ML20012D029
| ML20012D029 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 03/19/1990 |
| From: | Berry K CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| REF-GTECI-A-47, REF-GTECI-SY, RTR-NUREG-1218, TASK-A-47, TASK-OR GL-89-19, NUDOCS 9003260335 | |
| Download: ML20012D029 (9) | |
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.o-i Consumers Power xwm Derector i
MM bclear Ekeruiq MENCEANT PROGRE55 General Offwes: 1946 West Pernell Road. Jackson. MI 49201 e (617) 7881636 l
March 19, 1990 i
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1 Nuclear Regulatory Commission Document Control Desk l
Washington, DC 20555 DOCKET 50-155 - LICENSE DPR BIG ROCK POINT PLANT -
RESPONSE TO GENERIC LETTER 89 REQUEST FOR ACTION RELATED TO RESOLUTION OF UNRESOLVED SAFETY ISSUE A-47 Generi.: Letter 89-19, Safety Implications of Control Systems, requests that all BVR plente provide autometic vessel overfill protection, and that plant precedures and technical specifications for all plants should include provisf ons to verify periodically that the protection is available to mitigate main feedwater overfeed events during powet operation. In addition, the Staff reco:mnouds that all BWR recipient.s reassess and modify their operating procedures and opereter training, if necessary, to assure that operacors can mitigste reactor overfill events that may occur via the condensate pumps during reduced system pressure operation. The letter recommen6s that plants without automatic overfiM protection, such as Big Rock Point. have a design to prevent reactor overfill, or provide justification why overfill protection should not be included. Consumers Power Company % response for Big Rock Point is provided herein.
In addressing the concerns of the f,eneric lette: which specifically relate to those severe events that could potentially lead to r. steam line break, it is important to review the history of the primtry system overfill issue for Big Rock Point (BRP).
On February 1,1978, the Staff issued a letter entitled,
" Evaluation of Incidents of Primary Coolant Release From Operating Boiling Water Reactors." The report referenced in this letter describes eight incidents involving unintentional discharge of primary coolant through safety relief valves during power operation. Based on this document, a Staff recommendation was made that the BWR feedwater systems should be designed to automatically control vessel level during anticipated transients without flooding the main steam line or the lines to other safety related equipment.
The memorandum also stated that although BWR-3 and BWR-4 had satisfied the intent of the Staff's recommendations through inclusion of a feed pump trip on high level, many older plants gG326033590o339 pR ADOCK 05000155 g.
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Big Rock Point Plant Response to GL 89-19 March 19, 1990 l
t had not included this feature. The correspondence then requested the plant to
. notify the Commission within 30 days as to its intentions and schedule for Installation of such protective features, t
i By letter dated March 7, 1978, Consumers Power concluded that installation of a feedwater pump trip based on high water level at Big. Rock Point was unwarrant-l ed.
The basis for this conclusion was three-foldt
.2.
The existence of a steam drum at BRP and the large volume it affords.
1 3.
The requirement for the feedwater system to mitigate specific LOCA conditions, i.e. a trip circuit would decrease availability.
t The reply also described the operation, including the operator response time I
to failures, and the 15-year history of the feedwater system at BRP.
By letter dated December 1, 1978, the Commission concluded that the instal-1ation of a feedwater pump trip on high reactor water level at Big Rock Point was not necessary to assure safe operation of the plant. This safety evaluation-was based on the following two points:
1.
The maxir..um water icvel attained should not initiate isolation of any safety feature such as high pros <are coo.1 ant injecti n system, or disable any system or component required for thu erderly shutdown of the reactor.
and 3.
The minimum level attained should not require the a:tivation of any ectety system.
'Because the primary feedwater cystem at Big Rock Point is required to perform the high pressure coole:nt injection function during ecrtnin LOCA ovente, the 5
Staff concl9ded that by not eripping the feedwater pump on recetor high level.
the lineensee is consistent with performance objective (1) abovu. Furthermore.
the Ftaf f concluded that CPCo's decisior, not to install a feedsnt(c pump trip is acceptable.
All'of the elements contained in the March 7, 1978 letter to the Commission
- remain valid with the following exceptionst 1.
The'feedwater system has operated for 27 years with no known problems relating to inadvertent flooding of the primary system steam drum while at power, with the one exception noted in the December 1, 1978 SER, and 1
2.
The second steam drum high level alarm setpoint has been increased from 13 L
' inches to 20.8 inches above steam drum centerline. The reason for this setpoint change was that the 13-inch alarm was distracting the operator during normal plant start-ups and shutdowns. The first high level alarm setpoint' remains unchanged at +4 inches above steam drum centerline.
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N'c10 r.R gulct;ry Commission 3
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Big Rock Point Plant Response to GL 89-19 l
. March 19, 1990
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i As a part of the Systematic Evaluation Plan (SEP),' questions were asked with
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respect to primary system overfill events.
In response to these questions, CPCo created an Excessive.Feedwater Event Tree (Attachment 1) as part of the
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BRP Probabilistic Risk' Analysis (PRA). The quantification of this' event tree
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concluded that there were no significant core damage sequences. All sequences t
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were calculated to be less than IE-7/ year.
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.t Comparison of the Big Rock Point Excessive Feedwater Event Tree results to that of NUREG-1218 indicates differences in initiating frequency.. operator-
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error probability, and steamline break probability.. The initiating frequency l
'i used was based on BRP plant-specific data; a sensitivity analysis was perform-
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ed on the human error analysis; and the main steamline was not assumed to fail r
- as a result of overfilling. The main steamline'at BRP is 12-inch schedule 120 l
pipe and is filled with water for hydrostatic testing.
j Based on the justification provided above, which demonstrates the small contribution that primary system overfill has on the Big Rock Point core j
damage frequency and the fact that the plant has 27 years of operating exper-ience with no overfill events while at power, Consumers Power Company concurs with the conclusions of NUREG-1218, that implementation of overfill l ~
protection at Big. Rock Point is not warranted.
sf Q c'Y'Y
-Kenneth W Berry l
Director, Nuclear Licenning c
CC Administrator, Region I!1, USNRC
.s NRC Resident Inspector - Dig Rock Point
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-Attachment OCO390-0003-NLO2 1
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i CONSUMERS POWER COMPANY F
p Big Rock ^ Point Plant L
Docket 50-155 License DPR-6 L
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Response to Generic Letter No 89-19 dated September 20, 1989
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- At the request of the Commission and pursuant to the Atomic Energy Act of 1954 and the Energy Reorganization Act of 1974, as amended, and the Commission's Rules and.Regufations thereunder, Consumers Power Company submits our response to NRC letter dated September 20, 1989, entitled, " Request for Action Related to Resolution of Unresolved-Safety Issue A-47 " Safety Implication of Control Systems in LWR Nucicar Power Plants" Pursuant to 10 CFR 50.54(f) - Ceneric Letter 89-19".
Consumers Power Company's response in dated March 19, 1990.
- CONSUMERS POWER COMPANY l
'E ToLthe. best.cf my knowledge, -information and belief, the cor. tents of this
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' submittal are truthful and complate.
t By
$4--- --- -
David P Hoffman Vice N sident Nuclear Operations q?
9' Eworn:And subscribed to before me this 19th dat o' March, 1990.
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Elaine E Buehrer,-Notary Public Jackson County, Michigan My commission expires October. 11, 1993 i
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ATTACHMENT-l'
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's EXECESSIVE FEEDWATER EVENT TREE N
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Excessive reedwater Event Tree The event tree considered here is any fault or failure in the main feedwater 6
system or steam drum level control system which results in excessive feedwater p
flow. Like other plant transient events, this initiating event will eventually lead to reactor and turbine trip and subsequent demands on the plant shutdown systems. Thus, this event was considered under the class of initiating events L
entitled, Turbine Trip, in the Big Rock Point PRA. Because the possibility exists for relief valve operation under liquid or two-phase flow conditions (note the BRP safety relief valves were not designed for liquid relief), the potential for a LOCA due to a stuck open relief valve is higher for this transient initiator than for other events.
i In order to evaluate the importance of the excessive feedwater event as a possible LOCA, an event tree (Figure 1) was constructed. The event tree is i
very similar, and the potential consequences are nearly identical to the event tree for a small steam line break inside containment. A description of each branch follows.
Event Tree Description Branch Point 1: At Branch Point 1, excessive feedwater is being delivered to tha primary steam Jrur ard the' operator either manually controls.ieedwater or 4
fails to do so.
The operator would first become aware of the excessive leod-l wcter addition when the steam drum icvel reached the.first high water level The initial hi h water level snuunciator is annunciator setpoint (+4").
F backed up'by a second conplately diverse annuncIcter un high water level
(+ 20"). The firnt high water. level alarm uses the rame Ic. vel instruments ae the feedwater regulating controls and is, therefore, suscepcible to the same i
failuren that caused the event. Itowever, the second high 4#ter level alate in
't represtated by four separate annunciators which are supplied by tho steam drt.m level inst ruments associated with - the reactor depresserization system.
The RDS level instrumants tre completely diverse from the feedwater regulating system level control.
Failure to control feedwater at Branch Poict 1 Jerds to i
a reactor trip.
Branch Point 2:
Success at this point leads to successful core shutdown.
Failure to scram the reactor at this point Icada to ATWS condition which is addressed separately in the BRP PRA.
Branch Point 3: With feedwater delivery continuing at a high rate, the primary system will fill until one or more of the primary safety valves open to control pressure at 1600 psia. At least two of six primary safety valves are required to open to prevent vessel rupture and core damage (between one and two valves are required to pass liquid at the rate of maximum feedwater addition).
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Branch point 4: With the turbine tripped, the feedwater pumps will eventually trip due to loss of hotwell inventory, unless the operator intentionally trips the pumps first. In either case the sufety valves close. With the valves closed..the operator may proceed to cold shutdown using the available plant systems. This sequence is included in the turbine trip event tree. Failure
-of the valves to close results in a small steam line break.
F With a break in the primary system coolant boundary it is Branch point 5:
necessary to provide adequate inventory makeup. The reactor depressurization r
and core spray systems provide this function. The RDS causes rapid system blowdown when low level is reached in the reactor._ This permits delivery-of water for core cooling and inventory maintenance from the low pressure core spray system. Failure of the RDS or core spray system leads to core damage.
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Branch point 6: In this sequence, fluid from the primary system is continually being pumped intol the containment and the core is being cooled by the RDS l
working together.with the Core Spray System.
If containment water level rises to the 587 foot elevation, the operator is instructed to switch to the core spray recirculation mode to ensure containment and core integrity. Failure to switch to recirculation may lead to containment structural failure due to overfilling with water.
r Summarv-- Excessive Feedwater Event i
There are ne signtficant core damage sequences for this f. vent as all sequences araLless than IF-7/ycur. This is becauce of the low expectad frequency of e
liquid relief through the primary xelief valves during pcwer opcration (1.4E-6/yr).
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?t esticy _cf Core D mege (vr'I)
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recpen to high level altre as-i suming two control operators and moderate dependence (0)
.14E-4 Safety valves fail to open (J) 2.2E-3 Tex 0KL 1.E-8 Safety valve fail to close (K) 1.0 Failure of Post Incident System assuming 1 month mission time 1.1E-2 L
l Tex 0KC 5.6E-9 Failure of RDS/ Core Spray (C) 4E-3 t
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Excessive Feed To Steam Drum Excessive Operator RPS SRV SRV Enff Long Tern Feed to Controls Opens Clores Core.
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6' TexOJ AYm?S (0) Operator fails to control feedvster; (A) Reactor Protection System failure; - (J) Safety valves fail to open; (K) Safety valve-fails..to close; -(C) Failure cf seactor depressurization and core spray; (L) Long term cooling failure; Tex - Transient initiated by excessive feedwater.
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