ML19250A506
| ML19250A506 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 10/17/1979 |
| From: | Herbein J Metropolitan Edison Co |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| GQL-1294, NUDOCS 7910230536 | |
| Download: ML19250A506 (14) | |
Text
e
C Metropolitan Edison Company g
Post Office Box 542 Reading Pennsylvania 19640 215 929 3601 Writer's Direct Dial %rnber October 17, 1979 CQL 129h Office of Nuclear Reactor Regulation Attn: Harold R. Denton, Director U. S. Nuclear Regulatory Co= mission Washington, D.C.
20555
Dear Sir:
Thre' Mile Island Nuclear Station, Units 1 & 2 (TMI-l i TMI-2)
License Nos. DPR-50 L EPR-73 Docket Nos. 50-289 & 50-320 Safety and Non-Safety Grade System Interaction This letter and the attached report responds to your September 17, 1979 letter
" Potential Unreviewed Safety Question on Interaction Between Non-Safety on Grade Systems and Safety Grade Systems", and is being submitted late as discussed by Mr. D. DiIanni of NRC and Mr. J. R. Stair of my staff on October h, 1979 Within the time frame allowed by your letter, our efforts have been focussed on developing and understandine of the issue and making a preliminary assess-ment of the impact of this issue on the conclusions of the safety analyses presented in the FSAR.
We have not discovered any adverse interactions which we feel warrants license modific ations.
There are, however, areas which vill require additional investigation.
These areas v e detailed in the attached evaluation.
The attached report, which was prepared in conjunction with Babcock & Wilcox and Metropolitan Edison Company, provides the details of our review and identifies further actions that we are undertaking to address the longterm system response under adverse environmental conditions.
Because of the existing situation at TMI-2, the site specific part of this evaluation only addresses TMI-1.
Sincerely,
/
l n
V J. G. Herbein Vice President-Nuclear Cperations c
3 JGH :LWH :t as h79g Attachment a0 i
7r 120.7; 7 0102 3 0JrJ g vetrcsoe Ec sen Ocmcary s a 3/e-ser o me Gern 6c w m Siser-
ME"'EOPOLITXI EDISCH CCMPXFf JERSEY CE'ITP?L PCWER & LIGFT CCMPA'IY XID PE !'IS?LVA' IIA ELECTRIC COMPEIY THREE '4ILE ISLAND NUCLEAR STATION, UNI'"S 1 & 2 0;eratinr; license Nos. DPR-50 and EPR-73 Docket Nos. 50-289 and 50-320 This letter is submitted in support of the Nuclear Regulatory Commission's request concerning safety and non-safety grade system interaction, dated September 17, 1979, for Three Mile Island Nuclear Station, Units 1 and 2.
As a part of this request, an " Evaluation of Potential Advers e Environmental Effects on Non-Safety Grade Control Systems" is attached. Further, all statements contained in this report have been reviewed and all such state-ments made and
.ters set forth therein are true and currect to the best of my kucvledge, information, and belief.
METECPCLITKI EDISCN CCMP%Ff 0
u,,
J
/
lice President
/
N day of
, 1979 Sworn and subscribed to me this 0
Notary Public #
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E7ALUATICN CF FCTE'ITIALLY ALVERSE DVIRCIIMEiTAL EFFECTS C?I !ICN-SAFE"'Y GPADE CCIGOL SYSTDIS Prepared by:
h.nCFCLI~'NT DISCII CCMPXiY and BABCCCK A'iD '4ILCCX CCMPriY 1203 352 Cetober
, 1979
TAELE OF CCUTENTS Page I.
Introduction 1
II.
Plant Licensine Basis 2
1.
Safety Analysis Functions and Parameters 2,
Plant Unique Features III.
Safety Assessment h
1.
Potential Environmental Effects 2.
l= pact on Plant Safety Analysis IV.
Justification for Cperation 6
7 Recommended Follow-ur Action 7
Tables I.
Typical Equipment Response During High Energy Line Breaks II.
Potential Environmental Effects on Non-Safety Grade Control Systems III.
Impact of Control System Effects on Safety Analysis IV.
Fuel Performance at Operating Plants 1203 453
I.
Introduction This report is in respense to !?1rold R. Denton~s letter of Septe=ber 17, 1979, on the subject of "?ctenti11 Unreviewed Safety 0.uestien on Interaction 3etween Non-Safety Grade Systems and Safety Grade Systems. "
It addresses the concerns listed in Information Notice 79-22 sad fulfills the ec==ittent =ade during our meeting with your staff on Septe=ber 20, 1979 During that meeting, we ec==itted to:
Evaluate the impact on the Licensing basis accident analyses due to consequential environmental effects on non-safety grade cont rol systems.
- Identify Licensing tasis accidents which cause an adverse environment for each plant.
- Define Safety Analysis inputs and responses used during Licensing basis accidents
- Verify Safety Analysis conclusions or reco==end actions justifying operation.
The scope of this response includes an evslua. ion of the plants equipment actuation and performance and effects on the Licensing basis analysis.
A = atrix of potential environmental effects on non-safety grade control syste=3 is presented.
Where non-safety grade equipment performance could be affected by the adverse environment, a
safety assessment vill be prepared.
The safety assessment vill define any potential problems which may arise due to the effects of an adverse environ =ent on ncn-safety grade control syste=s.
Work beyond the scope of the 20-day response and work to provide a ore detailed assessment are included in recc== ended followup actions.
1203 354
II.
Plant Licensing Sasis 1.
Safety Analysis Punctions and Farameters The plant licensing basis analyses, as presented in the FSA3, were reviewed to define the inputs, assumptions and responses used for non-safety grade control systems.
This information is summarized in Table I, which lists typical equip =ent actions and actuation times used in the safety analyses for B&W 177 fuel assembly plants.
The data has been categorized to reflect the functional requirements as follows:
A.
Reactor Power Control and Shutdown B.
Reactor Pressure Control C.
Steam System Isolation and Pressure Control D.
Feedvater System Isolation and Control This categorization has been developed to focus upon those primary functions which have a potential for control system interaction.
The table identifies the range of equipment actions and actuation times used in the plant safety analysis for steam line break, feedwater line break and large and small LOCA.
2.
Plant Unicue Features A.
Steam Line Pressure Switches TMI-l has steam line pressure switches located inside containment that actuate to isolate feedvater on a Steam Line Break (SL3).
These pressure ssitches would be exposed to an adverse environment on a line break inside containment.
Failure of a single pressure switch will not prevent isolation of feedvater nor will failure of a single pressure switch cause it to incorrectly isolate feedwater.
The pressure switches have previously been evaluated as suitable for steam line bresk environment. This previous conclusion will be reverified.
B.
Feedwater Isolation System The feedwater isolation system for SLB was modified during the 1979 refueling outage to close the main and startup feedvater block valves on the affected steam generator in addition to the control valves.
The block valves are motor operated and therefore are unlikely to inadvertently reopen after. closing.
Although the v.alves may not be qualified for adverse environ =ent the redundancy provides additional assurance of feedwater isolation.
1203 355 2
C.
Fressuriser Heaters The pressurizer heater centrol and power supply could be exposed to adverse envircnnent cn a feedvater line break.
A feedvater line break cculd affect pressurizer heater operation, hcVever, this vould not adversely affect the course of the accident.
D.
Auxiliary Feedvater System T'e main steam lines and one feedvater line go through the Inter-mediate building.
The Auxiliary Feedvat er System is also located in the Intermediate Building.
The effects of adverse environ =ent on the Auxiliary Feedvater System have not been fully evaluated.
Althcugh there is sc=e distance separating the Auxiliary Feedvater cceponents and thenain feedvater and steam lines it is not possible to state at this time that Auxiliary Feedvater cc=pctents vould not be affected.
1203 356
III.
Safety Assessment 1.
Potential Envirot= ental Effects The non-safety grade control syste=s have been reviewed to determine if an accident enviren=ent could adversely affect the analyzed course of the event.
Specifically, the approach taken was to use the safety analysis function and par 1=eters frc= Table 1 as a basis to identify where potential control system effects could have an i= pact.
The result of this evaluation is st=uarized in Table II, Fotential Environ = ental Effects on Non-Safety Grade Control Systems.
The = atrix identifies, for six accident types, the non-safety grade control syste=s which could be~ adversely affected by the environ =ent caused by the event. '4here no entry is made in the = atrix, no potential for environ = ental effects exists due to the physical location of the equip =ent with respect to the high energy line break, i.e., brecks inside contain=ent do not affect equipment outside contain=ent and vice versa.
If 1n entry is =ade (X or Y), a potential effect exists as follevs :
X - The adverse environ =ent caused by the break could affect the equipment and, equipment =alfunction could affect safety analysis functions identified in Table I.
Y - The adverse environ =ent caused by the break could interact with the equipment, but the equip =ent =alfunction would not affect safety analysis functions identified in Table I.
This structuring of the potential effects = atrix provider a focus on those non-safety grade control syste=s which are i=portant and identifies areas for further evaluation of the impact on the safety analysis (i.e., X's).
2.
I:;act on Plant Safety Analysis Potential enviren= ental effects which could adversely impact the plant safety analysis are identified in Table II with an "X".
For each potential adverse effect, a safety assess =ent vill be prepared to ecnfir= plant safety or identify a potential problem area.
A.
Turbine 3nass /At=csrheric Felief Valves. 'G'4 Control and AF'4 Control Under Larze LOCA Environ =ent The large break loss-of-coolant accident relies upon safety grade equipment for =itigation.
The pctential effects presented in Tables I and II indicate that the control syste= functions, ; hough considered in the analysis, are =cdelled censervatively such that postulated =alfuncticcs of these syste=s vill not invalidate the analytical results.
The reactor shutdevn and pressure control during the blovdcvn and reficod phases do not rely upon ncn-safety grade centrol syste=s.
The stea= and feedvater syste: control features are conservatively =odelled in the analyses as folicus :
1203 357 4,
1.
The secondary steam system is conservatively assumed to remain intact (bottled up) to provide a large heat source during the late stages of blowdown.
The steam safety valves are assumed to maintain a concervately high steam pressure.
Fotential control syctem effects which provide steam relief would tend to improve the analytical res"Its.
2.
The feedwater system flow is conservatively assumed to quickly decrease to zero following the break. This loss of feedwater minimizes the effect of the OTSG secondary as a heat sink for a conservative analysis.
B.
MFW and AW Control and Turbine Syrass/ Atmospheric Relief Under Small LOCA Environment The small break loss-of-coolant analysis has been revised since TMI-2 to include a parameterization of potential equipment and operator actions during the accident.
As a result of this re-analysis, operating guidelines have been prepared by the NSSS vendor for use in operator training and revised operating procedures.
This change to the small break operating procedures provides a consistency between the small LOCA safety analysis and the requi. ed eqnipment and operator actions.
A review of Table II indicates a potential problem with the main or auxili ary feedwater level control.
The small break analysis and operating guidelines utilize CTSG level for RCS cooling and depressurication.
In the adverse environment caused by the small LCCA, the OTSG level indication could potentially be misleading to the operator and cause an inadequate amount of OTSG water inventory.
The effect of errors due to reference leg heating are being evaluated in response to II Bulletin 79-21.
Table II also indicates a potential interaction between small LOCA and Turbine Eypass and Atmospheric Felief Valve control.
The steam generator outlet pressure transmitters are 1ccated inside containment and thus could be exposed to an adverse environment. The transmitters were procurred with environmental qualifications.
It is expected that adverse interaction is unlikely; however, further investigation is planned.
1203 358 5
C.
Pressuriser PCRV under SLB (Inside Containment ), FWLE (Inside Cortainment ) end LOCA Environments The analysis and consequences of inadvertent opening or failure to close of the pressuriser PCEV as a result of SLB, FWLB or small LCCA enviren=ent vill be addressed in the "TMI-l Festart Safety Analysis Report" and vill be submitted prior to TMI-l rest art.
D.
CRDCS Under All Environments Potential for a significant increase in initial power level as a result of spurious rod withdrawal prior to reactor trip for SLB, FWLB and LOCA vill be addressed in the 'TMI-1 Restart Safety Analysis Report" and will be submitted prior to TMI-l restart.
E.
MFW and AF'J Control and AF'4 Isolation Valves Under SL2, ?4L3 ;
and Small LCCA Environments As indicated in Table 1, these control systems are i=portant elements of the safety analyses for steam and feedvater line breaks. These accidents vill be addressed in the "TMI-l Restart Safety Analysis Peport" and will be submitted prior to TMI-l restart.
Y Justification for Operation Based on the evaluations and safety assessments which vill be presented in the "TMI-l Restart Safety Analysis Report" it vill be demonstrated thro Igh conservatisms that safe operation of TMI-l vill te completely justified.
1203 359
V.
Recommended Future Action The "TMI-l Restart Safe y Analysis Report" vill demonstrate that TMI-l actual equipment actuation and performance are consistant with that used in the licensing basis analysis.
The report vill address potential effects of non-safety grade control systems in an adverse environment and assess conclusions reached in the original safety analysis. In addition, the report vill assess the environmental effects of equipment required to maintain safe shutdown following accidents which cause an adverse enviroc=ent.
This effort can be closely coupled to the Abnormal Transient Operating Guidelines Program currently under consideration, and can focus upon additional operator training to recognize and respond to the impact of an adverse environment on non-safety grade control syste=s.
The schedule for submittal of the Safety Assessment can be consistant with the current schedule for the Abnormal Transient Cperating Guide-lines Program (i.e., =id-lo80).
A more detailed evaluation of potential effects of high energy line break accidents on non-safety control systems vill be performed in the long term, with particular emphasis on the potential program areas identified in the safety assessment.
1203 360 1
sf),
TABLE I TYPICALEQULPMENTRESPONSEDURINGHIGI!ENERGYLINEBREAKS b'l CZ)
B&U 177 FA PLANTS cxa s
Steam Line
.Feeduater Large Small Break Line Break LOCA LOCA I.
Reactor Power Control and Shutdown Trip Function Utilized High 4 or Lov High RC Pressure Reactor Trip Low RC Pressure RC Pressure Not Used Time of Reactor Trip 1.1-8.0 sec.
8.2-13.4 sec.
II.
_ Reactor P.ressure Control Tire to PORV Actuation PORV Not 4-8 sec.
PORV Response PORV not acsumed
^' ""
- Time at which PORV Closes 420 sec.
U U I"P #'""'
"P""
Steam Line Break III.
Steam System Isolation and Pressure Control (1)
Steam Line Isolation Time 1.6-8,5 sec.
6.0-12.0 sec, Code Safety Code Safety Valves are Used Valves are Used (2)
Time to Steam Relief Valve Opening 7.0-16.0 sec.
7.0-7.5 sec.
in the Analysec in the Analyses (2)
Time for Steam Relief. Valve Closure 20-30 sec.
25-30 sec.
f r Conservatism for conservatism IV.
Feed nter System Isolation and Control 1
(1)
Main Facdwater Isolation Time l'9-34 sec.
N18 sec.
Analysis Con-Not Required
""[8tiV"lY (1)
Auxiliary Feedwater Isolation Time 19-34 sec. 418 sec.
not nequired
,g3 (2)
Auxiliary Feedwater Initistion Time 440 sec. 440 sec.
of All Feed-nAo nee.
(2)
Ma'n or Auxiliary Feedwater Control Maintain Maintain Maintain Prenet
""E"#
Minimum Minimum UTUG Ievel OTSG Level OTSG Level (1) Affected Steas Sencrator' (2) Unaffected Steam Generator
r r
l TABLE Il N
POTENTIAL ENVIRO:2! ENTAL EFFECTS ON NON-SAFETY CRADE CONTROL SYSTFJ1S.
c M
Licensing Basis Accidents SLB Inside SLB Outside FWLB Inside FVLB Outside Large Small C
Non-Safety Crade Control Systems Containment Containment Centainment Containment LOCA LdCA C
,g' N
I.
Reactor Po 'er Control and Shutdown X
Co,ntrol Rod Drive Control System X
X X
X X
II.
Reactor Pressure Control Power Operated Relief Valve X
X Y
X Pressurizer Heaters Y
Y Y
y Y
Y Y
Y Y
Y
~Y Pressurizer Spray III.
Ste.1n System Isolation and Prescure Control X
Turbine Trip / Turbine Stop Valves X'
X X
Steam Line Isolation Valves
- Turbine Bypass /Atm Relief Valves **
X X
X X
X X
IV. Int 4gater Systan 15011Licn_and Control Main Feedwater Con *co1**
X X
X X
X X
l Main Feedwater Isolation" X
X X
X Y
X X
X X
X Y
X Auxiliary Feedwater Isolation X
Auxiliary Feeduster Initiation **
X Auxiliary Feedwater Level Control ** -
X X
X X
X X
Environmental c.f fects Cannot occur Due to Location of Equipment Affected Steam Cenerator
- Unaffected Stents Generator (inside containment vs. outside containment)
Y Environment will not affect Safety Analysis Results*
X Environ:nent could affect Safety Analysis Results b
N 1
TMI-l & TMI-2 "bec" LIST
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Mrn. Pat lliccinn b
Edison'Riectric Inntitute Mr. W. R. Gibson h;
90 Park Avenue Babcock & Wilcox P.O. Box 1260 fiev York. 'tew Yr-t 10016 b!
Tynchburg, V*
?? 5( 3 Mr. E. L. Bl ake, Jr.
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Shaw, Pittman, Pottn & Trnwbridca Mr. T. F. Hartley, Jr.
1800 "M" Street, NW Marsh & McLennan, Inc.
y Wnchington, D.C.
?0036 1221 Avenue of the Americas How York, New York 10020 vt Mr. R. Sanaccre jt American Nuclear Innurcen Mr. A. S. Dam The Exchange - Suite 2'3 Burns & Roe, Inc.
270 Farmincton Avenue 650 Wintern Avenue Fnnnincton, CT n603?
Paramun, New Jersey 07652 O
l President's Commission on the 3
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Accident at Three Mile Ir. land Ms. Margaret Reilly Attn: Stnnley M. Gorinson Enq.
Chief Div. of Reactor Review
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2100 "M" Otrent TTW PA. Dept. of Environmental Resources Fulton Rank Buil. ding
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Washincton, D.C.
20037
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liarrisburg, PA 17120
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D SmW:
B. C. Rusche 6
S c'.:.L C ; ?
4 J. J. Rarton J. R. Stair W. N. Moreau L. W. IIardinc 9
W. Shmauss
/GRC Chairman TMI-l J. T. Collins (iTRC) Trailer h 7 H. Kar.anas E. G. Wallace
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R. F. Wilson J. F. Wilson J. P. Loran GEC Chairman - TMI-2 T. A. Mackey GRC Facretary - TMI-2 2
G. P. Miller T't! GORB Gecretary R. C. Arnold Chaircan - TMT-1 PORC D. G. Mitch911 Secretary - TMI-1 PORC C. A. Ulxdorf rhairman - TMI-2 PORC J. L. Seelirrer Cecretary - TMI-2 PORC R. M. Prabhakar Chairman - PORC
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Cecreta'ry - RORC C.
W.
Smyth 4
y (Mwa 7 l{
D. Haverkamp 3
GRC Secretary - TMI-l File $02.001
.0001.0001.02 Mr. Robert L. Rider i
i Bechtel Trailer 107A i203 363