouse on a generic basis, Westinghouse would not provide this detailed information.
SCE&G stated that they had no problem having Westinghouse provide the specific ir. formation on the Sumer design and would tell Westinghouse to set up a meeting with us in Monroeville.
Mechanical Engineering We discussed with SCE&G their response to RAI 110.37. This item dealt with the applicant's identifying which systems and components they designed using the square-root-sum-of-the-squares (SRSS) method for combining the loss-of-coolant accident (LOCA) loads and the safe shutdown earthquake (SSE) loads and the justification. We stated that a special working group had reviewed the SRSS method and had concluded (in report NUREG-0484) that it is an acceptable method for the design of reactor coolant pressure boundary components. We stated that we were currently examining the extension of the SRSS method to other systems and components.
We stated that for these systems and components it was necessary for the FSAR to identify the components that do not meet the aFselute sum load combination criteria. This should include both upset and faulted loads.
In response to the meeting's discussion, SCE1G stated that they would consider a revised response to RAI 110.37.
In addition, they would look at components outside of the reactor coolant pressure boundary and identify in the response the requested components.
Materials Engineering In a telecon prior to the meeting, SCE1G had inquired if the inspection of the steam generator tubing prior to installation in the steam generators could be used in lieu of a full inspection of the steam generation tubes prior to fuel load. Our preliminary response was that it could not and SCE&G requested a furtner discussion in the captioned meeting.
In the meeting, SCE&G stated that their interpretation of the Westinghouse Standard Technical Specifications (Tech. Specs.) and, Pegulatory Guide 1.33,
" Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes,"
MAR 2 1 G79 Revision 1 was that the factory inspection of the tubing was an acceptable base line (preservice) inspection. They recognized that going that route meant they would have tc inspect tubes from all three steam generators at the first refueling outage.
Their concern was that a similar plant, had factory inspection of the tubing and also had to perform full inspection of the tubing prior to fuel load. They stated that they were considering a full inspectien, but they did not find it to be a licensing requirement.
We reviewed their bases for the conclusion that the factory inspection was an acceptable preservice inspection. Our conclusion was that the SCE&G was correct.
Dean Tibbitts Light Water Reactors Branch L. 2 Division of Project Management
Enclosures:
1.
Attendance List 2.
Discussion of Items in the Area of Reactor Systems 3.
Discussion of Items in the Area of Initial Tests and Operation 4.
Loss of CCW or Seal Injection to Reactor Coolant Pumps and Reactor Coolant Pump Motors ccs w/ enclosures:
See next page
Mr. E. H. Crews, Jr., Vice President
'#" ' I ~?73 and Group Executive - Engineering and Construction South Carolina Electric & "ias Ccapany P. O. Box 764 Columoia, South Carolina 29218 cc:
Mr. H. T. Babb, General Manager South Carolina Electric & Gas Ccapany P. O. Box 764 Columoia, South Carolina 29218 G. H. Fischer, Esq.
Vice President & General Counsel South Carolina Electric & Gas Company P. O. Box 764 Columbia, South Carolina 29218 Mr. William C. Mescher President & Chief Executive South Carolina Public Service Authority 223 North Live Oak Drive Moncks Corner, South Carolina 29461 Mr. William A. Williams, Jr.
Executive Assistant to the General Manager South Carolina Public Service Authority 223 North Live Oak Drive Moncks Corner, South Carolina 29461 Wallace S. Murphy, Eso.
General Ccunsel South Carolina Public Service Authority 223 North Live Oak Drive Moncks Corner, South Carolina 29461 Troy B. Conner, Jr., Esq.
Conner, Moore & Corber 1747 Pennsylvania Avenue, N. W.
Washington, D. C.
20006 Mr. Mark B. Whitaker, Jr.
Licensing and Staff Engineer South Carolina Electric & Gas Company P. O. Box 764 Columoia, South Carolina 29218 Mr. O. W. Dixon Group Manager, Prodaction Engineering South Carolina Electric & Gas Company P. O. Box 764 Columbia, South Carolina 29213
Mr. E. H. Crews, J r.
v;.R 2 1 1979 cc:
Str. Brett Allen Sursey Route 1 Box 93C Little Mountain, South Carolina 23075
ENCLOSURE 1 ATTENDANCE LIST MEETING WITH SCE3G ON THE SUMMER STATICN FEBRUARY 6 !; 7, 1979 NRC - STAFF GILBERT ASSOCIATES, I'lC.
B. Siegel P. Lanouette G. Ma:etis J. Wermiel B. Clayton WESTINGHOUSE S. Goldberg J. Kovacs
- 0. Oudek J. Branmer D. Marburger D. Sellers R. Surman P. Norian R. Stough S. Salah R. Hollard SOUTH CAROLINA ELECTRIC AND GAS C0ffPANY M. Whitaker C. Price A. Koon J. Connelly
- 0. Bradham R. Clary
ENCLOSURE 2 DISCUSSICN OF ITEMS IN THE AREA 0F REACTM SYSTEMS RAI No.
Discussion 211.30 Overpressure Protection - SCE&G had submitted their response to Branch Technical Position RSB 5-2 six days prior to the meeting.
In the meeting, Westinghouse presentad the response.
They said the response was very similar to the submittal on RESAR 414 The concerns we presented at the meeting were:
(1) the procedures, operator actions, aad other assumptions that defined the starting point for the overpressurization transient and (2) credit taken for non-safety grade equipment.
211.36 Branch Technical Position RSB 5-1 (Cold Shut-211.38 down Issue) - SCE&G's resonse on this matter was 211.107 to be submitted in Amendment No. 12 on February 15, 1979. Westinghouse briefly described the proposed response; the submittal included sections on (1) the cold shutdown scenario, and (2) a single failure evaluation. Two areas that we were interested in were:
(1) justification of operator action and (2) a comrritment to perform a naturai circulation cooldown test. Regarcing the seccr:
a-'e, 3033 aa: :-'ng to ee'erence a nat/11
- lition test -- ce cer#Seme
- at another ve:
ngn~ee fa:-
Regulatory Guioe 1.63.2 - We noted that SCE&G had not committed to conduct a test demonstrating the capability to shutdown and cooldown the reactor from outside the control room. We stated that our position that they conduct such a test would be sent out in the near future.
211.86 Loss of Shutdown Cooling - We noted that there had recently been several licensee event reports (LER's) on loss of shutdown cooling.
At SCE&G's request we identified the specific LER's.
We requested that they evaluate the LER's to see if a similar situation might arise on the Summer design.
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Discussion Reactor Coolant Pumo $eal Failures - A loss of seal injection flow ap;arently lead to a failure of the reactor coolan. pump seals at the Salem plant in the fall of 1973.
A representative from Westinghouse gave their understanding of the event; they attributed the e/ent to a failure by the operator to isolate the No. I seal leak off line.
Westinghouse stated that normally loss of seal injection from the CVCS will not lead to seal failure. We requested that tiey document this.
We presented our position on this matter. Our position for the Summer plant is provided in Enclosure No. J.
(A new RAI on this matter will be issued to the applicants.)
211.100 Recirculation Test - It is our position that a test be conducted on their sump to verify (1) NPSH for the pumps and (2) suppression of vortices in the sump.
We stated that we would accept a commitment to do a model or sump test, subject to a review of the test plan.
211.102 LOCA Analysis - A revised LOCA analysis with the corrected Westinghouse model had been completed.
SCE&G stated that it would be submitted in a future amendment.
211.110 Limiting Section 15.0 Transients - SCEaG's response to this item was not acceptacle. Vo wanted them to identify the transient that pro-duced the lowest departure frcm nucleate boiling ratio and the transient that produced tne peak pressure.
From the curves in Section 15.0, we were unable to determine the most limiting transients.
Refueling Water Storage Tank - We discusred with SCE&G a design deficiency discovered at Seabrook.
Our concerns with the Summer design were (1) the time available for manual actions after the Lo-Lo setooint is reached and (2) vortex formation in the tank.
We stated that a request for infor-mation would be sent out on this item.
(This was covered in a later RAI 211.121) wh4 ch was sent to the applicant on February 23, 1979.
UR2159
,RAI No.
Discussion 211.48 RWST Manual Isolation Valve - We require that position indication be provided in the control room for this valve. They agreed to reevaluate their position indication for this valve.
211.84 Leakage Detection Systems - We require that all instruments used to detect leakage be alarmed in the control room. SCE&G said they would confirm that their design satisfied this require-ment.
SCE&G also clarified that the unidentified leakage detection sumps are not the same as the emergency core cooling system sumps.
211.90 FMEA for RHR & ECC Systems - We were concerned that SCE&G might not have identified all the valves where power must be locked out.
In particular we requested that they look at valve 8885.
211.116 Loss of Instrument Air - SCE&G stated that they had examined the case where there is a complete loss of instrument air but not a partial loss of instrument air. We were particularly concerned about interactions between the control systems that utilize instrument air.
SCE&G stated that thet would reexamine their response in light of the discussion.
211.115 Operator Action Times for Section 15.0 Events -
211.120 It wasn't clear from the response that all transients where operator action is assumed have been identified. Westinghouse asked us to look at their response to RAI 212.144 in the FESAR 414 docket. Regarding RAI 211.120, we stated that subparts 2, 3, & 6 were inadequate and the response to subparts 4, 5, & 7 were not provided.
up o1 3pg RAI No.
Discussion 211.43 We briefly discussed their response to Item 211.43. The response stated that the RHR pumps must be manually restarted if a loss-of-off-site power occurs following a safety injec-tion signal. We stated that the Power Systems Branch would look at tnis and it may not be acceptable.
211.81 We stated that Section 5.2.2.4 should be expanded to verify that check valves in the ECCS would be individually leak tested. We also asked why valve 8377 was not leak tested; SCE&G stated that this valve was not in an ECCS line.
MAR 01 3 73 ENCLOSURE NO. 3 DISCUSSICN OF ITEMS IN THE AREA 0F INITIAL TESTS AND GPERATION RAI No.
Discussion 423.14 This item dealt with the abstracts for 33 tests that were submitted in Amendment No. 11. The entries in the "RAI Nc." column identify the t abparts of this item. Our comments are presented below:
(1)
Fire Protection System - The test abstract needed to be expanded to discuss now the capability of the fire protection system would be demonstrated. The abstract should be expanded to include a discussion of verificaticn of system parameters like spray patterns, flow rates and pressures.
(2)
Transformers - The test abstract should be expanded to state that fast transfers between power sources will be demonstrated.
(4)
D-C System - If ine operation of a d-c component at minimum voltage was verified at the site we wanted them to A;scuss how the test will be performed.
Second we wanted them to state in the test abstract that design load for the d-c system is verified.
(5)
Diesel Generators - We request that SCE3G identify l
how the load-carrying duration test of each diese would be conducted. We al30 noted that it appeared that some diesel generator auxiliaries such as cooling water were not being tested; we requested that they clarify their description of the initial test program to verify that these systems would be included in a preoperational test.
(6)
Response to Loss of Air Test - The test abstract should be expanded or other abstracts provided to include (1) tests for particulates and moisture in the air system, (2) tests of the air system's operability and (3) tests of safety grade air accum-ulators.
(10)
Reactor Coolant System Heatup for Hot Functional Testing -
The acceptance criteria in the test abstract should in-clude a statement that the capacity of the cressurizer relief valves and the steam gene rator safety and dumo valves will be determined by test and checked against the design values.
W t 1 '9M RAI No.
Discussion (12)
Emergency Feedwater System - The test abstract accep-tance criteria for the steam driven pump should be 5 successive, successful cold starts to an appreciable flow rate.
(13) 0.esidJal Heat Removal System - The test abstract snouldinclude demonstration of valve interlocks which prevent overpressurization of the RHR System from tne reactor coolant system.
(14)
Control Room Ventilation System - A test abstract should be added for the control building ventilation system.
(17)
Containment Isolation System - It was not clear from the response that the total time to isolation, including instrumentation delay, would be verified to be less than that required. We suggested that the test absti Ict state that the total isolation time would be verified to be less than that required by the Tech. Scecs.
(19)
RPS Response Measurements - The response should be modified to include analysis or testing of the response of sensors and instrument lines. However, we stated that they did not have to address response times of nuclear instruments or thermocouples.
(20)
Rod Drop Time Measurement - The acceptance criteria in the test abstract should include a check of the proper operation of the dashpot. Measuring the time to pass through the dashpot region was not necessary; they should verify that the rod was slowed down by the dashpot and did not stick.
(22)
Reactor Coolant System Flow Measurement - The abstract should also require verification that the flow does not exceed any design or technical specification limits.
(23)
RTD Bypass Loop Flow Verification - The acceptance criteria in the test abstract should refer back to the Tech. Spec resconse time of the overtemperature IT channels which includes loop delay time and RTD response time.
(24)
RCS Leak Test - The title of the abstract should be changed to the 0-ring leak test.
(25)
Low Pcwer Test - The resconse should provide the basis for (1) the acceptance margin on r.' 3 critical boron concentration and (2) the acceptable difference between measured and predicted rod worths.
una e 1 379 RAI No.
Discussion (26)
Process Computer - The test should verify the alarm functions and cneck out any LCO monitoring that the process computer does.
(31)
Plant Loss of Electrical Load - The respcnse should verify that the plant behaved as predicted by realistic or best estimate calculations. We stated that comparing the plant response to a Section 15.0 worst case analysis may lead to the wrong conclusion.
(32)
Loss of Offsite Power - We reiterated our position that this test should include a turbine trip simul-taneous with the loss of offsite power and inter-ruption of offsite power should be maintained for 30 minutes.
423.26 Based on our review of the description of the method 423.27 used to develop preoperational and startup tests 423.28 procedures and results, we could not conclude that it was acceptable.
The information was not clear on who must review and concur in the test procedures, how comments were resolved, and the qualifications of those people concurring in the test procedures.
We stated that at least five qualified people should review the test procedures and test results.
- However, SCE&G did not agree that this should be done for the Phase II tests. As a result of our discussion with SCE&G it appeared tnat their method was probably acceptable, but the description in the FSAR was not adequate.
SCE&G was to revise the FSAR.
423.30 (2)
We noted that SCE&G's plans for testing primary and secondary relief valves was incomplete. We requested that they modify the test abstract when their plans are developed.
(6)
The response did not verify the heat removal capa-bility of the post-accident containment heat removal system. We stated that tests measuring cooling capacity should be run under ambient conditions and then test results extrapolated to accident conditions.
(9)
Load Testing of the Polar Crane - SCE&G stated that the crane had been tested to 125% of the design load.
The test abstract will be modified to include this information and to describe the maintenance and acministrative programs used through construction and preocerational testing of other systems to show tnat the original tests of the crane have not been invalidated.
. vaA ' i 17s RAI No.
Discussion (10)
Natural Circulation Te:t
'.s'e noted that SCE&G had not committed to perform this test nce had they provided enough information to reference a test in a similar plant.
They should provide this information when available.
423.31 The response to this item failed to descr";e the testing of the system or methods used to detect leakage from emergency coce cociing components outside of containment.
A new test abstract addressing leakage out of ECC Systams will be generated.
423.32 The problem with this response is the same as that for RAI 430.30(6).
MAR 2 1 599 ENCLOSURE NO. 4 Loss of CCW or Seal Injection to Reactor Coolant Pumos and Reactor Coolant Famo Motors 1.
A single failure in the compcnent cooling water system or in the CVCS (seal injection) shall not result in fuel damage or damage to the reactor coolant pressure boundary beyond normal makeup capability.
Single failure includes operator error, spurious actuation of motor-op2 rated valves, and loss of a pump.
2.
A moderate energy leakage crack or an accident that is initiated frcm a failure in the component cooling water system or in the CVCS (seal injection) shall not result in excessive fual damage (10 CFR 100 limits) or exceeding the requirements of a loss of coolant accident in 10 CFR 50.46 Those p! ants for which it has been determined that a single failure or pipe break will cause : loss of CCW or seal injection to reactor coolant pumps and motors (and automatic protection is not provided) may assume operator action 10 minutes after indication in the control room if the following information is submitted:
1.
A demonstration that the reactor cociant pumps and motors are capaole of operating with loss of the system which is subject to single failure or break (CCW or seal injection) without loss of function or tne neec for operator action for a least 10 minutes af ter indication is provice in the control rcom.
.z.
Mio s 1 y 2.
Safety-grade instrumentation to detect the loss of CCW or seal injectica to the reactor coolant pumps and motors and alarm in the control rocm 3.
An analysis whien shows that simultaneous multiple pump seizure is unlikely.
4.
An analysi of a locked RCP rotor transient followed by a flow degradation due to additional pump failures (both core performance and system pressure are to be considered).
5.
A commitment for an operating procedure for a loss of ccmponenet cooling water or seal injection to reactor coolant pumps and motors.
6.
A description of the reactor coolant pump testing to support item I above.