ML17258A924
| ML17258A924 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 03/27/1981 |
| From: | Maier J ROCHESTER GAS & ELECTRIC CORP. |
| To: | Crutchfield D Office of Nuclear Reactor Regulation |
| References | |
| TASK-05-11.A, TASK-06-07.C1, TASK-5-11.A, TASK-6-7.C1, TASK-RR NUDOCS 8104020530 | |
| Download: ML17258A924 (48) | |
Text
t REGULATORY IN MATION DISTRIBUTION SYSTE RIDS)
ACCESSION NBR:8104020530 DOC SEDATE: 81/03/27 NOTARIZED:
NO, DOCKET FACIL':50 24'obert Emmet Ginna Nuclear Planti Unit ii Rochester G
05000240 AUTH'AME AUTHOR AFFILIATION MAIERgJ ~ E ~
Rochester Gas L Electric Corp'ECIP
~ NAME RECIPIENT AFFILIATION CRUTCHF IELD p D, Operating Reactors Branch 5
SUBJECT:
Ack receipt of 610227 ltr w/fina) assessments for SEP Topics V"11,A 8 VI-7.C, 1.None of utils alter natives to criteria addressed by NRC ~ Modified proposed topic assessments encl include new info re automatic closure,,of RHR valves'ISTRIBUTION CODE:
A035S COPIES RECEIVED:LTR ENCL
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TITLE: SEP Topics NOTES:1 copy:SEP Sects Ldr ~
05000240 EXTERNAL:
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NSIC 14 05 RECIPIENT ID CODE/NAME ACTION:
CRUTCHP IELD 04 INTERNAL:,A/D MATLLQUAL43 HYD/GEO-BR 10 NRC 02 F IL 01 COPIES LTTR ENCL 7
7 1
1 2
2 1
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RECIPIENT ID CODE/NAME CONT SYS A
07 I8E'6 OR ASSESS BR 11 SEP BR 12'PDR 03 COPIES LTTR ENCL' 1
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3 3
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SR 3 I81 TOTAL NUMBER OF COPIES REQUIRED:
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ROCHESTER GAS AND
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laal 0 laaO) 7 ELECTRIC CORPORATION o
89 EAST AVENUE, ROCHESTER, N.Y. 14649 JOHN E.
MA I ER VICE PRESIDENT TCLCPHONC Aec* cope Tc-i
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- 4. 0
SUMMARY
The R.E. Ginna Nuclear Station has three systems with a lower design pressure rating than the
The CVCS, SIS, and RHR systems do not use the methods prescribed by current regulatory criteria for isolation of high and low pressure systems.
- However, acceptable alternative
- methods, including system design and administrative controls, provide assurance that proper separation of low pressure systems from the RCS at Ginna exists.
5.0 REFERENCES
2.
3.
4 5.
6.
7.
8.
NUREG-075/087, Standard Review Plan 6.3.
Updated Final Facility Description and Safety Analysis
- Report, Ginna Nuclear Power Plant, Unit No. 1.
RG&E drawings 33013-422,
-424, -425, -426, -427, -428,
-432, -433, -434, -435, and -436.
RG&E drawings 10905-280,'285,
-287, -295,
-296, -300, and -301.
Regulatory Guide 1.139, Proposed Revision 1, February 25, 1980.
Amendment No.
16 to the Ginna Operating License, May 14, 1975.
- Letter, D.
H. Lauderbach, EGSG, to James
- Knight, NRC, "SEP Safety Topics V-ll.A and V-11.B for R.E. Ginna",
February 21, 1980.
Letter, Dennis L. Ziemann, NRC, to L. D. White, Jr.
RG&E "Verification of Plant Information on SEP Topic V-ll.A", December 21, 1978.
/
4
9.
Letter, L. D. White Jr.,
RG&E, to Dennis
- Ziemann, NRC, "Verification of Plant Information on SEP Topic V-11.A",
January 25, 1979.
10.
Letter, L. D. White, Jr.,
RG&E, to Dennis L. Ziemann,
- NRC, "LWR Primary Coolant System Pressure Isolation Valves," March 14, 1980.
11.
Letter, Dennis M. Crutchfield, NRC, to Leon D. White, Jr.
RG&E, "SEP Topics V-10.B, V-ll.A, V-11.B, VII-3, and IX-3 (Safe Shutdown Systems)"
and enclosure "SEP Review of Safe Shutdown Systems for the R. E. Ginna Nuclear Power Plant, Revision 1," dated November 14, 1980.
12.
- Letter, Dennis M. Crutchfield, NRC, to Leon D. White Jr.
- RGRE, "SEP Topics III-lO.A, V-11.A, VI-7.C.l, VI-7.F, and VIII-3.B," August 20," 1980.
13.
Letter, John E. Maier, RG&E, to Dennis M. Crutchfield,
- NRC, "SEP Topics III-10.A, VII-A, VI-7.C.1, VI-7.F, and VIII-3.B", dated January 8, 1981.
14.
Letter, John E. Maier, RGSE, to Dennis M. Crutchfield,
dated January 8,
1981.
15.
- Letter, Dennis M. Crutchfield, NRC, to John E. Maier,
- RGGE, "SEP Topics V-11.A and VI-7.C.1, dated February 27, 19 81.
16.
Letter, John E. Maier, RGB E, to Dennis M. Crutchfield,
- NRC, "SEP Topics V-11.A and VI-7.C.1", dated March 27, 1981.
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Enclosure 2
SEP TECHNICAL EVALUATION TOPIC VI-7.C.1 INDEPENDENCE OF REDUNDANT ONSITE POWER SYSTEMS FINAL DRAFT R.
E.
GINNA NUCLEAR STATION Docket No. 50-244 March 27, 1981
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TABLE OF CONTENTS
1.0 INTRODUCTION
Pacae 2.0 CRITERIA
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2.1 AC Supplies 2.2 DC Supplies 3.0 DISCUSSION AND EVALUATION 3.1 AC Supplies 3.2 DC Supplies 4. 0
SUMMARY
5.0 REFERENCES
1
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2 2
2 3
5 5
0 I
A g J
I d
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SEP TECHNICAL EVALUATION TOPIC VI-7.C. 1 INDEPENDENCE OF REDUNDANT ONSITE POWER SYSTEMS FINAL DRAFT R.
E.
GINNA NUCLEAR STATION
1.0 INTRODUCTION
The objective of this review is to determine if the onsite electrical power systems (AC and DC) are in compliance with current licensing criteria for electrical independence between redundant standby (onsite) power sources and their distribution systems.
General Design Criterion 17 requires that the onsite electrical power supplies and their onsite distribution systems shall have sufficient independence to perform their safety function assuming a single failure.
Regulatory Guide 1.6, "Independence Between Redundant Standby (Onsite)
Power Sources and Between Their Distribution'ystem,'"
and IEEE Standard 308-1974, "IEEE Standard Criteria for Nuclear Power Generating Stations" provide a basis acceptable to the NRC staff for meeting GDC 17 in regards to electrical independence of onsite power systems.
- 2. 0 CRITERIA As stated in Regulatory Guide 1.6, Section D.4, the following independence criteria apply.
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When operating from standby sources, redundant load groups and redundant standby'sources should be independent of each other at least to the following extent.
2.
3.
The standby source of one load group should not be automatically paralleled with the standby source of another load group under accident conditions.
No provisions should exist for automatically transferring one load group to another load group or loads between redundant power sources.
If means exist for manually connecting redundant load groups together, at least one interlock should be provided to prevent an operator error that would parallel their standby power sources.
As stated in Regulatory Guide 1.6, Section D.3, each d-c load group should be energized by a battery and battery charger.
The battery-charger combination should have no automatic connection to any other redundant d-c load group.
3 ~ 0 DISCUS SION AND EVALUATION Discussion.
Ginna onsite emergency AC power system consists of two redundant diesel-generator power trains.
Diesel generator 1A (DG1A) supplies 480 V buses 14 and 18 while diesel generator 1B (DG1B) supplies buses 16 and 17.
Manual means exist to tie buses 17 and 18 through a tie breaker and to tie buses 14 and 16 through a tie breaker.
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The control circuit for each'breaker provides interlocks such that the breaker cannot be shut if either DG is closed on either bus or if the normal feeders to the bus are closed."'dditionally, if the tie breakers are closed, they will trip open upon restoration of normal power, DG closing on the bus, or any safety injection signal.
Means exist to power safety injection pump SI-IC from either bus 14 or 16.
The control circuit for the breaker from each bus is designed such that shutting of one breaker prevents shutting the other breaker so that paralleling the redundant DGs is prevented.
Instrument buses 1A, 1B, 1C, and 1D are capable of being supplied by multiple sources.
Each bus is supplied by a pair of mechanically interlocked breakers such that paralleling of redundant sources is prevented.
Evaluation.
The redundant onsite AC power trains have no automatic transfers of loads and/or load groups.
The manual transfer of load groups or manual interconnection of emergency buses have the required interlocks to prevent inadvertent paralleling of redundant sources.
Therefore, the onsite emergency AC system is in compliance with current licensing 0
, requirements for independence of onsite power systems.
Discussion.
Ginna Nuclear Station has two redundant battery and charger trains to supply 125 V DC emergency loads.
Each train consists of a battery, a 75-amp charger, and a 150-amp charger.
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Means exist to interconnect both trains by manually shutting a tie breaker.
This breaker is padlocked open and the key is maintained by the shift foreman.
Current operating procedures require removal of the feeder fuse from one of the buses feeding the tie breaker prior to closing the tie breaker Although no interlocks exist to prevent closure 3
of the tie breaker if the feeder fuse has not been
- removed, the administrative procedures in effect at Ginna ensure that there would be no paralleling of the redundant DC trains.
Automatic transfer of 125 V DC load groups from train A to B r
ll (or vice versa) occurs in seven locations.
Control power I
for 480 V switchgear on buses 14, 16, 17, and 18, DG1A control panel, DG1B control panel, and the rod drive MG set control panel automatically transfers to the redundant train on a loss of power from the normal source.
Each load will automatically transfer back to the normal supply when it is regained.
Separation between trains is provided by redundant fuses (in series) which are sized and coordinated to prevent a load fault from affecting both redundant safeguards trains.
Evaluation.
The 125 V DC system has one manual tie between redundant trains and seven automatic transfers of power from one redundant train to the other.
Administrative controls and fuse protection are provided to prevent paralleling redundant trains via the tie breaker, in lieu of physical or electrical interlocks to prevent parallel operation of the two trains.
Although the 125 V DC system is not in compliance
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with current regulatory criteria with respect to independence of onsite power systems, independence is maintained by alternative acceptable means.
4. 0
SUMMARY
The review of docketed information and plant electrical drawings indicate that the Ginna Nuclear Station onsite AC redundant power sources and distribution system meet the current licensing requirements for independence of onsite power systems.
The 125 V DC system has seven automatically transferred loads and one manual tie breaker which are not in compliance with current regulatory criteria for independence of onsite power systems.
However, administrative controls and fuse protection provide acceptable alternative means to ensure the required independence.
5.0 REFERENCES
2.
3.
4.
5.
General Design Criterion 17, "Electrical Power System,"
of Appendix A, "General Design Criteria of Nuclear Power Plants," to 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities."
"Independence Between Redundant Standby (Onsite)
Power Sources and Between Their Distribution Systems,"
Regulatory Guide 1.6, March 1971.
Rochester Gas and Electric Corp. letter (White) to NRC (Ziemann) dated April 18, 1979.
RGSE Corp. drawings 10905-59, 62, 63, 74 and 75.
RG&E Corp. drawings D-206-51, 21489-269, and 33013-652.
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k J
I,
6.
RG&E Letter, L. D. White, Jr. to Dennis L. Ziemann (NRC), "Technical Specification Change for the Inclusion of Two 75-Amp Battery Chargers,"
August 10, 1979.
7.
Letter and attachments from Dennis M. Crutchfield,
- NRC, to Leon D. White, Jr.,
- RGSE, "SEP Topics III-10.A, V-ll.A, VI-7.C.1, VI-7.F, VIII-3.B," dated August 20, 1980.
8.
Letter, John E. Maier, RGGE, to Dennis M. Crutchfield, NRC g "SEP Topics III-10 ~ A, V-11 ~ A, VI-7~ C ~ 1, VI-7~ F, VIII-3.B, " dated January 8,
19 81.
9.
- Letter, Dennis M. Crutchfield, NRC, to John E. Maier,
- RGSE, "SEP Topics V-11.A and VI-7.C.1," dated February 27, 1981.
10.
Letter, John E. Maier, RG&E, to Dennis M. Crutchfield,
- NRC, "SEP Topics V-11.A and VI-7.C.1", dated March 27, 1981.
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(wc/ocuva EQUIP!ANT SPECIPICATION Design temperature of the motor operated. valve assembly use maximum service temperature listed in the valve specification'heets.
4.2.4 Electrical supply to motor - see Valve Specification Sheets.
4,2.5 4.2.6 Maximum differential pressure across valve-'disc during opening and closing see Valve Speci ication sheets.
Differential pressure acro"s valve disc during hydr'os"atic test, both directions, shall be 1.5 times the design pressure referred by, Paragraph 4.2.2 above, unless noted otherwise.
4.2.7 4.2.8 4.2.9 Ambient temperature -
20 F (49 C) except, as noted.
Open-shut cycles 50 per year for 40years.
The properties of the working fluids WF are defined 'n Addendum "A".
4.P. 1/!
All valves shall be capable of acing heated 'and cooled at a rate cf 100 F(38 C) per hour between 40 F (4.44 C) and the design temperature of thc valve.
In addition, certain valves sha!.1 be capable of sustaining cyclic thermal transients (TT) as defined in Addendum "A".
4.2.11 The maximum allowable i'luid head loss coefficient, Lli), or Cv for each valve is specified in tne valve specification sheet.
The coefficient L/D is the equivalent length in pipe diamet rs of connected straight pipe which will produce the same head loss as the valve itself under 'c same flow conditions with the valve fully
- open, The Cv value is the flow coefficient of the valve measured
'in gallons per minute corresponding to a pressure drop of one pound per square inch.
4.3 Mechanical 4.3,1 The pressure containment, parts of'he valve assembly shal.l be designed in a<:c<irdance with ASA ~!6,5, "Steel Pipe Flange." and Flanged F'ttings,"
or MSS-SP 66 "Pressure
'i'cmperature Ratings for Steel
.?ut,t We..din~
=..'nd Valves."
4.3,2 The valve" stuf'fing box designed for radioactive va3vc specification sheet.",)
s):: ll be provided with a lant~':.
ring leakoff connection wi~.!> a minimum of a ful'l::-;..t o" u',c;.i~/g below the lantern ring and:
maximum of one half of a ct of packing above the lantern ring.
(A fuxx set of p c!.ing is WESTINGHOUSE EIZCTRIC CORPORATION ATMfC POWER DIVISION WAPD FORM 412 Revision No.
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