ML19341C167
| ML19341C167 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 12/22/1980 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Linder F DAIRYLAND POWER COOPERATIVE |
| References | |
| TASK-05-10.B, TASK-05-11.A, TASK-05-11.B, TASK-07-03, TASK-09-03, TASK-5-10.B, TASK-RR NUDOCS 8103020178 | |
| Download: ML19341C167 (50) | |
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December 22, 1980
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Docket No. 50-409
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.Mr. Frank Linder
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General Manager Dairyland Power Cooperative 2615 East Avenue South La Crosse, Wisconsin 54601
Dear Mr. Linder:
RE: SEP TOPICS V-10.B. V.ll. A, V-11.B. VII-3 and IX-3 (Safe Shutdown Systems) - LA CROSSE BOILING WATER REACTOR (LACBWR)
Enclosed is a copy of our current evaluation of Safe Shutdown Systems for'LACBWR. This assessment compares your facility, as described in Decket No. 50-409 with the criteria current'y used by the regulatory staff for licensing new f acilities.
Please inform us if your as-built facility differs from the licensing basis assumed in our assessment within 90 days of receipt of this letter.
We note that you have proposed an alternate dedicated shutdown system as a part of our review of the liquefaction potential at the site.
Our evaluation of that system has not be en completed. We will update this report as E.oon as our review is finished. Nevertheless, you should proceed with your review of this report.
This evaluation will be a basic input to the integr ted safety assessment I
for your facility unless you identify changes needeG to reflect the as-built conditions at your facility.
This assessment may be revised in the future if your facility design is changed or if NRC criteria relating to this subject is modified before the integrated assessment is completed.
l l
Sincerely, Dennis M. Crutchfield, ief Operating Reactors Branch #5 Division of Licensing Encl oi;ure:
Cer.pleted SEP Topics -
Safe Shutdown Systems cc w/ enclosure:
See next page 810302 0/Y
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Mr.' Frank Linder December 22, 1980 cc Fritz Schubert, Esquire Director, Standards and Criteria Staff Attorney Division Dairyland Power Cooperative Office of Radiation Programs 2615 East Avenue South (ANR-460)
La Crosse, Wisconsin 54601 U. S. Environmental Protection Agency
- 0. S. Heistand, Jr., Esquire Washington, D. C.
20460 Morgan, Lewis & Bockius 1800 M Street, N. W.
Washington, D. C.
20036 U. S. Environmental Protection Agency Federal Activities Branch Mr. R. E. Shimshak Region V Office La Crosse Boiling Water Reactor Dairyland Power Cooperative ATTN: EIS C0ORDINATOR P. O. Box 135 230 South Dearborn Street Genoa,' Wisconsin 54632 Chicago, Illinois 60604 Coulee Region Energy Coalition Charles Bechhoefer, Esq., Chairman ATTN: George R. Nygaard Atomic Safety and Licensing Board P. O. Box 1583 U. S. Nuclear Regulatory Comission La Crosse, Wisconsin 54601 Washington, D. C.
20555 La Crosse Public Library Dr. George C. Anderson 800 Main Street Department-t.T Oceanography La Crosse, Wisconsin 54601 University of Washington Seattle, Washington 98195 U. S.- Nuclear Regulatory Comission Resident Inspectors Office Mr.' Ralph S. Decker Rural Route #1, Box 225 Route 4, Box 1900 Genoa, Wiscons.in.54632 Cambridge, Maryland 21613 Town Chairman Dr. Lawrence R. Quarles Town of Genoa Kendal at Longwood, Apt. 51 Route 1 Kenneth' Square, Pennsylvania 19348 Genoa, Wisconsin 54632 Thomas S. Moore Chairman, Public Service Comission Atomic Safety and Licensing Appeal Board of Wisconsin U. S. Nuclear Regulatory Comission Hill Farms State Office Building Washingt'on, D. C.
20555 Madison, Wisconsin 53702 Ms. Anne K. Morse Alan S. Rosenthal, Esq., Chairman Coulee Region Energy Coalition Atomic Safety and Licer. sing Appeal Board Post Office Box 1583 U. S. Nuclear Regulatory Comission
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Lacrosse, Wisconsin 54601 Washington,' D. C.
20555 U. S. ' Nuclear Regulatory Comission Mr. Frederick Milton Olsen, III Resident Inspectors Office 609 North lith Street Rural Route #1, Box 225 Lacrosse, Wisconsin Genoa, Wisconsin 54632 4
s i
SEP REVIEW 0F SAFE SHUTDOWN SYSTEMS FOR THE LA CROSSE BOILING WATER REACTOR i
Date: December 22, 1980 i
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i TABLE OF CONTENTS P_ age 1.0 INTR 00VCTION...........................................
1 2.0 GENERAL DISCUSSION.....................................
6 2.1 Normal Plant Shutdown and Cooldown................
6 2.2 Shutdown and Cooldown with Loss of Offiste Power..
7 3.0 CONFORMANCE WITH BRANCH TECHNICAL POSITION 5-1 FUNCTIONAL REQUIREMENTS................................
11 12 3.1 Background........................................
3.2 Shutdown Systems..................................
15 Table 3.1 Classification of Shutdown Systems..........
30 4.0 SPECIFIC RESIDUAL HEAT REMOVAL AND OTHER REQUIREMENTS OF BRANCH TECHNICAL POSITION 5-1.......................
35 4.1 RHR System Isolation Requirements.................
35 4.2 Pressure Relief Requirements......................
36 4.3 Pump Protection Requirements......................
37 4.4 Test Requirements.................................
38 38 4.5 Operational Procedures............................
5.0 RESOLUTION OF SEP T0 PICS...............................
40 5.1 Topic V-10.B RHR System Reliability...............
40 5.2 Topic V-11.A Requirements for Isolation of 41 High and Low Pressure Systems...................
5.3 Topic V.11.8 RHR Interlock Requirements...........
41 5.4 Topic VII-3 Systems Required for Safe Shutdown....
42 5.5 Topic X Auxiliary Feed System (AFS)...............
45 45
6.0 REFERENCES
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1.0 INTRODUCTION
The Systematic Evaluation Program (SEP) review of the " safe shutdown" subject encompassed all or parts of the following SEP topics, which are among those identified in the November 25, 1977 NRC Office of Nuclear Reactor Regulation document entitled " Report on the System-atic Evaluation of Operating Facilities":
1.
Residual Heat Removal System Reliability (Topic V-10.8) 2.
Requirements for Isolation of High and Low Pressure Systems (Topic V-11.A) 3.
RHR Interlock Requirements (Topic V-11.B) 4.
Systems Required for Safe Shutdown (Topic VII-3) 5.
Statir
~ 2rvice and Cooling Water Systems (Topic IX-3) 6.
Auxiliary Feedwater System (Topic X)
The review was primarily performed during an onsite visit by a team of SEP personnel.
This onsite effort, which was performed during the period May 22-24, 1978, afforded the team the opportunity to obtain current information and to examine the applicable equipment and procedures,-and it also gave the licensee,-Dairyland Power Cooperative, the opportunity to provide input into the review.
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. The review included specific system and equipment requirements for remaining in a hot shutdown condition (> 212 F) and for proceeding to a cold shutdown (< 212*F).
The review for transition from operating to hot standby considered the requirement that the capability exists to perform this operation from outside the control room.
The review was augmented as necessary to assure resolution of the applicable topics, except as noted below:
Topic V-11.A (Requirements for Isolation of High and Low Pressure Systems) was examined only for applica'. ion to the Residual Heat Removal (RHR) system.
Other high pressure / low pressure interfaces were not investigated.
-Topic VII-3 (Systems Required for Safe Shutdown) was completed except for determination of design adequacy of the systems.
Topic IX-3 (Station Service and Cooling Water Systems) was only reviewed to consider redundancy and to a limited extent seismic and quality classification of cooling water systems that are vital to the performance of safe-shutdown' system components.
(No discussion of Topic IX-3 is included ia this report.
The information gathered in support of this-topic will be used to resolve the topic later in the SEP.)
. The criteria against which the safe shutdown systems and components were compared in this review are taken from the:
Standard Review Plan (SRP) 5.4.7, " Residual H< eat Removal (RHR) System"; Branch Technical Position RSB 5-1, " Design Requirements of the Residual Heat Removal System"; and Regulatory Guide 1.139, " Guidance for Residual Heat Removal." These documents represent current staff criteria and are used in the review of facilities being processed for operating licenses.
This comparison of the existing systems against the current licensing criteria led naturally to at least a partial comparison of design criteria, which will be input to SEP Topic III-1, " Classification of Structures, Components and Systems (Seismic and Quality)."
As noted above, the six topics were considered while neglecting possible interactions with other topics and other systems and components not directly related to safe shutdown.
For example, Topics II-3.8 (Flooding Potential and Protection Requirements),
II-3.C (Safety-Related Water Supply), III-4.C (Internally Generated Missiles), III-5.A (Effects of Pipe Break on Structures, Systems, and Components-Inside Containment), III-6'(Seismic Design Considera-tions), III-10.A'(Thermal-Overload Protection for Motors of Motor-Operated Valves), III-11 (Component Integrity), III-12 (Environmental Qualification of: Safety-Related Equipment) and V-1 (Compliance with
. Codes and Standards) are among several topics which could be affected by the results of the safe shutdown review or could have a safety impact upon the systems which were reviewed.
These effects will be determined by later review.
This review did not cover, in any significant detail, the reactor protection system, nor the electrical power distribution, both of which will be reviewed later.
The major factor in, assessing the safety margin of any of the SEP facilities depends upon the ability to provide adequate protection
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-for postulated Design Basis Events (DBEs).
The SEP topics provide a major input to the DBE review, both from the standpoint of assessing the probability of the event and that of determining the consequences of events.
As examples, the safe shutdown topics below pertain to the-listed DBEs (the extent of applicability will be determined during the DBE review for Lacrosse BWR.
Completion of the safe shutdown topic review (limited in scope as noted above), as documented in this report, provides significant input in assessing.the existing safety margins.
i
. Impact.Upon Probability Topic DBE Group
- Or Consequences of DBE V-10.8 VII (Spectrum of Loss of Coolant Consequences Accidents) 4 V-11.A VII (Defined above)
Probability V-11.8 VII (Defined above)
Probability VII-3 All (Defined as a generic topic)
Consequences IX-3 III (Steam Line Break Inside Consequences Containment)
(Steam Line Break Outside Containment)
IV (Loss of AC Power to Station Consequences Auxiliary)
-(Loss of all AC Power)
.V (Loss of Forced Coolant Flow)
Probability (Primary Pump Rotor Seizure)
(Primary Pump Shaft Break)
.VII (Defined above)
Consequences X
II (Loss of External Load)
Consequences (Turbine-Trip)
(Steam Pressure Regulator
[ closed])
(Loss of Feedwater Flow)
-(Feedwater System Pipe Break)
III (Defined above)
Consequences
-IV (Defined above).
Consequences Consequences V (Defined above)
VII (Defined above)
Consequences "For a listing of DBE groups and generic topics,~see Reference 5.
.H 2.0 DISCUSSION 2.1 Normal Plant Shutdown and Cooldown (Offsite Power Available, All Equipment Operable)
The plant conducts a controlled shutdown in accordance with written procedures by reducing generator load gradually with the main steam bypass valve closed and turbine inlet pressure controlled at 1225 psig. The initial pressure regulator is in " automatic".
Turbine load is decreased by control rod insertion at about 1.5 MWe/ minute to 10%.
The regulator closes down the inlet valve to maintain 1225 psig and feedwater flow is controlled in automatic.
At about 10% power, station load is transferred to the reserve auxiliary transformer and control rods are inserted to achieve about 2 MWe power at which time the turbine is tripped.
Power is then decreased to the point that it just offsets heat losses and effects of reduced steaming.
Feedwater is automatically reduced to match steam flow and secured when this flow reaches its minimum value..These operations minimize pertubations and require minimum actions by operators.
Seal injection _and purification add on the order of -70 gpm of
.subcooled water to the reactor coolant inventory.
Steaming continues to the steam jet air efectors and gland seal thus removing heat and water.
The decay heat blowdown valve is manually controlled to blow down via the decay heat removal system to the main condenser to 4
maintain constant inventory.
Component cooling flow to the decay heat exchanger (100 gpm minimum) is then established.
Flux is maintained in the low heating range if this condition (hot standby) is desired.
When proceeding to cold shutdown, all control rods are inserted; purification, seal injection, and the air ejectors are secured; main condenser vacuum is reduced to atmospheric pressure; and the main steam isolation valves are closed to minimize the cooldown rate.
At a reactor coolant system (RCS) temperature of 470*F, the Decay Heat Cooling System may be placed in service circulating reactor coolant through the tube side of the decay heat cooler.
Component cooling water is circulated through the shell side and is in turn cooled in the shell side of the component cooling water coolers by circulating low pressure service water through the tube side.
The cooling path to the Mississippi River is thus established and maintained to t
complete cooldown and remove decay heat while cold.
Cool down rates are controlled to maintain reactor vessel temperatures in a range to avoid excessive stresses.
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- 2. 2 Shutdown and Cooldown with Loss of Offsite Power On loss of offsite power the main condenser is unavailable for heat removal for cooldown and the feedwater pumps cannot be used for b
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reactor coolant makeup because onsite sources do not have sufficient capacity.
After the reactor is tripped, the shutdown condenser is activated manually, automatically by closure of either MSIV, or automatically by system pressure 1325 psig or greater. A cooling path is established by opening either of two inlet valves and either of two condensate return valves.
The shutdown condenser is a closed loop which establishes natural circulation by condensing steam boiled off from the reactor vessel in the tube side of the condenser and returning the condensed steam via gravity flow to a feedwater line and then to a reactor forced circulating loop.
Being a closed loop the need for makeup water is minimized.
The condenser can transfer about 8% of rated reactor power from the reactor coolant to the shell side water which is boiled off to.the atmosphere.
The condenser shell side water level is controlled to provide makeup from the demineralized water storage tank and at a lower level (if demineralized system supply is insufficient, exhausted or fails) by high pressure service water.
The demineralizer water transfer pumps are powered from the diesel backed essential buses and are capable of adding up to 30,000 gallons of water stored in the virgin water storage tank to the condenser as needed.
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Additional demineralized water may be obtained from the adjacent Unit 3 by making a flexible hose connection.
Normally, however, the lower controlled level is reached, and high pressure service water system automatically provides makeup using either of two diesel powered pumps which take suction directly from the river.
Redundant and diverse methods of makeup water with onsite or offsite power are thus available to the shell side of the condenser.
The reactor coolant system cooldown rate may be controlled by controlling the position of the steam inlet valves.
The system will cool the RCS to cold shutdown (about 212 F) and maintain it there indefinitely.
Needed instrumentation is powered from the essential busses and provided in the control room.
Another method of cooling is avafiable, but is to be used only if other methods fail. This method requires the activation of the manual depressurization system (MOS) and subsequent use of alternate core spray.
This is not included as a means for removal of decay heat in plant procedures since venting to containment requires plant downtime for cleanup and restoration to normal conditions.
Use of alternate core spray has as its primary purpose responding to loss of coolant events. The depressurization system contains two vent valves which open to the containment atmosphere.
This function can be performed only if part of_ the shutdown condenser system _is activated, i.e.,-the venting is~ performed by passing steam through 1.
. the shutdown condenser steam inlet valves to and through the reactor vent valves.
Each vent valve requires the opening of its corre-sponding shutdown condenser steam inlet valve in order for it to function to blowdown reactor system pressure.
Each vent valve is operated from the control room by its independent manual switch.
The vent valves open on loss of air or loss of power to the solenoids.
Following the decrease of system pressure to less than 150 psi, the alternate core spray is activated and delivers ample flow (about 900 gpm at a vessel pressure of 50 psig).
This water is provided by two separate diesel driven pumps that take suction from the river and discharge through valves to a 4" line which penetrates the reactor vessel head. This water flows through the open end 4" pipe on to the high pressure core spray system tube bundles, l
through the flow deflector plate area and downward to the core.
Note that operation of this system is procedurally required if the vessel water level is low enough (-12 inches) coincident with containment pressure of 5 psig.
The manual depressurization and alternate core spray systems are designed to comply with the interim acceptance criteria for ECCS.
(Reference No. 1)
I
4.
3.0 CONFORMANCE WITH BRANCH TECHNICAL POSITION 5-1 FUNCTIONAL REQUIREMENTS The system (s) which can be used to take the reactor from normal operating conditions to cold shutdown shall satisfy the functional requirements listed below.
1.
The design shall be such that the reactor can be taken from normal operating conditions to cold shutdown using only safety-grade systems.
These systems shall satisfy General Design Criteria 1 through 5.
2.
The system (s) shall have suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities to assure that for onsite electrical power system operation (assuming offsite power is not available) and for offsite electrical power system operation (assuming onsite power.is not available) the system function can be accomplished assuming a single failure.
3.
The system (s) shall be capable of being operated from the control room with either only onsite or only offsite power
.available with an assumed single failure.
In demonstrating that the system can perform its function assuming a single failure, limited operator action outside of the control room would be considered acceptable if suitably justified.
. 4.
The system (s) shall be capable of bringing the reactor to a cold shutdown condition, with only offsite or onsite power available, within a reasonable period of time following shutdown, assuming the most limiting single failure.
3.1 Background
A " safety grade" system is defined, in the NUREG-0138 (Reference 2) discussion of issue #1, as one which is designed to seismic Category I (Regulatory Guide 1.29), quality group C or better (Regulatory Guide 1.26), and is operated by electrical instruments and controls that meet Ir,stitute of Electrical and Electronics Engineers Criteria for Nuclear P
. ant Protection Systems (IEEE 279).
The Lacrosse Plant was built and received its Provisional Operating Authorization prior to the issuance of Regulatory Guides 1.26 and 1.29 (as Safety Guides 26 and 29 on 3/23/72 and 6/7/72, respectively).
- Also, proposed IEEE-279, dated August 30, 1968, was not used in the design of the fa ility.
Therefore, for this evaluation, the systems that should be " safety grade" systems are the systems identified in Table 3.1 and in Section 3.2.
General Design Criterion (GDC) 1 requires that_ systems important to safety be designed, fabricated,_ erected, and tested te quality standards,'that a Quality Assurance (QA) program be implemented to
, assure these systems perform their safety functions, and that appropriate records of design, fabrication, erection, and testing are kept.
Regulatory Guide (RG) 1.26 provides the current NRC criteria for quality group classification of safety-related systems.
Table 3.1 provides a comparison of the safe shutdown systems with RG 1.26.
The classification of all systems important to safety for the Lacrosse BWR will be determined under SEP Topic III-1, "Classifi-cation of Structures, Systems, and Components (Seismic and Quality)."
Table 3.1 of this report will be used as input to Topic III-1.
At the time the Lacrosse BWR Plant was licensed, the NRC (then AEC) criteria for QA were not developed.
However, the QA program for operation of Lacrosse has been -revi2wed by the staff and found to be in conformance with 10 CFR 50, Appendix B (Reference 3). Appro-priate records concerning design, fabrication, erection and testing of equipment important to safety are maintained by the licensee in accordance with the QA program and the plant Technical Specifications.
GDC 2 states that structures and equipment important to safety shall be designed to withstand the effects of natural phenomena without loss of capability to perform their safety function.
Natural phenomena considered are:
hurricanes, tornadoes,. floods, tsunami, seiches and earthquakes.
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The effects of tornadoes will be reevaluated during the course of the SEP in Topics II-J A, " Severe Weather Phenomena," III-2, " Wind and Tornado-Loadings," and III-4.A, " Tornado Missiles." The effects of flood on the Lacrosse Plant will be reassessed in the SEP review under Topics II-3.8, " Flooding Potential and Protection Requirements,"
and III-3, " Hydrodynamic Loads." And within the SEP review, the potential for and consequences of a seismic event at the Lacrosse site will be reassessed under several review topics.
GDC 3 requires structures, systems, and components important to safety to be designed and located to minimize the effects of fires and explosions.
The Lacrosse fire protection reevaluation resulting from the Browns Ferry fire is currently underway in the NRC Division of Operating Reactors.
The results of this reevaluation will be integrated into the SEP assessment of the Lacrosse Plant.
GDC 4 requires that equipment important to safety be designed to withstand the effects of environmental conditions for normal opera-tion, maintenance, testing, and postulated accidents.
Also the equipment should be protected against dynamic effects including internal and external missiles pipe whip, and fluid impingement.
The SEP will consider the various aspects of this criterion when reviewing topics III-12. " Environmental Qualification of Safety-Related Equipment," III-5.A " Effects of Pipe Breaks Inside Contain-ment," III-5.B, " Pipe Breaks Outside Containment," and III-4,
" Missile Generation and Protection."
GDC 5 is not applicable for the Lacrosse Station because it does not share any equipment with other facilities.
3.2 Shutdown Systems The safe shutdown systems which should be " safety grade" are:
1.
Reactor Protection and Control Systems (no discussion included) 2.
Shutdown Condenser 3.
Demineralized Water Transfer System 4.
Manual Depressurization System (HOS) 5.
Alternate Core Spray (ACS)*
6.
Reactor Building and Turbine Building Main Steam Line Isolation Valves "A single check valve isolates the High and Low Pressure Service Water Systems from the ACS.
Therefore, these systems should be of the same seismic and quality classification as the ACS.
These service water systems are not required to function for safe shutdown.
O t 7.
Instrumentation for the Above Systems and Equipment 8.
Emergency Power (AC and DC) for the Above Systems and Equipment In addition to these systems, other systems may function as backup for the above systems and components.
Some of the backup components are discussed in this and other Sections of the report to identify alternate ways that may provide an acceptable level of safety.
Shutdown Condenser System The shutdown condenser system provides the capability to take the reactor from hot shutdown to cold shutdown; i.e., BTP 5-1 Functional Requirement No. 1, is described below:
t The shutdown system consists of the shutdown condenser, piping, valves, and associated instrumentation.
The shutdown condenser is located on a platform 10 feet above the main floor in the reactor building.
Steam from the 10-inch main steam line passes through a 6-inch line, two parallel inlet steam
. control valves, back to a 6-inch line and into the tube side of the condenser where it is condensed by evaporating cooling water on the shell side.
The steam generated in the shell is exhausted to the atmosphere through a 14-inch line which penetrates the reactor building. An area monitor is located next to the steam vent line near the containment shell penetration in order to detect excessive activity release in the event of a shutdown condenser tube failure.
The main steam condensate is collected and returned to the reactor i
vessel by gravity flow. The condensate line leaving the condenser is a 6-inch line along the horizontal run and is reduced to a 4-inch line for the remainder of the vertical section.
Two parallel condensate outlet control valves are located in the 4-inch return line.
The condensate line also contains two 2-inch vent lines
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which join together and return to the lower channel section of the condenser.
The vents are provided for returning vapors and/or noncondensible gases which are carried into the condensate line back to the condenser to prevent perturbations in the condensate flow leaving the condenser.
The lower channel section in turn is vented to the off-gas system through a 1-inch vent line.
Flow in i
this vent line is restricted by a 1/16-inch orifice, which is built into and is an integral part of the shutdown condenser off gas control valve seat.
. A vent line containing two parallel control valves is connected to the 6-inch condensate return line.
The valves discharge directly to the reactor building atmosphere and can be used, in an emergency, to vent air from the reactor vessel in the event that it should become necessary to flood the reactor building due to a large leak below the reactor core.
This would permit the water level in the building to equalize with that in the reactor vessel.
The shell, steam inlet channel, condensate outlet channel, and tube sheet are made of carbon steel.
All parts in contact with the reactor steam are clad with monel, except the tubes which are made of 70 percent copper and 30 percent nickel.
The shutdown condenser is designed to absorb, without camage, the thermal and physical shock of going from ambient temperature to full load conditions in 5 seconds for 50 cycles during a 20-year unit lifetime.
The thermal shock is equivalent to a temperature transient of 500 F in 5 seconds.
The tubes are seal welded to the tube sheet and the tube sheet is welded to the shell; however, a cutting ring is provided for tube bundle removal should the need arise.
Due to the large temperature differential between the reactor steam and the shell side cooling water, a thermal barrier is provided to reduce thermal stresses in the tube sheet.
The barrier cnnsists of N
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a shield plate and individual tube ferrules which insulate the steam inlet channel and tube sheet from the cold water.
The large temperature differential also has led to the use of finned tubes.
The finned tubes present an irregular surface on the outside tube, which limits film boiling by breaking up the film.
Film boiling tends to insulate the tubes, thus decreasing the heat transfer coefficient.
The tube surface area is 3705 square feet and will 6
handle 39.3 x 10 Btu /hr at a transfer rate of 27.7 Btu /hr/sq ft/*F.
The system piping is designed for a maximum working pressure of 1400 psig at 650'F.
System operating pressure is 1250 psig.
The steam piping from the biological shield to the inlet of the shutdown condenser is constructed of Schedule 120 low-alloy steel.
All other steam and condensate piping are of Type 304 stainless steel.
The steam piping within the biological shield and the 6-inch section of the condensate return line is Schedule 120 and the 4-inch section is Schedule 80.
Valves in the system meet and are in accordance with ASME c:. des and standards.
The condenser has 2 - 6" steam inlet angle valves.in parallel.
-They are air operated, air to close, controlled by 125V DC control
o
, power.
They fail open on loss of power and may be manually vented at the valve station on the platform at the shutdown condenser (about 5 minutes from the control room).
The solenoids for control air to each valve are provided from separate DC sources.
The positica of each valve is indicated in the control room by a hand indicator-controller by selection and by indicating panel lights.
The condenser condensate return to reactor has 2 - 4" angle valves in parallel. They are air operated, air to close, controlled by 125V DC control power.
They fail open on loss of DC or air and may be manually vented at the valve station on the platform at the shutdown condenser (about 5 minutes from the control room).
The solenoids for control air to each valve are provided from separate DC sources.
The position of each valve is indicated in the control room by panel lights.
The only instrumentation required to know the status of Shutdown condenser cooling is reactor coolant system pressure.
Redundant measurements are provided in the control room.
Also, secondary side water level is provided in the control room and at the condenser platform.
Other indications in the control room are tube side vent pressure, shell and tube side temperature, valve positions, and controls.
1
. Reactor pressure channels 1 or 2 will automatically activate the condenser system at a pressure of 1325 psig.
Cicsure of either main steam line reactor or turbine building isolation valves (1 of 2 for each valve) will automatically activate the condenser system.
Simultaneous failure of either valve closure circuits (2) will result in high reactor pressure and activation of both channel 1 and 2 reactor pressure protection circuits and the condenser.
Thus the protection is redundant and diverse on isolation valve closure.
J The condenser system may be manually activated.
Secondary side water level is indicated locally by two gauge glasses; remote indication is provided by an air to current converter.
The current to the indicator in the control room also provides high and low level alarms. The air supply is provided by plant air system.
The level controller provides a 3 - 15 psi air signal split, 3 - 9 psi to high pressure service water system makeup valve and 9 - 15 psi for the demineralized water makeup valve.
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Secondary makeup water is provided by the demineralized water system with the high pressure service water (HPSW) system providing a backup. supply.
The lines for demineralized makeup and HP59 are continually; pressurized by pumps. The domineralized water transfer pumps are M
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powered from separate essential busses and these buses are diesel generator backed.
The :iPSW pump is AC motor driven and takes a suction on the low pressure service water (LP5W) system.
The HPSW pump is backed up by two pumps driven by separate diesel engines, the alternate core spray pumps.
The makeup water system relies on a single controller.
Although the makeup water is provided from redundant sources each with its own piping and valves, a single failure of this controller in a fashion to keep air on the control valves would not indicate the failure in the control room, nor would makeup be provided.
Following boil off of the contained water, system pressure would begin to rise.
If the operator _could not go to the condenser platform inside containment in time, the safety valves would open.
Air is supplied to the shutdown condenser valves and the level controller from the containment building control air by the station air system.
This system has two two-stage compressors.
The system air provided for control comes from these compressors and is filtered and dried.
This air is backed by an instrument air system that has two compressors driven by a single motor.
These compressors come on at a drop from a normal 100 psig to 75 psig.
An adequate air supply is thus provided.
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With only onsite power available, the backup instrument air system is the only source available since it is the only compressor powered from the essential bus.
It is needed for operation of the secondary side level controller; however, on loss of instrument air the makeup valves fail open and provide a continuing saakeup supply to the secondary side.
Alternate Ccre Spray System Alternate core spray (ACS) in conjunction with the manual deprc;=uifta-tion system (MDS) is redundant to the shutdown condenser.
Its use in this mode is not included in plant procedures.
To use alternate core spray requires reactor pressure to be below 150 psig.
Therefore, the vent valves must be opened and pressure reduced if the shutdown conden' ser is not func'tioning'.
Heat added to the primary containment by the vent valves is transferred to the atmosphere through th'e containment dome wall.
The vent valves are manually operated in this situation.
They are separate and redundant and no single failure will prevent the manual depressurization system from performing its function (Reference 1).
However, to use either vent valve requires its corre-sponding shutdown condenser system steam inlet valve be open.
These operations are controlled manually and-understood by senior operators but not incorporated into existing plant procedures.
The reason they are excluded is because the consequences of unneeded venting are severe and would require significant interruption in operations.
We believe these operations should be identified in plant procedures.
L__.
Decay Heat Cooling System Following a normal reactor shutdown the decay heat cooling system is used to cool the reactor to 120*F.
Usually the main condenser is preferred for decay heat removal until the RCS is considerably cooler than the allowed 470*F decay heat cooling initiation temper-ature because the cooling rate can be controlled better.
After initiation it is used to maintain the reactor water temperature below 120*F while the reactor vessel is open for refueling or alterations. Also it is used to provide additional heat to the reactor to satisfy loop piping Nil Ductility Transition (NDT) temperature requirements and provides a path via the blowdown line to remove excess reactor water from the reactor to the main condenser.
The decay heat cooling system takes its suction from the inlet line to the forced circulating pump 1A, reactor water then flows to the decay heat pump which discharges to the tube side of the decay heat cooler and returns it to the reactor side of the forced circulation pump isolation valves.
The shell side of the decay heat cooler is cooled by the component cooling water system which is completely redundant in equipment and power supply.
The component cooling water system heat exchangers shell sides are cooled by the service water system which obtains its water from the Mississippi River.
Since the service water system is not powered from the essential
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busses, a loss of offsite power renders the service water pumps inoperable thereby interrupting the heat exchange from the decay heat cooling system to the river.
To reestablish the continuity of heat exchange the component cooling system heat exchangers shell side can be connected, via a flexible hose connection, to the fire water system which utilizes two diesel driven pumps that also receive their water supply from the Mississippi River.
This is covered in operating procedures.
The decay heat pump is a single stage centrifugal pump powered from the essential bus diesel backed and fed through reactor building motor control center lA.
It is designed for continuous operation 4
at 1500 psig and 585'F.
Control stations for the pump are located in the control room and locally.at the pump, indication of pump status is provided at both locations.
In addition to the pump status indications (green light pump energized,-red light pump deenergized, white light-auto trip), an audible alarm'annunciates in the control room whenever the automatic trip functions.
The decay heat cooler can serve as a cooler (normal mode) or as a heater.
In the cooling mode, reactor water enters the tube side and makes two passes then exits, component cooling water enters the
26 -
shell side also making two passes.
In the heater mode, which is used to facilitate loop NDT requirements, heating steam enters the shell side.
The shell side of the cooler is rated by design at 150 psig and 375'F, the tube side is rated by design at 1500 psig and 470*F.
The system design limit of 470'F is based on thermal stressing of the cooler tube sheet.
All piping and valves in the decay heat system with the exception of the blowdown line are designed to 1400 psig and 595*F and are made of 304 stainless steel. The blowdown line which is used to maintain a constant volume of water in the reactor is made of carbon steel and is designed to 1450 psig and 650 F.
The blowdown valve which is a 2-inch, 1500 psig, air operated valve can be controlled locally by handwheel or remotely from the control room.
Approximately 70 gpm of subcooled water enters the primary system through the seals of the forced circulation pumps, rod drive mechanisms, and the purification system.
During operation this 70 gpm serves as makeup water and is of no consequence, however, during shutdown when normal feedwater is reduced, the 70 gpm becomes excessive; therefore, the decay heat system is utilized to blowdown the' excess inventory to the main condenser.
The blowdown valve.is electrically interlocked with the reactor protection system such k.
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27 -
4 that a low reactor water level scram will close the blowdown valve to preclude further reduction of reactor water level.
The blowdown valve fails closed on a loss of control air or electric power thereby-removing any possibility of draining the reactor.
i During local operation of the blowdown valve reactor system pressure is av'ailable for the operator to monitor.
During remote operation the operator has available all of the parameters in the control room. - Local operation is required during shutdown from outside the control room.
-The decay heat system blowdown line can also be used during a " Feed and Bleed" operation using the ECCS pumps to feed the system and the blowdown line to reduce and maintain the reactor water level at a predesignated level.
Thus the RCS can be cooled'by this mode as
' an alternate cooldown method when offsite power is available.
Although the decay heat system is'normally used during routine
~
I:
shutdown of the reactor, it has no redundant components and lacks redundant power supplies; however', the function of cooldown is provided'for by use~of the shutdown condenser or manual depressur-
.ization and alternate. core spray systems thereby providing redundant diverse methods'for cooldown.
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We have concluded that all operations needed for safe shutdown with offsite or only onsite power available can be done from the control room.
Nevertheless, with credit for limited action outside the control room, the shutdown condenser reliability can be improved.
By dispatching an operator to the platform inside containment, the single failure vulnerability of the secondary side level controller can be overcome.
The sight glasses show water level, the makeup valves can be adjusted to provide makeup water and reactor pressure and water level are indicated in the control room.
However, it should be noted that the condenser does not have to be immune from single failure considering the redundant and diverse depressuriza-tion coupled with alternate core spray which provides an alternate shutdown and cooldown method without offsite power.
Likewise, in the temperature range below 470 F, the Decay Heat Cooling System can be used to cooldown and remove decay heat while cold. Manual operations are required for the loss of offsite power situation.
The shutdown condenser will bring the reactor to cold condition from operating conditions. The condenser relies on boil off of secondary water to accomplish cooldown.
Therefore, it will bring the RCS temperature to 212*F.
The heat removal capacity is so large (equivalent to 8% of rated power) that steam inlet flow must be controlled to avoid excessive thermal stress to the reactor
29 -
vessel.
Even with the failure of the secondary side level controller, and dispatch of the operator to the platform, the rate of cooldown will have to be controlled to avoid excessive thermal stresses.
The controlled cooldown rates are reasonable.
When the manual depressurization and alternate core spray systems are used to cool down, the vent valves provide a rapid depressuri-zation to atmospheric pressure and the alternate core spray will cool the core indefinitely.
This system has been analyzed by the staff and found to be acceptable for emergency core cooling during a loss-of-coolant accident with only onsite or offsite power available and with the most limiting single failure.
The capacity of the combined cooling of the shutdown condenser (8%
of rated power) from operating temperature (or alternatively manual depressurization with the vent valves) to 470 F and the Decay Heat Cooling System below 470*F with a heat removal capability about 1/6 l
that of the shutdown condenser, the cooldown can be accomplished.
This alternative will require much longer to cooldown since fire l
water must be connected to provide cooling to the secondary sida of the component cooling heat exchanger, t
TABLE 3.1 CLASSIFICATION OF SHUTDOWN SYSTEMS LACROSSE BOILING WATER REACTOR Quality Grcup Seismic Plant Plant System, Subsystem, Component R.G. 1.26 Design R.G. 1.29 Design Remarks Shutdown Condenser (tube side)
ASME III ASME VIII Category I Note 1 Note 1: Although not Class 1 Case 1270N originally designed to withstand a seismic (shell side)
ASME III ASME VIII Category I Note 1 event, an analysis of Class 3 Case 1270N the LACBWR concluded that the facility Piping from reactor vessel ASME III
?
Category I Note 1 could withstand an to shutdown condenser up to Class 1 earthquake with and including safety valves, corresponding maximum main stear
- isolation valve horizontal ground 64-25-030, aain feed shutoff acceleration of 0.12g valve, MDS valves (62-25-013, 014) and vent and drain lines larger than 1" diameter.
Vent and Drain piping smaller ASMF III
?
Category I Note 1 Footnote 2(a) to than 1" diameter Class 2 10 CFR 50.55a.
Demineralized Water System Transfer pumps (2)
ASME III
?
Category I Note 1 Supplies water to the Class 3 shutdown condenser Piping from Demin. Water Storage tank to Shutdown Condenser, OHST, and Alternate Core Spray System via transfer pumps Demin. Water Storage Tank ASME III API-12G Category I Note 1 Class 3 i
Overhead Storage Tank (OHST)
ASME III AWWA D-100 Category I Note 1 Class 2
TABLE 3.1 (Continued)
Quality. Group Seismic Plant Plant System, Subsystem, Component R.G. 1.26 Design R.G. 1.29 Design Remarks Alternate Core Spray (ACS)
Pumps (2 diesel driven)
ASME III
?
Category I Note 1 These pumps are Class 2 same as diesel fire pumps Diesel engine fuel suppI.-
ASME III
?
Category I Note 1 Portions of steam Class.2 line are used for both MDS and Piping from pumps to outermost ASME III
?
Category I Note 1 Shutdown Condenser.
containment isolation valve up Class 2 to and including relief valves
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and HPSW supply, drain vent, and test line isolation valves, strainers, and valves which isolate non-essential portions of the system.
Piping from outermost contain-ASME-III
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Category I Note 1 ment isolation valves up to Class 1 reacter vessel including vent
. piping greater than 1" diameter.
Process Instrumentation and Category I Note 1 Controls NA Category I Note 1 Emergency Power Supply System N4 Diesel generators Category I Note 1
t TABLE 3.1 (Continued)
I Quality Group Seismic Plant Plant
- System, Subsystee, Component R.G. 1.26 Design R.G. 1.29 Design Remarks DC. power supply systems Category I Note 1 Distribution lines, switchgear, Category I Note 1 motor control centers Reactor Control and Category I Note 1 Protection System NA l
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. 4.0 SPECIFIC RESIDUAL HEAT REMOVAL AND OTHER REQUIREMENTS OF BRANCH TECHNICAL POSITION 5-1 4.1 RHR System Isolation Requirements The RHR system shall satisfy the isolation requirements listed below.
1.
The following shall be provided in the suction side of the RHR system to isolate it from the RCS.
(a) Isolation shall be provided by at least two power-operated valves in series. The valve positions shall be indicated in the control room.
(b) The valves shall have independent diverse interlocks to prevent the valves from being opened unless the RCS pressure is below the RHR system design pressure.
Failure of a power supply shall not cause any valve to change positior,.
(c) The valves shall have independent diverse interlocks to protect against one or both valves being open during an RCS increase above the design pressure of the RHR system.
2.
One of the following shall be provided on the discharge side of the RHR system to isolate it from the RCS:
(a) The valves, position indicators, and interlocks described in item 1(a)-(c),
(b) One or more check valves in series with a normally closed power-operated valve.
The power-operated valve position shall be indicated in the control room.
If the RHR system discharge line is used for an ECCS function, the
. power-operated valve is to be opened upon receipt of a safety injection signal once the reactor coolant pressure has. decreased below the ECCS design pressure.
(c) Three check valves in series, or (d) Two check valves in series, provided that there are design provisions to permit periodic testing of the check l
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. valves for leak tightness and the testing is performed at least annually.
Lacrosse Boiling Water Reactor does not have a low pressure redundant residual heat removal system.
The system which performs the residual heat removal function for the latter stage of normal cooldown and for long term decay heat removal while depressurized is the Decay Heat Cooling System.
Since it is designed for reactor coolant system (RCS) pressure, the isolation requirements listed above do not apply; i.e.,
the potential for an RCS break from overpressurization because of valving errors is of no safety consequence.
4.2 Pressure Relief Requirements The RHR system shall satisfy the pressure relief requirements listed below.
1.
To protect the RHR system against accidental overpressuriza-tion when it is in operation (not isolated from the RCS),
pressure relief in the RHR system shall be provided with relieving capacity in accordance with the ASME Boiler and Pressure Vessel Code.
The most limiting pressure transient during the plant operating condition when the RHR system is not isolated from the RCS shall be considered when selecting the pressure relieving capacity of the RHR system.
For example, during shutdown cooling in a PWR with no steam bubble in the pressurizer, inadvertent operation of an additional charging pump or inadvertent opening of an ECCS accumulator valve should be considered in selection of the design bases.
2.
Fluid discharged through the RHR system pressure relief valves must be collected and contained such that a stuck open relief valve will not:
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. I a.
Result in flooding of any safety-related equipment.
A b.
Reduce the capability of the ECCS below that needed to mitigate the consequences of a postulated LOCA.
c.
Result in a nonisolatable situation in which the water provided to the RCS to maintain the core in a safe condition is discharged outside of the containment.
3.
If interlocks are provided to automatically close the isolation valves when the RCS pressure exceeds the RHR system design pressure, adequate relief capacity shall be provided during the time period while the valves are closing.
The Decay Heat Cooling System does have code required relief valves set at design pressure to protect against overpressurization while isolated, and it is protected from overpressurization while in service by code safeties.
Again, it is a full pressure system and protection from accidental overpressurization is adequate with reactor vessel code safety valves.
Its heat exchanger is tempera-ture limited to operation below 470*F, and this limit is administra-tively controlled.
4.3 Pump Protection Requirements The design and operating procedures of any RHR system shall have provisions to prevent damage to the RHR system pumps due to over-heating, cavitation or loss of. adequate pump suction fluid.
The pump is designed for continuous operation at 1500 psig and 585'F.
The-pump requires a net positive suction head (NPSH) of approximately 10 ft. to preclude cavitation and eventual impeller corrosion.
Due to its location it has an available suction head of
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' approximately 100 ft., cooling water to the pump bearings is supplied by the component cooling water system.
4.4 Test Requirements The isolation valve operability and interlock circuits must be designed so as to permit on line testing when operating in the RHR mode.
Testability shall meet the requirements of IEEE Standard 338 and Regulatory Guide 1.22.
Based on the previous discussion, the Decay Heat Cooling System does not require isolation and interlock circuits.
4.5 Operational Procedures The operational procedures for bringing the plant from normal operating power to cold shutdown shall be in conformance with Regulatory Guide 1.33.
For pressurized water reactors, the opera-tional procedures shall include specific procedures and information required for cooldown under natural circulation conditions.
The licensee has procedures to perform safe shutdown operations including shutdown to hot standby, operation at hot standby, hot shutdown, operation at hot shutdown and cold shutdown including long-term decay heat removal.
The licensee has also provided the operating staff procedures covering offnormal and emergency conditions for shutting down the reactor and decay heat removal under conditions of loss of system or parts of system functions normally needed for shutdown and cooling the core.
Procedures for systems operation for systems used in safely shutting down the reactor are also
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. i included in the plant operating procedures.
These procedures include provisions identified in Regulatory Guide 1.33.
These precedures were reviewed and are in conformance with Regulatory Guide 1.33.
Some operations are not covered that should be addressed are:
1.
Providing demineralized water from the onsite Unit 3 to the reactor plant demineralized water system.
2.
Contingencies for failure of shutdown condenser secondary side water level controller and loss of air supply to controller.
3.
Use of manual depressurization system'in conjunction with alternate core spray as a backup safe shutdown system.
Altern-atively, we believe this method may be covered in the licensee's retraining program and simply identif.ied as a means of achieving cooldown in procedures, if the licensee desires.
He has indicated that the severity of contamination following unneeded use of this method would faterrupt plant operation for a long period to clean up the'contamidation, and he would prefer to omit from procedures to avoid use unless other methods 1. ave proven unsatisfactory.
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5.0 RESOLUTION OF SEP TOPICS The SEP topics associated with safe shutdown have been identified in the Introduction to this assessment.
The following is a discussion of how the Lacrosse Plant meets the safety objective of 1
these topics.
5.1 Topic V-10.8 RHR System Reliability The safety objective of this topic is to ensure reliable plant shutdown capability using safety grade equipment subject to the guidelines of SRP 5.4.7 and BTP RS8 5-1.
The Lacrosse BWR systems have been compared with the criteria of BTP 5-1 and the results of these comparisons are discussed in Sections 3.0 and 4.0.
Section 3.0 4
discusses th'e way the functional requirements are met and Section 4.0 discusses the Decay Heat Removal System which performs the function identified in BTP RSB 5-1 as Residual Heat Removal.
The Decay Heat Removal System has very limited use in the Lacrosse plant and it l
does not contain system redundancy.
However, we have concluded that the other Lacrosse systems acceptably fulfill the safety l
objective subject to the resolution of the following in the SEP l
integrated assessment:
a 5
as
. 1.
The requirement for using only safety grade equipment to 4
accomplish the shutdown and cooldown.
The seismic and quality classification of safe shutdown systems identified in Section 3.0 will g2 established in further review of Topic III-1.
2.
The need for improved shutdown condenser shell side level control to preclude a single failure disabling the condenser.
Resolution of this item can be postponed until the integrated assessment because of the existence of a redundant cooldown method (MOS and ACS).
5.2 Topic V-llA Requirements for Isolation of High and Low Pressure Systems The safety objective of this topic is to assure adequate measures are taken to protect low pressure systema connected to the primary system-from being subjected to excessive pressure which could cause failures in some plants have the pctential for causing a LOCA outside of containment.
The topic in this review is concerned only with the decay heat cooling system; high/ low pressure interfaces with other systems were not reviewed.
Since this system is completely contained within containment, except for a portion of the blowdown line, and since it is designed for system pressure, the overpressure
. potential is minimal (i.e., the same as the rest of the RCS); and the topic is resolved for the decay heat cooling system.
5.3 Topic V-11.8 RHR Interlock Requirements The safety objective of this topic is identical with V-11.A.
The interlock would close low pressure isolation valves when open and high pressure excursion occurs and would prohibit opening when high pressure exists.
Again, this system is designed for full system pressure and the interlocks are unnecessary.
This topic is resolved for the decay heat removal system.
In addition to these requirements, and as a matter to be resolved separately from the SEP, the NRC staff has determined that certain isolation valve configurations in systems connecting the high-pressure Primary Coolant System (PCS) to lower-pressure systems extending outside containment are potentially significant contributors to an intersystem loss-of-coolant accident (LOCA). Such configurations have been found to represent a significant factor in the risk computed for core melt accidents (WASH-1400, Event V). The sequence of events leading to the core melt is initiated by the failure of two in-series valves to function as a pressure isolation barrier between the high-pressure PCS and a lower-pressure system extending beyond containment. This causes an overpressurization and rupture of the low-pressure system, which results in a LOCA that bypasses containment.
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, The NRC has determined that the probability of failure of these valves as a pressure isolation barrier can be significantly reduced if the pressure at each valve is continuously monitored or if each valve is periodically inspected by leakage testing, ultrasonic examination, or radiographic inspection. NRC has established a program to provide increased assurance that such multiple isolation bar;iers are in place in all operating Light Water Reactor plants. This program has been designated DOR Generic Implementation Activity B-45.
In a generic letter of February 23, 1980, the NRC requested all licensees to identify susceptible valve configurations which may exist in any of their plant systems communicating with the PCS.
For plants in which valve configuratioris of concern were found to exist, licensees were further requested to indicate: 1) whether, to ensure integrity, continuous surveillance or periodic testing was currently being conducted, 2) whether any valves of concern were known to lack integrity, and 3) whether plant procedures should oe revised or plant modifications be made to increase reli abil. i ty.
LACBWR is one of those plants identified as being susceptible to the potential failure because of the configuration of valves (two check valves in series) in the Alternate Core Spray system. Therefore, as noted, action will be taken independently of the SEP effort to resolve the "Ev2nt V" problem.
hhk 5.4 Topic VII-3 Systems Required for Safe Shutdown The safety objectives of this topic are:
'l.
To assure the design adequacy of the safe shutdown system to (a) initiate autoniatically the operation of appropriate systems, including the reactivity control systems, such that specified acceptable fuel. design limits are not exceeded as a result of
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.and (b) initiate the operation of systems and components required'to bring the plant to a safe shutdown.
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To assure that the required systems and equipment, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown are located at appropriate locations outside the control room and have a potential capa-bility for subsequent cold shutdown of the reactor through the use of suitable procedures.
3.
To assure that only safety grade equipment is required for a PWR* plant to bring the reactor coolant system from a high pressure condition to a low pressure cooling condition.
Safety objective 1(a) will be resolved in SEP Design Basis Event Reviews.
These reviews will determine the acceptability of the plant response, including automatic' initiation of safe shutdown related systems, to Design Basis Events, i.e., accidents and
' transients.
Objective 1(b) relates to availability in the control room of the control and instrumentation systems in the control room are capable of following the-plant shutdown from its initiation to its conclusion-at cold shutdown conditions.
The ability of_the Lacrosse plant to l;
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- follow the shutdown is discussed in the preceding sections of this report.
Based on these discussions and knowledge gaineo from the site visit, we conclude that safety objective 1(b) is met by the safe shutdown systems, actuation mechanisms, and control room displays, at Lacrosse subject to the findings of related SEP Electrical, Instrumentation and Control topic reviews.
Safety objective 2 requires the capability to shut down to both hot shutdown and cold shutdown conditions us'ing systems, instrumentation and controls located outside the control room.
Lacrosse plant has a procedure, " Emergency Reactor Shutdown and Cooldown When the Control Room is Inaccessible." The procedure assumes lack of time to trip reactor prior to leaving the control room and indicates the locaticn from which the reactor can be tripped.
It covers emergency communications and locations of portable radios.
It designates the operators stations and actions to be taken by them, provides adequate instructions for ascertaining the operability and condition of the essential plant equipment and indicates the surveillance instrumen-tation and instructions for interpreting the information..The plant relies on the Shutdown Condenser and Decay Heat System to control shutdown and cooldown.
The review team visited each designated operators station and-assessed the capability of the plant staff to perform the necessary t
t operations. We conclude that the plant can perform these shutdown operations and indeed has done so.
Early in plant life power cables were cut inadvertently and the plant lost all AC power.
Operators were dispatched to their local stations and the plant was successfully shut down and controlled.
The adequacy of the safety grade classification of safe shutdown systems at Lacrosse, to show conformance with safety objective C, will be completed in part under SEP Topic III-1, " Classification of Structures, Components and Systems (Seismic and Quality)", and in part under Design Basis-Review Event reviews.
5.5 Topic X Auxiliary Feed System (AFS)
The safety objective ~of this topic is to assure that the AFS can provide adequate cooling water for decay heat removal in the event of loss of all-main feedwater using the guidelines of SRP 10.4.9 and BTP ASB 10-1.
This topic is'not applicable to Lacrosse.
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6.0 REFERENCES
i
' l.
Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No. 6 to Provisional Operating License No. DPR-45, Dairyland Power Cooperative, Lacrosse Boiling Water Reactor, Docket No. 50-409, August 12, 1976.
2.
Staff Discussion of Fifteen Technical Issues Listed in Attachment to November 3,1976 Memorandum from Director, NRR to NRR Staff, NUREG-0138, November 1976.
3.
NRC letter, R. Reid to J. Madgett, dated February 14, 1977.
- 4.
Lacrosse Boiling Water Reactor Operating Manual, Volumes I through V.
-5.
Systematic Evaluation Program,-Status Summary Report, NUREG-0485.
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