ML19341C169
| ML19341C169 | |
| Person / Time | |
|---|---|
| Site: | Midland |
| Issue date: | 02/23/1981 |
| From: | Jackie Cook CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19341C170 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.E.1.1, TASK-TM 11223, NUDOCS 8103020210 | |
| Download: ML19341C169 (12) | |
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Vsce PressJent - Projects, Engsneenng
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!d Genered o'11ces: 1945 West Parnell Road, Jackson. MI 49201. ($17) 788-0453 r
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1 ui Harold R Denton, Director of Nuclear Reactor Regulation
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US Nuclear Regulatory Commission Washington, DC 20555 MIDLAND PROJECT MILLAND PLANT AFW RELIABILITY ANALYSIS ItIDLAND DOCKET N01 50-329, 50-330 FILE: 0917.1 UFI: 08*06*04*03 SERIAL:
11223
REFERENCE:
1.
NRC letter from D F Ross Jr to S H Howell dated 4-24-80.
2.
NUREG-0737 " Clarification of TMI action Plan Requirements,"
October 1980.
3.
CP Co letter S H Howell to H R Denton, Serial #7999, dated 11-30-79.
4.
CP Co letter S H Howell to H R Denton, Serial #8026, dated 12-4-79.
5.
CP Co letter S H Howell to H R Dentan, Serial #8563, dated 4-1-80.
Consumers Power is hereby submitting the Midland Plant Auxiliary Feedwater System Reliability Analysis prepared by Pickard, Lowe and Garrick, Inc. This report provides a complete response to Part 2 of Reference #1 and NUREG-0737, Item II. E.1.1 (1) and is referenced by Appendix 10A.2 of the Midland FSAR.
Also included is a preliminary response to Part 4 of Referc2ce #1.
This document is currently undergoing pre-submittal review and wit be submitted in the next amendment of the Midland FSAR in Appendix 10A.4.
However, it is included at this time for completeness since CP Co understands that sa NRC staff reviewer is available.
This submittal in addition to the information previously submitted in the Midland FSAR sections 7.3.3.2.6, 7.4.1.1.1, 7.4.2, 10.4.9 and Appendix 10A comprises a complete response to all of the AFW related requests for information contained in references #1 and 2.
We wish to emphasize the following points regarding the AFW Reliability Analysis:
1.
This analysis examines three alternative designs. The double crossover case represents the current Midland AFW System Design as described in the FSAR. The base case represents the Midland AFV System Design prior to incorporation of the modifications described previously in our responses to your 10 CFR 50.54 (f) requests (references V3, 4 and 3).
'a'e wish to I
oc0181-0217a100 810 sos jr/0 o
2 point out that results from this analysis were utilized in the AFW system design review discussed in references 3, 4 and 5.
The three pump case represents a theoretical comparison of the Midland 2-100% pump design with the 2-50%, 1-100% pump design utilized on several other B&W plants.
It does not represent an analysis of any actual plant design nor is the design feasible for Midland. The results shown on Table #3 (page 11) and summarized in Section 2 clearly demonstrate the lower unavailability of the Midland AFW System Design in comparison to the three pump design analyzed.
2.
Figures-2.(a), (b), (c) and Table 5 except for the Midland data points, were taken direct 13 from an AFW reliability analysis of other B&W plants performed by B&W (report reference #2). Some of this information, therefore, may-have been altered by recent modifications at these plants.
~ The "on demand" unavailability presented for Midland in these figures is Lconservative relative to the 5, 15, and 30 minute unavailabilities since no. credit is.taken for an Anticipatory Reactor Trip on Loss of Main Feedwater or-operator actions to initiate AFW.
3.
Section 6.1 of the report lists several recommendations that could potentially improve system availability beyond the current level. Ci Co has evaluated these items and has reached the following conclusions:
a.
-Reduce pump unavailability due to maintenance - The maintenance frequencies and durations used in this analysis were taken from WASH-1400 since CP.Co maintenance schedules and procedures were not available for input to the study. These maintenance schedules and procedures are currently. under development and every effort is being made to consolidate scheduled maintenance, minimize-the duration of maintenance outages and eliminate any non-essential maintenance-
'during power operation. CP Co estimates, however, that the WASH-1400 data will approximate the actual frequencies and durations of AFW pump maintenance at Midland.
-b.
Improve the reliability of the AFW turbine controls'- The control fluid to be used in the Midland AFW pump turbine drivers will, in accordance with the manufacturer's-recommeniations, contain a corrosion' inhibiter to minimize the potential effects af moisture con (<nsation in the control system. The control fluid level will be checked regularly and the~ fluid will be periodically changed to insure its:sood quality. To Consumers Power's knowledge the only
~ instance of a' condensation / corrosion related problem in turbine.
-controls has occurred where an incorrect control fluid was in use.
Therefore, CP Co believes that heating of the control fluid is not required.
Positive indication ~of AFW' pump turbine overspeed trips and reset is provided.at Midland by amber lights on the main ~ control board, the auxiliary-shutdown panels and the local AFWLturbine. control panels.
These; lights. illuminate to. indicate an.overspeed trip and extinguish when the trip linkage is reset.
.oc0181-0217a100-de' w~..
3 c.
Reduce the chance of human error during testing - The AFW pumps wili be tested every 31 days in accordance with Section 16.3/4.7.1.2 of the Midland Technical Specifications and the applicable codes and standards. To increase AFW system availability during pump testing, the Midland procedures for periodic surveillance testing of AFW will include provisions for operator communication with the control room any time an AFW train is removed from service for testing.
In addition, the position of the AFW full flow test valve will be indicated in the control room to permit regular verification of its correct positioning. Therefore, CP Co believes that adequate measures have been taken to reduce the probability of human errors during AFW system surveillance testing.
In conclusion, CP Co believes that this study in addition to the information previously submitted and incorporated into the Midland FSAR clearly demonstates the adequacy of the Midland AFW System Design. We request that the AFW System Design Review Board convene on April 22, 1981.
JWC/JPK/cir CC RJCook, Midland Resident Inspector oc0181-0217a100 4
QUALITY ASSURANCE PROGRAM 1h-SAR CHANGE NOTICE
- 1. Q 3 FSAR JOB NO.
7220
- 2. DISCIPLINE /COMPANYEP CO - 3&L
- 3. No.
- 4. ORIGINATOR M
- d*9t**-/M Fe2'ca--^u 5.DATE 1 27/9'
/
- 6. REFERENCED SECTIONS OF SAR Appendix 1CA.4 and 1CA.2 PRELIMIN ARY
- 7. DESCRIPTION OF CHANGE Respenses to the NRC requests for infcz :ation
- 8. REFERENCED SPECIFICATIONS OR DRAWINGS None
- 9. JUSTIFICATION DFRoss Jr's letter to SEowell dated 4-24-60
- 10. BECHTEL DISCIPLINE INTERFACE REVIEW:
INTERFACING STAFF REVIEW:
O ARCH C PLANT DSN O ARCH.
O WCH O CIVIL C PGAE OCML.
O NUCLEAR O CONTROL SYS O STRESS O CONTrb)LSYSTEM O PLNiTCSN O ELEC D OTHCR O ELEC.
O REUAEtuTY E MECH. NUCLEAR O GEOTECH O STRESS CM&CS G OTHER
- 11. REVIEWED BY DATE
- 12. REVIEWED BY DATE l 13. REV!EWEC BY DATE iGrouo Sc;: minor)
(EAR COCRDIN ATCm i i
tNUr. EAR ENGWE2m j
t-l 14.CO*4CUERENCE OY OATE i 1 E. APPROVED SY (C? Col IDATE
- 16. CONCURRENCE SY DATE iPriOJECT F.MN'IEo!
WC5's sif?PLIETD sa a nsers rce s :s
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n PR ELIMIN A R Y M im. 1u _,,,
10A.4 RESPONSES TO TI;E REQUESTS FOR INFORMATION RECARDINC TifE BASIS FOR ArW SYSTEMS FLOW REQW REMENTS TRANSMITTED IN ENCLOSURE 2 OF M. ROSS JR. LETTER TO S.II. IlOWELL, APRIL 24, 1980 Question
Response
Question 10A.4 10A.4.1 See Attached Response a.
Identify the plant transient and accident conditions considered in establishing AFWS flow requirements, including the following events:
1)
Loss of Main Feed (LMFW) 2)
LMrW w/ loss of offsite AC power 3)
LMFW w/ loss of onsite and offsite AC power 4)
Plant cooldown 30 5)
Turbine trip with and without bypass 6)
Main steam isolation valve closure 7)
Main feed line break 8)
Main steam Lane break 9)
Small break LOCA
- 10) Other transient or accident conditions not listed above b.
Deecribe the plant protection acceptance criteria and corresponding technical bases used for each initiating event identified above. The acceptance criteria should address See Attached Response e--
plant limits such aat Maximum RCS pressure (PORV or safety valve actuation)
Fuel temperature or damage limits (DNB, PCT, maximum fuel central 10A Revision 30 10/80 e
e
HIDLAND 1&2-FSAR
!92MJHEMMEHiind!!nr 9"F%3!PdFM " "" ^m PRELIMIN ARY
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ROSS JR. IATTER TO Lit. IlOWt:LL,3v4 T 24, 1980 (contin,ug )
Question
Response
temperature)
RCS cooling rate limit to avoid excessive coolant shrinkage Minimum steam generator level to i
assure sufficient steam generator heat transfer surface to remove decay heat and/or cooldown the primary system.
10A.4.2 Describe the analyses and assurptions and corresponding See Attached Response 30 technical justification used with plant condition considered in 1.a. above including l
a.
Maximum reactor power (including instrument error allowance) at the time of the initiating transient or accident.
b.
Time delay from initiating event to reactor trip.
c.
Flant parameter (s) which initiates AFWS flow and time delay between initiating event and introduction of AFWS flow into steam generator (s).
d.
Ministam steam generator water level when initiating event occurs.
j e.
Initial steam generator water inventory and depletion rate before and after AFWS flow commences - identify reactor decay heat rate used.
f.
Maximum pressure at which steam is released from generator (s) and agcinst which the AFW pump must develop sufficient head.
g.
Minimum number of steam generators that must 10A-14 Revision 30 10/80 se
-s.
PRELIMIN ARY MIDLAND IC-FSAR 10A.4 RESTONSES TO THE RE EESTS FOR INFORMATION REGARDING THE BASIS FOR AFW SYSTEMS FLOW REQUt hi>1ENTS TRANSMITTED IN f.NCLOSURE 2 OF D.F. RoSS JR. LETTER TO S.H. HOWELL. AFRil,24. 1380 (continuedj l
Question
Response
receive AFW flows e.g.1 out of 27, 2 out of
(
47 h.
Rr flow condition - continued operation of RC pumps or naturel circulation.
1.
Maximum AFW inlet temperature.
L j.
Following a postulated steam or feed line break, time delay assumed to isolate break and direct AFW flow to intact steam generator (s).
AFW pump flow capacity allowance to accommodate the time delay and maintain minimum steam generator water level. Also 30 identify credit taken for primary system heat i
removal due to blowdown.
k.
Volume and maximum temperature of water in main feed lines between steam generator (s) and AFWS connection to main feed line.
L 1.
Operating condition of steam generator normal blowdown following initiating event.
m.
Primary and secondary system water and metal sensible heat used for cooldown and AFW flow 4
sizing.
n.
Time at hot standby and time to cooldown RCS to RHR system cut in temperature to size AFW water source inventory.
10A.4.3 Verify that the AFW pumps in your plant will cupply the See Attached Response necessary flow to the steam generator (s) as determined by items 1 and 2 above considering a single failure.
Identify the margin in sizing the puap flow to allow for pump recirculation flow, seal leakage and pump wear.
r 4
4 IOA-15 Revision 30 10/80
RESPONSE TO QUESTION 10A.4.la The minimum Auxiliary Feedwater System (AFWS) flowrate was set by functional requirements. That flowrate was then verified to be acceptable using the transient which would require the greatest AFWS flow. The transients considered were analyzed for the FSAR and are identified in Table 1.
The events listed in this question which are not included in Table I will also be addressed.
The functional requirements for the AFWS flowrate are to remove the decay heat generated after a reactor shutdown and to provide a smooth reactor coolant flow transition from forced circulation to natural circulation should a loss of offsite power (LOOP) occur simultaneous with the need for AFW. The functional requirements resulted in an AFWS system flowrate of 720 gpm to be delivered to the steam generator within 40 seconds of the initiation signal.
The 40 seconds time was chosen to allow the AWS to inject feedwater and begin increasing steam generator level to the 50% operating range level, that required for natural circulation, prior to completion of the RC pump coastdown. At that time, the design flowrate was selected to be equal to or greater than the decay heat generation rate. Since decay heat rate changes with time, other values than 40 seconds and 720 gpm could have been used and been acceptable.
These parameters were then used in transient and accident evaluations. The loss of feedwater (LOFW) transient is the most limiting for AFWS flow. The analysis assumptions for this event are addressed in the response to Question 10.A.4.2.
All other transients which either require or assume the availability of AEW in the Safety Analysis use the design values derived from the functional requirements.
The events listed in Enclosure 2 of D F Ross, Jr letter to S H Howell, dated
-April 24, 1980 which are not included in Table 1, are discussed below:
T.MFW With Loss of Onsite and Offsite AC Power - This event was not a design Casis of plant and subsequently is not included in Chapter 15 of the FSAR.
However, following a temporary loss of all A-C power, the steam turbine driven AFW pump can be used to supply sufficient feedwater to both steam generators as discussed in FSAR Section 10.4.9.3.
Plant Cooldown - Plant cooldown with AFV is a controlled event with decay heat levels equal to or lower than in the emergency condition identified as the design basis event. The design basis event bounds this case for AFW flow required. Protection against potential AFW overcooling is provided by the AFW 1evel control system described in FSAR subsection 10.4.9.
Turbine Trip With and Without Bypass - This event does not affect the AFi!S unless MW fails in which case the loss of MFW event previously addressed would bound the AFWS design.
Main Steam Isolation Valve Closure - Again, this event does not directly affect the AWS unless MW is lost as discussed above.
-Small Break LOCA - The AFW criteria assumed for this event is described in Topical-Report BAW-10052 updated by letter report, J H Taylor.(B&W) to S A Varga (NRC), 7/18/78, and the recently submitted B&W report entitled
" Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 FA Plant", 5/07/77. These documents discuss and show that the AFWS flowrate will not lead to the violation of the acceptance criteria.
miO181-0205a100 t
P M L1/A!M ARY
2 PRELiMIN ARY 7-RESPONSE TO QUESTION 10A.4.lb:
The plant protection acceptance criteria for each accident are listed in Table 1 along with the technical basis for each acceptance criterion. The transient events identified in Question 1 which are not included in Table 1 are not bounding events, have not been analyzed, and as such do not have acceptance criteria. The acceptance criteria for a small break LOCA are included in the documents identified in Response to Question la.
The RCS cooling rate is not a limit relative to accident acceptance criteria.
The safety limit for all transients which use AFW for mitigation is that the core remain cooled with ultimate acceptance criteria being those addressed in Table 1.
For transients which result in draining the pressurizer or for which natural circulation is slowed or interrupted, restoration of pressurizer level and subcooling is accomplished by swelling due to core heat imput and inventory restoration by RPI.
Steam Generator level is not based on decay heat removal rate or cooldown capability. Steam generator level is variable depending on the plant condition. The level is normally low when removing decay heat with forced primary circulation. The level is normally high when removing decay heat with natural circulation.
It is also set high for small LOCA as described in Topical Report BAW-10052, and in the B&W report, " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks".
RESPONSE TO QUESTION 10A.4.2 As discussed in response to la above, the design basis event which verifies the AFWS design requirements is loss of main feedwater. The analysis assumptions for this event are listed below keyed to the letters of the question. Corresponding technical justification where not specifically listed below, is based on licensing requirements and prudent engineering judgment at the time of the analysis. The information is not provided for the other events identified in Question la and Table 1 because the LOFW event is the most limiting.
a.
Maximum Rx Power - 102% based on a 2% instrument error in neutron flux measurement.
b.
Time Delay Initiating Event to Rx Trip - The reactor will trip on high RCS pressure approximately 9 seconds after a LOFW event, AFWS Initiating Signal and Time Del'ay - The AFW initiation signal for the c.
LOFW event is a low steam generator level signal to the Auxiliary Feedwater Actuation System (AFWAS). The design basis time delay from initiation signal to full AFV flow into SG is 40 seconds. The FSAR LOFW analysis shcws that the time from the initiation event to full AFW flow into SG is 142 seconds.
d.
SG Level at Initiation Event - Steam Generator Inventory is dependent on power level.
miO181-02G5a100
I PRELIMIN ARY 3
I.
SG Inventory and Decay Heat - For discussion of water inventor /, see (d) e.
above. Reactor decay heat rate is based on one times ANS Standard 5.1 -
1979.
f.
Maximum SG Pressure - 1128 psia assuming actuation of only one primary safety valve.
g.
Minimum Number of SG - The number of generators was not specified in the analysis, heat removal capability is the pertinent parameter and can be accommodated by one SG.
h.
RC Flow Condition - Continued RC pump operation was assumed.
This inlet temperature could exceed 90 F under the unlikely situation where Service Water must be used as the AFW water source.
However, the AFW pump capacity was based on 105 F inlet water temperature with adequate margins for wear and seal leakage. As such, the present capacity is adequate to handle the additional volume required with hotter water. Also see response 10A.4.3.
J.
Steam, Feedline Break Time Delay - Refer to FSAR Section 15. 2.8 and FSAR Appendix 15c for feedwater line break analytical information. Refer to FSAR Section 15.1.5 and FSAR Appendix 15D for steam line break analytical information.
k.
Main Feedline Volume and Temperature Between SG and AFWS - N/A. The MFWS and AFWF are cross connected. This provision has been provided so that the AFWS may be used to supply feedwater to the steam generators during periods of start-up, cooldown and hot standby; however, these lines are isolated by AFWAS, and therefore this item has no bearing on the design basis.
1.
SG Normal Blowdown - N/A. The OTSG's do not have a blowdown system.
Water and Metal Sensible Heat Used - Plant cooldown was ngt considered in m.
the design basis analysis. A heat capacity of 1.256 x 10 BTU / F was used for calculating the volume of feedwater required to cool the RCS to decay heat system parameters.
Time at Hot Standby, Etc., Relative to AFW Inventory - The condensate n.
storage tank is sized to accommodate the plant at hot shutdown for at least four hours followed by a cooldown to 280 F.
The maximum cooldown rate of the RCS is 100 F per hour.
RESPONSE TO QUESTION 10A.4.3 The Midland AFW pumps are rated to supply 885 gpm with a total dynamic leac of 2700 feet of water. The pumps can supply the necessary flow of 720 gpm to the steam generators at a pressure of 1050 psig (2426 feet of water) with adequate margins for seal leakage and pump wear. Since the pumps are not on continuous miO181-0205a100
.a
}
b i
PRELIMIN ARY 4
recirculation, no margin was provided for recirculation flow on the sizing of the pumps. Feedwater demand during any of the plant transient and accident conditions discussed in item 1 is met by the AFW pumps in conjunction with the i
AFW level control described in FSAR Subsection 10.4.9.
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miO181-0205a100
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TABLE 1
.s ACCIDE::T tr.SC:tIPTto::
FSAR SECTTON ACCEPTA?CE CP.IT '<I.\\(I}
,1) Loss of Coolant Flow 15.3
'A, B
~
- 2) Loss of Electric Powur 15.2.6 A, B, D
- 3) Steam Line Break 15.1.5 D
- 4) Main Feedwater Break 15.2.8 A, D
- 5) Start-Up Accident 15.4.1 A, B
- 6) Rod Withdrawal Accident at 15.4.2
' A, B Rated Power Operation
- 7) Moderator Dilution Aceddent 15.4.6 A, B C) Cold Water Accident 15.4.4 A, B
- 9) Stuck-out, Stuck-In, or 15.4.3.1, 15.4.3.2 A, B Dropped Control Rod Accident
- 10) Steam Generator Tube Failure 15.6.3 B, D
- 11) Rod Ejection Accident 15.4.8 C, D
- 12) Loss of Coolant Accident 15.6.5 D, E
- 13) Loss of Main Feedwater 15.2.7 A, B NOTE:
(1)
_ KEY
_ ACCEPTANCE CRITERIA TEC11NICAL BASIS A
Mcx. RCS Press. - 110% Design ASME Code B
DNB > 1.3 with BAW-2 SRP 4.4 C
280 Cal./ Gram Fuel Limit Reg. Guide 1.77 D
Acceptable Doses 10 CFR 100 E
Fuel Cladding < 2200 F 10 CFR 50.46 6
PRELIMIN ARY G
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