ML19263D919
| ML19263D919 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 03/23/1979 |
| From: | Baer R Office of Nuclear Reactor Regulation |
| To: | Crews E SOUTH CAROLINA ELECTRIC & GAS CO. |
| References | |
| NUDOCS 7904170059 | |
| Download: ML19263D919 (12) | |
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VAR 2 3 7979 Docket No. 50-395 Mr. E. H. Crews, Jr.
Vice President and Group Executive Engineering and Construction South Carolina Electric & Gas Company P. O. Box 764 Columbia, South Carolina 29218
Dear Mr. Crews:
SUBJECT:
REQUEST FOR ADDITIONAL ItlFOPO1ATION ON THE FINAL SAFETY ANALYSIS REPORT (FSAR) FOR THE VIRGIL C. SUffiER NUCLEAR STATION As a result of our review of your responses to the first round responses tn. the materials engineering requests for additional infornation, we find that we need additional information to complete our review. The informa-tion we need to complete our review is identified in the Enclosure. Also identified in the Enclosure are several staff positions that the Summer application must comply with.
We request that you respond to these items by April 30, 1979.
If you are unable to meet this date, please provide your schedule for responding within two weeks of the receipt of this letter. Also, if you have any questions about the Enclosure, please contact us.
Sincerely,
. MM e -r, Robert L. Baer, Chief Light Water Reactors 3 ranch No. 2 Division of Project Management
Enclosure:
Request for Additional Information
.ccs w/ enclosure:
See next page 7904170059
Mr. E. H. Crews, Jr., Vice President and Group Executive - Engineering and Construction South Carolina Electric & Gas Company P. O. Box 764 Columbia, South Carolina 29218 cc:
Mr. H. T. Babb, General Manager South Carolina Electric & Gas Company P. O. Box 764 Columbia, 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, Esq.
General Counsel 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 Columbia, Scuth Carolina 29218 fir. O. W. Dixon Group Manager, Productici agineering South Carolina Electric & Gas Company P. O. Box 764 Colurbia, South Carolina 29218
20 Mr. E. H. Crews, J r.
cc:
Mr. Brett Allen Bursey Route 1 Box 93C Little Mountain, South Carolina 29076 s
ENCLOSURE REQUEST FOR ADDITIONAL INFORMATION ON THE FINAL SAFETY ANALYSIS REPORT
_FOR THE VIRGIL C. SUMMER NUCLEAR STATION
MAR 2 31979 121.0 MATERIALS ENGINEERING 121.8 Your responses to 0121.1 and 0121.3 concerning fracture tougnness information for ferritic materials in tne reactor coolant pressure boundary are inadequate.
It is indicated in Table 5.2.3,
" Reactor Coolant Pressure Soundary Materials - Class 1 Primary Components" that SA 533 Grade A Class 2 and SA 508 Class 2a steel might be used in the pressurizer and steam generators. These steels are low alloy ferritic steel having minimum yield strengths of 70 ksi and 65 ksi, respectively. Appendix G of 10 CFR Part 50 states that the adequacy of the fracture toughness of ferritic steels having a minimum specified yield strength greater than 50 ksi shall be demonstrated to the Commission on an individual case basis.
To satisfy this reouirement your responses indicate that the recuired data are contained in Westinghouse topical report WCAP 9292, " Dynamic Fracture Toughness of ASME SA 508 Class 23 ASME SA 533 Grade A Class 2 Base and Heat Affected Zone Material and Applicable Weld Metals." This report is currently under review by the staff. Mcwever, if the topical report is found to be acceptable it will only satisfy the generic requirenents of Appendix G of 10 CFR Part 50. That is, the adequacy of the subject materials to be described by the X, of the ASME Code will have been demonstrateb; curve of Apoencix G however, the specific materials used in the V. C. Summer nuclear plant also must meet all the fracture toughness requirements of 10 CFR Part 50, Appendix G.
In addition, we require that the information required by Note 1 of Code Case 1528-3, later incorporated into the ASME Code as paragraph G-2110(b), be provided in the :SAR for any high strength ferritic material used in a pressure-retaining component of the RCPB.
121.9 Your response to Q121.4, which requested operating temoerature information on primary comoonent supoorts is incomplete. We require that the temperature for primary component suoports be identified for start up conditions (primary system pressure of 1000 psi and above).
121.10 Your response to Ql21.6, requesting information on your preservice inspection program of ASME Code Class 1, 2 and 3 comoonents in accordance witn 10 CFR Part 50 paragraon 50.55a(g) did not provide the requested information. The preservice inscec* ion plan must be submitted to support the safety evaluation re: ort finding on ISI.
..uR 2 31979 Additionally steam ger.erator tubing is covered by tne Technical Specifications in lieu of the recommendation of ne Reculatcry Guide 1.83, therefore tne ciscussion of that guide is not gerhane.
121.11 Your response to 0121.7 concerning turbine disk integrity is not satisfactory because you reference a report tnat treats the general aspects of turbine missile generaticn and strike probability, rather nan the aspects of turbine disk material i n tegri ty.
Specifically you should address the requirements of Branch Technical Position MTEB 10-1 (Revision 2), " Turbine Disk Integrity" dated July 10, 1975 which is appended to Standard Review Plan 10.2.3 of NUREG-75/087.
211.122 REACTOR SYSTE:iS BRANCH MAR 2 31973 Staff Position on ATWS for the Vircil C. Summer Nuclear Station We consider ATWS to F2 an unresolved safety issue.
However, we have described the type of plant modifications which, if provided, would reduce ATWS risk to an acceptable level. Volume 3 of NUREG-0460 which describes the rationale for specifying these plant modifications is being reviewed by the Advisory Committee on Reactor Safeguards.
The Regulatory Requirements Review Committee has completed its review and concurred with our approach described in Volume 3 of NUREG-0460 insofar as it applies to Summer. We plan to issue requests for the incustry to supply generic analyses of ATWS mitigation capability anc anticipate presenting to the Commission in May 1979 our recommendations for its actions to resolve the ATWS concern. Summer would be required to implement plant modifications in conformance with the Commission's final resolution on this issue.
We require that the applicant agree to implement modifications on a schedular basis in conformance with the Commission's final resolution of this issue.
In the event that Summer starts operation before necessary plant modifications are implemented, we require some interim actions be taken by the applicant in order to reduce, further, the risk from ATWS events. The applicant is required to:
1.
Develop emergency procedures to train operators to recognize an ATWS event, including consideration of scram indicators, rod position indicators, flux monitors, pressurizer level and pressure indicators, pressurizer relief valve and safety valve indicators, coolant average temperature, containment temperature and pressure indicators, steam generator level, pressure and flow indicators, a rii any other alarms annunciated in the control room with empnasis en alarms not processed through the electrical portion of the reactor scram system.
MAR 2 31979 2.
Train operators to take actions in the event of an ATWS including consideration of manually scramming the reactor by using tne manual scram buttons, prompt actuation of the auxiliary feedwater system to assure delivery of tne full capacity of this system, and initiation of turbine trip. The operator should also be trained to initiate boration by actuation of the high pressure safety injection system to bring the plant to a safe shutdown conditicn.
Early operator action as descrioed above would provide significant protection for all ATWS events which occur (1) as a result of common mode failure in the electrical portion of the scram systen anc (2) those which occur due to a common mode failure in the scram breakers or the rod drive system for which excessive primary pressures are prevented by actuation of turbine trip.
MAR 2 31979 211.123 STAFF POSITICN ON Loss of CCW or Seal Injection to Reactor Coolant Pumos and Reactor Coolant Pumo Motors 1.
A single failure in :ne component cooling water system or in :ne 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-operated valves, and loss of a pump.
2.
A moderate energy leakage crack or an accident that is initiated from a failure in the component cooling water system or in tne CVCS (seal injection) shall not result in excessive fuel damage (10 CFR 100 limits) or exceeding the requirements of a loss of coolant accident in 10 CFR 50.46 Those plants for which it has been determined that a single failure or pipe break will cause a 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 coolant pumps and motors are capable of operating with loss of the system which is subject to single failure or break (CCW or seal injection) without loss of function or the need for operator action for a least 10 minutes after indication is proviced in the control room.
- Ah o c cl3 2.
Safety-grade instrumentation to detect the loss of CCW or seal injection to the reactor coolant pumps and motors and alarm in the control rocm 3.
An analysis which shows that simultaneous multiple pump seizure is unlikely.
4.
An analysis of a locked RCP rotor transient followed by a flow degradation due to additional pump failures (both core performanca and system pressure are to be considered).
5.
A commitment for an operating procedure for a loss of componenet cooling water or seal injection to reactor coolant pumps and motors.
S'.. A description of the reactor coolant pump testing to support item 1 above.
MAR 2 3 I979 311.0 ACCIDENT ANALYSIS GRANC:1 STAFF POSITI0g3 311.31 With regard to a postulated fuel handling accident inside the reactor containrent, you have provided the results of an analysis of the radiological conse-quences of such an accident using assumptions wnicn are comparable to those given in Regulatory Guide 1.25.
We have independently evaluated the potential for releases of activity should this accident occur.
During refueling operations the containment atmos-phere is exhausted through non-seismic Category I ductwork and charcoal filters to the plant vent.
Since the filters are outside containment and the ductwork is not seismically designed, no credit can be given for the iodine removal by filtration of the purge system exhaust. Also, the design does not presently have the capability to preclude a significant release by rapid isolation of the seismic Category I purge line isolation values.
We estimated the potential radiological conse-quences assuming that all the activity at the sur-face of the pool was released witnout filtration.
We conclude that the estimated doses from this accident exceed our acceptance criteria.
It is our position that you demonstrate that the ventilation system design either (1) precludes the release of radioactivity by closure to the isolation valves with safety-grade instrumentation or (2) process any releases through seismic Category I filters and ductwork.
311.32 Sodium hydroxide solution is mixed with the re-fueling water supply by gravity flow rather than by eduction or positive displacement pump. As a result, the composition of the spray solution is dependent upon the containment pressure and the relative liquid levels in the refueling water and sodium hydrocide storage tanks, which are in turn affected by potential failures of this and other systems also draining upon the refueling water supply. You have analyzed several scenarios, and have concluded that the spray system can be regulated within acceptable bounds of spray solu-tion composition by choice of suitable orifices within the system piping. We agree that this method is acceptable in principle, but we re-quire that the orifices chosen be tested, prior to plant operation, to demonstrate that, under all combinations of extremes in containment pressure and tank liquid level, the system will deliver the required quantity of sodium hcyroxide.
19 23 57? It is our position that tne proposed preoperational test for the spray system will not provide assurance that the orifices will deliver tne required flow under all design basis conditions.
311.33 The containment spray system is designed to activate automatically, even in the event of a single active component failure.
Upon exhaustion of the refueling water supply, however, the system shuts down, and must be restarted in the recirculation mode by manual operations on the control room.
We have evaluated the consequer.ces of the postulated failure of the operating personnel to perform these manual operations and have concluded that doses could equal or exceed the guidelines of 10 CFR Part 100 under scme circumstances.
Our calculations show acceptable doses if it is assumed that the spray system is switched over from the injection mode to the recirculation mode within 60 minutes after termination of the in-jection mode and remain operational for at least an additional 90 minutes. Our analysis of tne loss-of-coolant accident indicates that, if this procedure were not followed, the 0-30 day LOCA dose would exceed the 10 CFR Part 100 guideline values.
Therefore, it is our position that your procedures specify switchover of the spray system from the injection mode to the recirculation made within 60 minutes af ter termination of the injec-tion phase following a LOCA.
.