ML17250A706
| ML17250A706 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 10/22/1980 |
| From: | Maier J ROCHESTER GAS & ELECTRIC CORP. |
| To: | Crutchfield D Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8010280420 | |
| Download: ML17250A706 (13) | |
Text
P REGULWRY INFORMATION DISTRIBUTIMSYSTEM (RIDS)
'CCESSION NBR: 8010280420 DOC ~ DATE: 80/10/22 NO fARIZED:
NO DOCKET FACIL:50-244 Robert Emmet Ginna Nuclear Plantq Unit ii Rochester G
05000244 AUTH,NAME AUTHOR AFFILIATION MAIERpJ ~ E ~
Rochester
'Gas 8, Electric Corp.
REC IP ~ NAME RECIPIENT AF F ILI ATI ON CRUTCHFIELDr D ~
Operating Reactors Branch 5
SUBJECT; Forwards comments on selected topics in preliminary clarification of TMI action plan requirements.
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.2-;,""" j ROCHESTER GAS AND ELECTRIC CORPORATION 0
89 EAST AVENUE, ROCHESTER, N.Y. I4649 JOHN E.
MAIEII VICE PRESIDENT TCLL'PHONE
- RCA CDOC >le 546.2700 October 22, 1980 Director of Nuclear Reactor Regulation Attention:
Mr. Dennis M. Crutchfield, Chief Operating Reactors Branch 55 U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Subject:
Preliminary Clarification of TMI Action Plan Reguirements R.E. Ginna Nuclear Power Plant Docket No. 50-244
Dear Mr. Crutchfield:
Mr. D G. Eisenhut's letter dated September 5,
1980 to All Licensees of Operating Plants and Applicants for Operating Licenses and Holders of Construction Permits provided preliminary clarification of TMI Action Plan Requirements.
We have reviewed that document and have comments on several sections which are presented in the attachment to this letter.
Please
-consider these comments before finalizing your requirements.
Sincerely, Jo n E. Maier attachment go/
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ROCHESTER GAS AND ELECTRIC COMMENTS ON SELECTED TOPICS IN PRELIMINARY CLARIFICATION OF TMI ACTION PLAN REQUIREMENTS I.A.1.3 Shift Manning Our comments and response to this item were provided in a letter dated October 13, 1980 to Dennis M. Crutchfield from L. D. White, Jr.
I.A.3.1 Revise Scope and Criteria for Licensing Examinations It is difficult and may be undesirable to examine license applicants on simulators which are not similar to the plant for which an applicant is seeking a license.
If the control board layout and the plant modeled for the simulator are significantly different. from the operator's plant, the operating and accident response characteristics will probably be different.
A great deal of time and effort would be required to learn a
new plant sufficiently well to be examined on it.
This learning may only cause.confusion on the operator's real plant.
It is valuable to receive simulator training and our operators attend simulator courses now, however, there does not appear to be a good reason for examination on a different machine.
II.B.l Reactor Coolant System Vents New requirements have been imposed which did not exist when we designed our system and submitted it. for NRC review over ten months ago.
A good faith effort was made to meet the NRC's earlier deadline for installation of January 1,
1981 and our system is now operational.
Good engineering judgment was used in the design of the system and we believe we meet all of the requirements, including those recently imposed.
Because the system is installed we do not plan to make modifications to it if new guide-lines are expressed without good justification.
During the design phase of the, project gas release rate limits from the vents were imposed and redundant valving provided to isolate the vent paths.
These provisions assure adequate time for operator action to monitor con-tainment hydrogen concentration during a venting process.
- Thus, no further studies are being conducted concerning the implications of venting either inside or outside containment.
I&
A test program for the vent system has not been finalized yet.
One proposed test method includes stroking the vent, valves at the control board during refueling shutdowns.
Design Review of Plant Shielding and Environmental Quali-fication of Equipment for Spaces/Systems Which May Be Used in Post. Accident Operations.
We have attempted to respond to NRC requirements for emergency facilities and we are now constructing a new building to house the Technical Support Center.
It is difficult to commit to large expenditures when require-ments are changing.
It is especially difficultto commit to major construction projects without.first having NRC approval of the proposed plan.
The implementation of major modifications should come only after NRC review of the proposed project..
As a minimum NRC should provide assurance that the requirements are finalized and will not be further changed.
Post implementation review is unacceptable.
Post Accident Sampling Capability Clarification item 3.b calls for provisions for reducing plateout in sample lines and for preventing blockage of sample lines by loose material.
It also calls for passive flow restrictors to limit reactor coolant loss or contain-ment air leak from a ruptured sample line.
These require-ments appear to be contradictory.
Performance Testing of BWR and PWR Relief and Safety Valves The EPRI valve testing program schedule calls for comple-tion of the relief and safety valve tests by July, 1981.
This is the same schedule that was established at the beginning of the program and provided'o NRC. It is not anticipated that testing will be completed prior to that date.
It will be impossible to provide plant specific submittals for qualifications of the rely.ef and safety valves immediately following completion of the testing.
Auxiliary Feedwater System Automatic Initiation and Flow Indication.
The requirements presented in this section are different from those stated in Harold Denton's October 30, 1979 letter.
RGE has designed and procured instrumentation that will provide a totally redundant flowrate indication which meets the following criteria:
a.
Full environmental qualification of field mounted transmitters.
3 b.
Redundant, design from transmitter to indicator.
c.
Flowrate indication powered by class IE supplies.
d.
Analog instrumentation mounted in qualified IE rack.
We have attempted to provide a well engineered system which meets the requirements that were previously published.
NRC should not expect licensees to implement systems with changing criteria while providing only a post implementa-tion review.
Assurance of stability of criteria or approval of designs is needed before implementation.
Additional Accident Monitoring Instrumentation.
Attachment 1
Noble Gas Effluent Monitors.
. Clarification 4.a.(4) requires a capability to obtain readings at least every 15 minutes during and following an accident..
Table II.F.1-1 then requires the display to be continuous and record-ing as equivalent Xe-133.
These requirements are inconsistent,
- however, we assume that the Eberline Sping-4's purchased for this application are acceptable.
We previously documented our intention to use these monitors on December 28, 1979.
The Sping-4's are capable of automatically logging activity levels every ten minutes when in alarm or at any time when manually requested by the operator.
Attachment.
4 Containment Pressure Monitor.
Clarification 3 provides new accuracy requirements not previously imposed upon the containment monitor-ing system.
The requirement for an accuracy of
%0.25 psi for pressures less than atmospheric, imposes, for our containment with a design pressure of 60 psig, an overall accuracy for the transmitter of 0.15 per-cent,.
Qualified transmitters have an accuracy of 0.5 percent and a repeatability of O.l percent.
Transmitters have been ordered for this application which have a lead time of six to nine months.
If additional narrow range transmitters are required for containment pressures which are less than atmospheric, adds.tional time will be required to procure them.
There does not seem to be a good basis for needing an accuracy greater than 0.5
- percent, however, so we do not intend to purchase a second set of transmitters unless good justifica-tion xs provided.
II.F.2 Clarification 5 requires that containment pressure be continuously recorded in the control room.
Miles of strip chart with nothing but a straight line will be produced over the years if the recorders are run continuously.
A better method would be to actuate a recorder on a safety injection or contain-ment isolation signal or have a data processor sample the containment pressure value, at a greater frequency during transient events.
Attachment 5
Containment Water Level Monitor.
The narrow range containment water level instrument will provide useful information only for leaks or small fluid line breaks.
It will provide no useful information for most LOCA break sizes.
Thus, this instrumentation should be designed and qualified for those postulated accident conditions for which operability of the instrumentation can provide meaningful indication.
Regulatory Guide 1.89 is an appropriate reference for this instrumentation; Regulatory Guide 1.97, Revision 2, is not.
Attachment 6
Containment, Hydrogen Monitor.
Clarification 3 requires that containment hydrogen concentrations be measurable with an accuracy within 1 percent of the monitored range.
This is a new require-ment with no basis provided.
Equipment which has already been ordered for hydrogen monitoring has a stated accuracy of i 2 percent.
This accuracy is sufficient to determined all appropriate post accident operator actions.
We do not intend to order different. equipment (we don't know if equipment with a greater accuracy is available) without good justification.
Instrumentation for Detection of Inadequate Core Cooling.
RGE's position on installation of reactor vessel water level instrument was presented in a letter from L. D.
White, Jr. to Dennis M. Crutchfield, USNRC dated July 2, 1980.
II.K.3.30 Revised Small Break LOCA Methods to Show Compliance with 10 CFR 50, Appendix K The present Westinghouse small break LOCA model used to analyze R. E. Ginna is in conformance with 10 CFR 50, Appendix K.
- However, Westinghouse has indicated that they will, nevertheless, address the issue and have scheduled it for completion by January 1,
1982.
Westinghouse will separately provide a detailed outline of the scope and schedule of this effort.
4 1
I N
II.K.3.31 Plant Specific Calculations to Show Compliance with 10 CFR 50.46.
If the results of the new Westinghouse model (and sub-
- sequent, NRC review and approval) indicate that the present small break LOCA analysis for R.
E. Ginna is not in conformance with 10 CFR 50.46, a new analysis utilizing the new and approved Westinghouse model wall be submitted to the NRC in accordance with the NRC Schedule.
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