ML19271A172
Text
Distribution: Docket File NRR RDG
~
JJjif 5 G80 RSB RDG Newberry Chron.
Mt No. 50-3382_
MEMORANDUM FOR:
R. Vollmer Director, TMI Support THRU:
G. R. Mazetis, Section Leader, Reactor Systems Branch, DSS FRCH:
S. F. Newberry, Peactor Systems Branch, DSS
SUBJECT:
TMI-1 RESTART REVIEW: EMERGENCY FEEDWATER SYSTEM CAPACITY AND ANTICIPATORY REACTOR TRIP Two of the Bulletins and Orders Task Force requirements for THI-1 and the other B&W reactors are the upgrade of the Erergency (Auxiliary) Feedwater System reliability and the installation of Anticipatory Reactor Trips on loss of feedwater and turbine trip. This memcra.ndum is intended to dccu-ment the status of our review of certain portions of these two areas on TMI-1 and to recocoend respective staff positions. The discussion regarding Erergency Feedwater is specifically related to the sizing or capacity design basis of the system. The attached enclosures address each of these areas.
We intend to proceed as indicated unless otheniise directed.
Scott F. "cwberry Reactor Systens Branch Division of Systems Safety
Enclosures:
Eccrgency Feedwater Sizing Calculation Anticipatory Reactor Trip on Loss of Feedwater 3.
Oconee loss of Feedwater Data 4.
TMI-1 Loss of Feedwater Data 5.
THI-1 Reactor Trip Parameters cc:
D. Ross J. Voglewede Z. Rosztoczy T. Novak H. Silver P. Norian S. Israel J. Wermiel D. Thatcher P. Matthews R. Fitzpatrick G. Mazetis M. Rubine
Contact:
Scott Newberry, NRR Ext. 27341 0
8002270 0,9 1 DSS:RSB f0SS:R.SB 0,,,ct sum me SNewber f
G
)jk)qp 1j J M -....---
~~-
D A T C m-N NRC form 318 A 14 79) NRCM O2040
++..**w====*=v**'aa****'":
g
Emergency Feedwater Sizing Calculation Our review and evaluation of the TMI-l EFW System must determine the bounding Emergency Feedwater System flow requirement for all transients and accidents (LCCA and non-LOCA) and the capability of the system to deliver the necessary water to the steam generators. The minimum acceptable requirement will appear in the TMI-l Technical Specifications.
Our current understanding with Met-Ed is that they will provide the staff the following:
1.
List of all events needing EFW to mitigate the consequences.
2.
Justification that the bounding non-LOCA calculation will serve as a conservative basis for sizing the EFW system for non-LOCA core cooling considerations.
In other words, they must show that the calculation they submit will bound all of the non-LOCA events requiring EFW.
3.
The non-LOCA bounding event will be a loss of feedwater using FSAR type assumptions to maximize heat removal requirements (1.2 ANS decay neat, 2% power level measurement uncertainty, RCP heat input).
The calculation will not take credit for " anticipatory reactor trip" since it will not occur under all conditions (see Enclosure 2).
The analytical method utilized will be CADOS.
The acceptance criteria for the event will be:
1.
Reactor Coolant System pressure remains less than 110% of design pressure (2750 psig).
2.
No fuel failure (ONBR >1.30).
This transient calculation would serve as the EFW sizing basis for all non-LOCA events (assuming our review finds no problems).
It should be noted that for this conservative sizing calculation, the above acceptance criteria do not preclude lifting the PORV, although credit for pressure relief through the valve will not be assumed. The basis for small LOCA events would still be the generic
2 calculations in reference I which assume 550 gpm with a 20 minute time delay for initiation.
Met-Ed expects to show that 500 gpm is adequate in tneir non-LOCA sizing calculation. Therefore, the flow requirement which would apcear in the Technical Specifications is 550 gpm.
The above approach to demonstrate adequate n:ing of the EFW system is a reasonable basis for allowing restart of TMI-1.
Anticioatory Reactor Trio System ( ARTS)
The ARTS were required by the Bulletins and Orders to help reduce the frequency of challenges to the PORY (in addition to lowering the high pressure trip setpoint and raising the PCRV setpoint). Most of the current generic ART modifications generate a loss of main feedwater reactor trip on low main feed pump control oil pressure on both pumps (which is what trips the main feed pumps).
It appears that the main feed pumps could be expected to trip on most of the loss of feedwater events that have occurred to date, including the TMI-2 accident, however, we have recogni:ed that they would not necessarily trip for all possible loss of tu.uwater scenarios.
For examole, review of the Oconee loss of feedwater data (reference 2) shows that if the ARTS nodification had been installed at the Oconee units,, n ART would have been initiated for approximately 65" of the reported loss of feedwater events. A suamary of the Oconee data is in Enclosure 3.
The change of the PORV and reactor trip setpoints (245C asig and 2300 psig respectively) are intended to reduce the frequency of opening the PORV for anticipated transients.
The bases for this approach are the best estimate calculations in reference 1 and the operating history of B&W plants since the changes were made (no PORV openings as a result of transients).
The ARTS were required to increase the margin attained by the revised setpoints (above).
Based on the similarity of the Oconee units to TMI-1, the margin available even without ART, and the design, procedural and operator training modifications required to facilitate recognition and mitigation of a stuck open PORV (should it open), the proposed ART parameters of MFP control oil pressure and turbine trip are reasonable to meet the short-term Bulletin and Order requirements.
1.
This similarity is based on the assumption tnat Met-Ed installs nain feedwater and condensate low suct on pressure pump trips similar to i
tnose at the Oconee plants.
. It is also interesting to note that of the two loss of feedwater events that have occurred in the operating history of TMI-1, it is not obvious that either event would have initiated on ART (reference 3). This is certainly true for the loss of instrument air event, but is not completely clear for the second event due to insufficient data.
(see Enclosure 4)
In the longer term, we recommend that a more complete generic study be conducted for all B&W plants to evaluate the need for a more " encompassing" reactor trip which is diverse to the high pressure trip for loss of heat sink events. While the 177 FA plants do have a high reactor outlet temperature reactor trip, we note that ARTS is the only time that secondary system raactor protection parameters are utilized (see Enclosure 5). A promising direction perhaps is a reactor trip based upon mismatch of the primary side (power) with the secondary side (feedwater).
Such a tr7p would probably cover all the Oconee loss of FW events, yet not compromise routine operations maneuvers (load changes, etc). Alternately, a more selective combination of secondary side parameters may provide a "better" coverage of pressure transien ts.
An additional consideration is the current SRP requirement of section 5.2.2, Overpressurization Protection, which gives no credit for the first direct reactor trip which would be high pressure. B&W plants should be evaluated considering this requirement.
In summary, the safety-grade ART inputs proposed by the licensee are acceptable for TMI-l restart; however, the need for a longer tenn generic study should be determined for all B&W plants concerning an alternate or additional diverse reactor trip.
Oconee 1, 2, 3 Data MFP Trip Prior Unit Reactor Trios Trip Due to LOFy to Reactor Trio 1
72 16 12 2
44 12 8
3 34 8
4 I
Total 150 36 24 Notes:
1.
21 of 24 MFP trips above 10% reactor power (10% reactor power is the currently proposed bypass setpoint for the loss of feedwater anticipatory reactor trip)
MFP Trip Prior to Reactor Power M)
No. of Events Reactor Trio Percent 0-10 3
3 100 11-20 10 6
60 21-50 12 10 83 51-75 5
4 80 I
76-100 6
1 (possibly 2) 17 Notes:
1.
1 (possibly 2) of 6 would nave generated ART; otners caused by ICS malfunctions which "ran-back" main feedwater; therefore, no MFP trip or ART would have occurred.
TMI-l Loss of Feedwater Data Date Power Level Cause/Descriction 6/18/74 7%
Loss of Instrument Air - Caused main feed control valves to shut.
Operator tripped reactor manually.
7/13/74 15%
High differential pressure across strainers on the suction side of C Condensate Booster Pump. Pump burned up due to low suction pres}ure.
Reactor tripped on variable pressure /
temperature reactor trip.
(Status of main feed pumps is not clear.)
TAllIE 2.3-1 itFAC'10lt PitOTECTION SYSTEM Tit 1P SLTPING LIMITS One iteactor Coolant hunp Four heactor Loolant Pumps '1hree Ileactor Coolant Pumpu Operatind in 10sch Loop
- .eus
- r Operating (Nominal Operating (Nominal (Nominal Operating
.ty p Operating Power - 100%)
Operatind Power - 75%)
Power IH)J) m
- 3 1.
Nuclear lower, Max.
I
% of ruted icwer 105 5 105.5 105 5
$.0 0
2.
Nuclear lover based 1.08 times flow minus 1.08 times flow minus 1.08 times riow minu: I t,* p:.
t 5*
on flow (2) and imbal-reduction due to reduction due to reduction due to ance, max. of rated imbalance (s) imbalance (s) imbalance (s) n puwer M
P gr-
{
3.
Nuclear gover based (5} NA NA 91%
/
on pump monitors, max.
y
% of rated lower E
p 7
4.
liigh reactor crolant 2355 2355 2355 17W g
c',
system pressure, psig, m
max.
5 low reactor coolant 1800 1800 IfiOO Iyr i
2 system pressure, psig.
n, min.
r*
'(756)(1)
(16.25 Tout - 7756)(l)
(16.25 Tout - 7756)(1,I l yg-6.
Variable low reactor (16.25 Tout. -
coolant system pressure, pulg, min.
7 Iteactor coolant temp.
619 619 619 4 F' F.,liax.
8.
liigh heaet.or liuilding la le 4
I pressure, psig, max.
(1) Tc,ut is in dedrees Fahrenheit ( F)
(2) heuctor coolant, system flow, %
(3) Administratively controlled reduction set only during reactor shutdown (Is) Automat.ically set when other segments of the ItPS (as specified) are bypassed (5) 'lhe pump monitors also produce a trip on:
(a) loss of two reactor coolant pumps in one reactor coolant loop, and (b) loss of one or two reactor coolant pumps during twu-pwnp operation.
References 1.
Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plants, May 7, 1979.
2.
Data on Oconee Loss of Feedwater Events - submitted informally to NRC staff by R. Wright (B&W), December 14, 1979.
3.
Reactor Trip Summary for TMI-l - submitted informally by Met-Ed during TMI-l restart review as part of their review of operating experience in accordance with IE Bulletin 79-05A.