ML17258A479
| ML17258A479 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 01/19/1982 |
| From: | Maier J ROCHESTER GAS & ELECTRIC CORP. |
| To: | Crutchfield D Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.1, TASK-2.B.2, TASK-2.D.1, TASK-2.E.1.1, TASK-2.E.1.2, TASK-2.F.1, TASK-2.F.2, TASK-2.K.2.13, TASK-2.K.2.17, TASK-2.K.2.19, TASK-2.K.3.05, TASK-2.K.3.25, TASK-2.K.3.30, TASK-TM NUDOCS 8201270423 | |
| Download: ML17258A479 (20) | |
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S9 EAST AVENUE, ROCHESTER, N.Y. 14649 JOHN E.
MAILER VICE PRESIOENT TELEPHONC AREA coof 7ld 546-2700 January 19, 1982 Director of Nuclear Reactor Regulation Attention:
Mr. Dennis M. Crutchfield, Chief Operating Reactors Branch No.
5 U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Subject:
Status of January 1,
1982 NUREG-0737 Items
.R.
E. Ginna Nuclear Power Plant Docket. No. 50-244 ggt,@<BED 9
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Dear Mr. Crutchfield:
Several items in NUREG-0737 are shown on Enclosure 1 of the document as having schedular commitment dates of 1/1/82.
Most of these items do not require that information be submitted at. this time, however, the purpose of this letter is to document the status of those items with a January 1982 schedular date.
Each of these items is discussed below.
Item II.B.l, Reactor Coolant System Vents, requires pro-cedures to be completed by January 1,
1982 for the vent system which is required to be installed by July 1, 1982.
We have previously provided procedure S-3.3J along with other information for our installed vent system with Reference l.
This procedure may be revised as new emergency procedures are implemented to meet the requirements of item I.C.l.
Items II.B.2, Plant Shielding, and II.B.3, Post-Accident
- Sampling, were to have plant modifications completed by January 1, 1982.'e have addressed the status of construction and provided a schedule for completion of these items in Reference 2.
Item II.D.l, Relief and Safety Valve Tests, originally had a
plant specific reporting requirement which was due January 1,
1982.
The schedules for completing the safety and relief valve testing by EPRI and their contractors and the submittal of plant specific reports were revised by Reference 3 to April 1 and July 1, 1982 respectively.
We expect to be able to complete our plant specific report in time to meet that schedule.
Item II.E.l.l, Auxiliary Feedwater System Evaluation, re-quired implementation by January 1,
1982.
This item and item 820i270423 820ii9 PDR,ADOCK 05000244, P
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January 19-,
1982 To Mr. Dennis M. Crutchfield SHEET NO.
II.E.1.2, Auxiliary Feedwater System Initiation and Flow, have been the subject of substantial correspondence and discussion between the NRC Staff and RGE Staff.
Some of the correspondence and requirements preceded publication of NUREG-0737.
Some of the system requirements and many of the requests for information for NRC review of our auxiliary feedwater system have not been refer-enced to NUREG-0737.
We have, however completed all of the requirements and supplied the requested information as documented in references 4 through 20.
Item II.F.l, Accident Monitoring, required that plants be equipped to monitor noble gas effluents, iodines and particulates, high range containment radiation, high containment pressure, containment water level and containment hydrogen concentration.
All of the necessary equipment except for the high containment pressure recorders and the containment hydrogen monitors is in place and operable.
The high containment pressure transmitters and control room indication are in place and operable but the permanent recorders to develop a history of containment pressure following an event in accordance with reference 22 have not yet been installed.
A temporary recorder is in place.
The permanent recorders are projected to be installed within the next two months although equipment delivery is not yet assured.
A revised schedule for completion of the hydrogen monitoring system was provided in Reference 2.
Item II.F.2, Instrumentation for Detection of Inadequate Core Cooling, includes a requirement to implement a reactor vessel water level instrument.
Previous correspondence (References 21 and 22) have stated our position that a "reliable, easy-to-interpret, unambiguous indication of vessel level" has not yet been demonstrated to exist.
We have committed,
- however, to install additional instrumentation when a device has been successfully demonstrated to function properly over the range of conditions for which it is intended to operate and has been shown to provide useful information to the operator.
)
We note that more than one ACRS member has expressed concern that proposed water level instruments may not perform as intended and may in fact confuse reactor operators under emergency con-ditions.
We are not aware that any water level system, even those installed on operating plants, has received NRC approval for use.
The installed cost, of currently proposed instruments being developed may be several million dollars.
A comparably valued benefit must be reasonably assured before installation.
Because some of the testing of reactor vessel level instru-ments is being performed at national laboratories under contract to the NRC, you may be receiving test results which are not yet available to RGE.
Please provide all available test results to aid us in our evaluation of reactor vessel water level instruments.
Y jI
ROCHESTER GAS ANP ELECTRIC CORP.
pATE January 19, 1982 TO Mr. Dennis M. Crutchfield SHEET NO.
Item II.K.2.13, Thermal Mechanical Report, required a
detailed analysis of the thermal-mechanical conditions in the reactor vessel during recovery from small breaks with an extended loss of all feedwater.
Westinghouse (in support of the Westinghouse Owners Group) has performed analyses for generic Westinghouse plant groupings to address this issue.
The generic study is applicable to the R.
E. Ginna plant.
A report. of the study was sent to the NRC by reference 23.
Item II.K.2.17, Reactor Coolant System Voiding, required that an analysis be performed to address the potential for void formation during natural circulation cooldown and depressurization transients.
Westinghouse (in support of the Westinghouse Owners Group) has performed the required anlysis.
The results of the
- analysis, which are applicable to Ginna, have been submitted to the NRC by Reference 24.
In addition, the Westinghouse Owners Group has developed a
natural circulation cooldown guideline that takes the results of the study into account so as to preclude void formation in the upper head region during natural circulation cooldown and depressurization transients, and specifies those conditions under which upper head voiding may occur.
These Westinghouse Owners Group generic guidelines have been submitted to the NRC by Refer-ence 25.
The generic guidance developed by the Westinghouse Owners Group (augmented as appropriate with plant specific considerations) has been utilized in the implementation of Ginna plant specific operating procedures as documented in Reference 32.
Item II.K.2.19, Sequential Auxiliary Feedwater Flow Analysis, no longer applies to R.
E. Ginna.
Subsequent to the issuance of NUREG-0737, and as documented in Reference 26, the NRC completed a generic review on this subject and concluded that the concerns expressed in this item are not applicable to plants with inverted U-tube steam generators such as those designed by Westinghouse and used at Ginna.
No further action is necessary.
Item II.K.3.5, Automatic Trip of Reactor Coolant
- Pumps, does not have a January 1,
1982 due date, however, it remains an open issue.
Westinghouse (in support of the Westinghouse Owners Group) has performed an analysis of delayed reactor coolant pump trip during small-break LOCAs.
This analysis is documented in Reference 27.
In addition, Westinghouse has performed test predictions of LOFT Experiments L3-1 and L3-6.
The results of these predictions are documented in References 28, 29, and 30.
Based on:
- 1) the Westinghouse
- analysis,
- 2) the excellent prediction of the LOFT Experiment L3-6 results using the
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ROCHESTER GAS AND ELECTRIC CORP.
January 19, 1982 TO Mr. Dennis M. Crutchfield SHEET NO.
Westinghouse analytical model, and 3) Westinghouse simulator data related to operator response time, the Westinghouse and RGE position is that automatic reactor coolant pump trip is not necessary since sufficient time is available for manual tripping of the pumps.
Our understanding of the schedule for final resolution of this issue is.-')
Once the NRC formally approves the Westinghouse
- model, a
3-month study period will ensue during which the Westinghouse Owners Group will attempt to demonstrate compliance with NRC acceptance criteria for manual RCP trip.
The NRC acceptance criteria will accompany formal approval of the Westinghouse models b)
If, at the end of the 3-month period, the West.inghouse Owners Group cannot.
show compliance with the acceptance criteria, the NRC may notify utilities that they must submit an automatic RCP trip design.
Item II.K.3.25, Effect of Loss of AC Power On Pump Seals, requires that the consequences of a loss of RCP seal cooling due to a loss of AC power (defined as loss of offsite power) for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> be demonstrated.
During normal operation, seal injection flow from the chemical and volume control system is provided to cool the RCP seals and the component cooling water system provides flow to the thermal barrier heat exchanger to limit the heat transfer from the reactor coolant to the RCP internals.
In the event of loss of offsite power the RCP motor is deenergized and both of these cooling supplies are terminated;
- however, the diesel generators are automatically started and component cooling water to the thermal barrier heat exchanger is automatically restored within seconds.
This cooling supply is adequate to provide seal cooling and prevent, seal failure during a loss of offsite power for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
Item II.K.3.30, Small Break LOCA Methods, requires that. the analysis methods used by NSSS vendors and/or fuel suppliers for small-brea'k LOCA analysis for compliance with Appendix K to 10 CFR Part 50 be revised, documented, and submitted for NRC approval.
Westinghouse has informed us that it is their position that the small-break LOCA analysis model currently approved by the NRC for use on R.
E. Ginna is conservative and in conformance with Appendix K to 10 CFR Part 50.
- However, Westinghouse believes that improvement in the realism of small-break calculations is a worthwhile effort and has committed to revise its small-break LOCA analysis model to address NRC concerns (expressed, for
, t
ROCHESTER GAS AND ELECTRIC CORP.
o+TE January 1 9, 1 982 TO Mr. Dennis M. Crutchfield SHEET NO.
- example, in NUREG-0611 and NUREG-0623).
This revised Westinghouse model is currently scheduled for submittal to the NRC by April 1, 1982 as documented in Reference 31.
Very truly yours,
II'J R
References RGE letter from John E. Maier to Dennis M. Crutchfield, USNRC, dated July 1, 1981.
RGE letter from J.
E. Arthur to Dennis M. Crutchfield, USNRC, dated November 25, 1981.
USNRC letter from Darrel G. Eisenhut, to All Licensees dated September 29, 1981.
USNRC letter from Darrel G. Eisenhut to L. D. White, Jr.,
RGE, dated October 22, 1979.
RGE letter from L. D. White, Jr. to Dennis
- Ziemann, USNRC, dated November 28, 1979.
USNRC letter from Dennis L. Ziemann to L. D. White, Jr.,
RGE, dated December 12, 1979.
RGE letter from L. D. White, Jr. to Dennis
- Ziemann, USNRC, dated December 14, 1979.
RGE letter from L. D. White, Jr. to Dennis
- Ziemann, USNRC, dated December 19, 1979.
RGE letter from L. D. Nhite, Jr. to Dennis L. Ziemann, USNRC, dated March 28, 1980.
USNRC letter from Dennis L. Ziemann to L. D. White, Jr.,
RGE, dated April 18, 1980.
RGE letter from L. D. White, Jr. to
- USNRC, dated May 22, 1980.
RGE letter from L. D. White, Jr. to
- USNRC, dated May 28,
- 1980, subject:
System Requirements.
RGE letter from L. D. White, Jr. to USNRC, dated May 28, 1980, subject:
Auxiliary Feedwater Systems.
Dennis M. Crutchfield, Dennis M. Crutchfield, Auxiliary Feedwater Dennis M. Crutchfield, NRC Requirements for RGE letter from L. D. White, Jr. to Dennis M. Crutchfield, USNRC, dated July 14, 1980.
USNRC letter from Dennis M. Crutchfield to John E. Maier, RGE, dated January 29, 1981.
USNRC letter from Dennis M. Crutchfield to John E. Maier, RGE, dated March 17, 1981.
I III I,
17.
18.
19.
20.
21.
22.
23.
24.
25.
RGE letter from John E. Maier to Dennis M. Crutchfield, USNRC, dated June 8,
1981.
USNRC letter from Dennis M. Crutchfield to John E. Maier, RGE dated August 19, 1981.
RGE letter from John E. Maier to Dennis M. Crutchfield, USNRC, dated September 22, 1981.
RGE letter from John E. Maier to Dennis M. Crutchfield, USNRC, dated January 8,
1981.
RGE letter from L. D. White, Jr. to Dennis M. Crutchfield, USNRC, dated July 2, 1980.
RGE letter from John E. Maier to Dennis M. Crutchfield, USNRC, dated December 15, 1980.
WCAP 10019, Summary Report on Reactor Vessel Integrity for Westinghouse Operating Plants, submitted by letter OG-66, dated December 30,, 1981 from O.
D. Kingsley,(Chairman, Westinghouse Owners'roup) to H.
R. Denton, USNRC.
ry, q
1 I
Ig Letter OG-57, dated April 20, 1981 from R.
W. Jurgensen (Chairman, Westinghouse Owners Group) to P.
S.
- Check, USNRC.
i Letter OG-64, dated November 30, 1981 from R.
W. Jurgensen (Chairman, Westinghouse Owners Group) to D. G. Eisenhut, USNRC.
26.
27.
28.
USNRC letter from Dennis M. Crutchfield to John E. Maier, RGE, dated June 29, 1981.
"Analysis of Delayed Reactor Coolant Pump Trip During Small Loss of Coolant Accidents for Westinghouse Nuclear Steam Supply Systems,"
WCAP-9584 (proprietary) and WCAP-9585 (non-proprietary),
August 1979.
Letter OG-49, dated March 3, 1981 from R.
W. Jurgensen (Chairman, Westinghouse Owners Group) to D. F. Ross, Jr.,
29.
30.
31.
32.
Letter OG-50, dated March 23, 1981 from R.
W. Jurgensen (Chairman, Westinghouse Owners Group) to D. F. Ross, USNRC.
Letter OG-60, dated June 15, 1981 from R.
W. Jurgensen (Chairman, Westinghouse Owners Group) to P.
S.
- Check, USNRC.
Letter NS-EPR-2524, dated November 25,
- 1981, from E. P.
- Rahe, Westinghouse, to D.
G. Eisenhut, USNRC.
RGE letter from John E. Maier to Dennis M. Crutchfield, USNRC, dated November 13, 1981.
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January 15, 1982 Docket No. 50-244 LS05-82-01-043 Mr. John E. Maier Vice President Electric and Steam Production Rochester Gas and Electric Corporation 89 East Avenue Rochester, New York 14649
Dear Mr. Maier:
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This confirms our telephone authorization given January 13, 1982 for a change in the Technical Specifications for the R.
E. Ginna facility as requested
.by your letter telecopied to us on January 13, 1982.
Facility Operating License No. DPR-18 was amended on January 13, 1982 by making the following revision in the Technical Specification 3.3.1.2:
Reference to the Reactor Coolant Drain Tank Pumps will be eliminated.
The Specification will be revised to permit the inoperability of one ECCS subsystem for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
This would cover inoperability due to pump, heat exchanger, valve, interlock, or pipe. If not repaired in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, hot shut down would be required in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, with TAVE 350'F in an additional 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
Copies of the license amendment, our evaluation and Federal Register Notice for thfs technical specification change will be sent to you when complete.
Sincerely, Original signed by Gus C. Lainas, Assistant Director For Safety Assessment Division of Licensing cc:
See next page OFFICE/
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NRG FORM 318 (10-80) NRCM 0240 OFFICIAL RECORD COPY USGPO: 1981-339.960
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Mr. John E. Maier - January 15, 198K CC Ha'r.ry H. Voigt, Esquire
- LeBoeuf, Lamb, Leiby and MacRae
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1333 New Hampsliire Avenue, N.
W.
Suite 1100 Washington, D. C.
20036 Mr. Michael Slade 12 Trailwood Circle Rochester, New York 14618 Ezra Bialik Assistant Attorney General Environmental Protection Bureau NNew York State Department of Law 2 Worl,d Trade Center New York, New York 10047 Resident Inspector R. E. Ginna Plant c/o U. S.
NRC 1503 Lake Road
- Ontario, New York 14519 Director, Bureau of Nuclear
~
Operations.
State of New York Energy Office Agency Building 2 Empire State Plaza
- Albany, New York 12223 Rochester Public Library 115 South Avenue Rochester,,
New York 14604 Supervisor of the Town of Ontario 107 Ridge Road West
- Ontario, New York 14519 Dr.
Emmeth A'. Luebke Atomic Safety and Licensing Board U. S. Nuclear Regul.atory Commission Washington, D. C.
20555 Dr. Richard F. Cole Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington, D. C.. 20555 U. S. Environmental Protection Agency Region II Office ATTN:
Regional Radiation Representative 26 Federal Plaza New York, New York 10007 Herbert Grossman, Esq.,
Chairman Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington, D. C.
20555 James P. O'Reilly, Regional Administr'ator Nuclear Regulatory Commission, Region II Office of Inspection and Enforcement 101 Marietta Street,. Suite 3100 Atlanta, Georgia 30303
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