ML19305E111
| ML19305E111 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 02/29/1980 |
| From: | Harold Denton Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19305E110 | List: |
| References | |
| DD-80-09, DD-80-9, NUDOCS 8004220422 | |
| Download: ML19305E111 (17) | |
Text
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l DD 9 UNITED STATES OF AMERICA l
NUCLEAR REGULATORY COPNISSION I
0FFICE OF NUCLEAR REACTOR REGULATION HAROLD R. DENTON, DIRECTOR
)
In the Matter of DAIRYLAND POWER COOPERATIVE Docket No. 50-409 I
(Lacrosse Boiling Water Reactor)
DIRECTOR'S DECISION UNDER 10 CFR 2.206 By petition dated May 21, 1979, Ms. Anne K. Morse requested that either the Nuclear Regulatory Comission's (NRC or the Comission) Director of Nuclear Reactor Regulation or the Director of Inspection and Enforcement order suspension of Provisional Operating License No. DPR-45 issued for operation of Dairyland Power Cooperative's (the licensee or DPC) Lacrosse Boiling Water Reactor (LACBWR).
This petition has been considered under the provisions of 10 CFR 52.206 of the Comission's regulations.
Notice of receipt of the petition was published in the Federal Register June 26,1979 (44 FR 37352). The licensee submitted a response to Ms. Morse's petition in a letter dated July 3, 1979.
Ms. Morse presents seven bases for her petition which she asserts show that continued operation of LACBWR is inimical to the health and safety of the public.
Each of these bases is discussed in this decision. Upon review of Ms. Morse's petition, the staff has determined that Ms. Morse has not presented any new information or reasons which would provide a basis for suspending operation of the Lacrossa facility at this time.
However, as discussed in this decision, the NRC staff does support Ms. Morse's concern about the liquefaction issue i
involving LACBWR and has issued to the licensee an " Order to Show Cause,"
l dated February 25, 1980, regarding this matter. Accordingly, for the reasons stated in this decision, Ms. Morse's petition has been granted in part and denied in part.
80042204,p
e 2
CONTAINMENT VENTING (ITEM 1)
Ms. Horse asserts that an NRC letter dated November 29, 1978, questions the safety of continuous venting of containment and that this letter directed the licensee to either cease entirely or limit containment purging to a maximum of 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> per year until such time as further evaluation of said practice could be completed.
The NRC November 29, 1978 letter discussed background information related to containment structures for nuclear power plants where the accident analysis that NRC has reviewed and approved assumes that the containment purge valves are closed during nonnal operation in contrast to the LACBWR where the ventilation dampers are normally op'en.
The letter was a request for information that would permit NRC to conduct a generic evaluation of containment purge valve use in r
practice in contrast to design expectations. The letter was not an order to close ventilation dampers except for 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> a year. The NRC's expressed concern relates to (1) accident consequences where containment pressure increases following loss of coolant accidents could be different from the documented analysis due to large (up to 48 inch diameter) open purge valves if the original analysis assumed the valves to be closed and (2) containment purge valve tests, analysis and qualification which provide assurance that the valves will close and seat properly against the dynamic forces of a design basis loss of coolant accident.
Operation of the LACBWR plant with the ventilation dampers open as designed was reviewed and approved by the Atomic Energy Commission (AEC-the NRC's predecessor) prior to initial operation of LACBWR.
The LACBWR Safeguards Report dated July 1965 states in Section 6.64 that the 20-inch containment ventilation system inlet and outlet isolation dampers are normally open. The redundant inlet and outlet l
. dampers shut off and seal ne plant's ventilating systen in the event of an accident, thereby preventing the release of fission products through these containment penetrations to the atmcsphere. These valves were designed to close on any of the following signals:
(1) high activity measurement by the gaseous monitor or either particulate monitor sampling the exhaust duct air leaving the building; (2) high reactor pressure; (3) high pressure in the containment building; (4) remote manual operation from the control room; and (5) loss of electrical power supply to the solenoid valves.
[Each ventilation damper has a spring-loaded air cylinder operation, twa limit switches, and a three-way solenoid valve. The solenoid valves are nonna11y energized by 115 v.d.c. to air load the damper (butterfly valve) air cylinder operators. When the solenoid valves are de-energized, the air cylinder operators are vented and the spring loading closes the butterfly valves within two seconds thereby isolating the containmentventilationsystem].
j 1
The following additional infonnation related to isolation damper reliability was provided by Allis Chalmers in response to AEC questions prior to grantino l
authorization for operation of LACBWR:
o Section 14.3.18 of the LACBWR Safeguards report of July 1%5 presents the analysis of the maximum credible accident including the containment pressure transient, the assumed containment leakage and the fission product release from the containment building.
o Allis Chalmers report ACNP-66501 (January 1966) presents containment pressure and temperature effects of loss of coolant accident responsive to NRC question 1-1.
y.
7-v
I
- i Allis Chalmers report ACNP-66531 (April 1966) addresses containment syster o
operation in response to NRC question IV-1.
o Allis Chalmers report ACNP-66516 (June 1966) provides analysis of containment building pressure following maximum credible accident responsive to NRC i
question I11-11.
Allis Chalmers report ACNP-66512 (February 1966) provides assurance that o
in accordance with emergency procedures the operator checks containment building dampers and the 4" vent header valves following containment isolation, for automatic closure in response to NRC question III-8.
Allis Chalmers reports ACNP-66523 and 66525 (f! arch 1966) states that the o
ability of the reactor building ventilation dampers to operate whenever the reactor building air exhaust, gaseous and particulate monitor indicates high activity will be tested twice a week as part of the radiation monitor test.
This was later modified to bi-weekly for the LACBWR Operating Technical Specifications.
(TS 5.2.15).
The AEC's review of the LACBWR contain.nent continuous ventilation system was completed prior to the issuance of the authorization for operation of LACBWR.
Automatic closure of the redundant ventilation dampers in the intake and exhaust system (4 valves) is required by the LACBWR Technical Specification 2.1.2.5(2).
In addition, the following requirements are also conditions of the LACBWR operating license:
TS - 2.10.1.4 requires radiation monitors to detect and indicate radiation levels and to cause reactor building ventilation system isolation if excessive radiation levels should occur within containment.
TS - 2.11.2.5 requires that containment building ventilation system exhaust be monitored for radioactivity prior to release through the stack.
- 5 TS - 5.2.15 items 11 and 12 require gas and particulate radiation monitor calibration at rach refueling, testing every two weeks, and daily checks.
TS - Table 1 requires closure of the redundant 20 inches inlet and outlet ventilation system dampers and 4 inch vent header valve before:
Table 1, item 3 - reactor coolant pressure exceeds 1305 psig, Table 1, item 16 - containment pressure exceeds 5 psig, and Table 1, item 18 - radiation levels exceed the maximum permissible concen-trations for continuous releases to the atmosphere as specified in Table II of 10 CFR Part 20, Appendix B.
The NRC's November 1978 request emphasized the importance of the mechanical qualification of purge and vent valves. Most facilities, including LACBWR, use butterfly valves for containment isolation because of quick closing capability.
The NRC's concern centered on the capability of the containment isolation valves in the purge and vent systems, af ter being opened during hot standby, hot shutdown, startup or power operation modes, to close against the fluid dynamic conditions of a postulated design basis accident condition upon receipt of an isolation signal, i
These fluid or aerodynamic forces originate from the pressure drop imposed across the closing valves by the ascending pressure in containment following the postulated design basis loss of coolant accident.
Normally these valves will receive an isolation signal either from high radiation monitors or high pressure. monitors, or both. Upot. receipt of the isolation signals the valves are required to seal
- losed within several seconds.
Potential failures affecting the purge and vent penetration valves could lead to degradation in containment integrity.
From staff stodh. and discussions with manufacturers involved in supplying these valves the follow 1 g conclusions can be made-1.
Most valves of this type will tend to close under the dynamic forces of a LOCA, if they are operated in a partially opened position.
r 2.
Partial opening of the valves between 30* and 50 of full open will
]
in most cases significantly reduce dynamic loads seen by valve component 3.
Demonstration of operability for most valves of this type can be obtained through analysis and previous testing data.
I Based on these conclusions, the NRC staff.has developed an interim position on use of containment isolation valves, which was sent to the licensee on October 23, 1979.
In accordance with this interim position, the licensee has committed to limit the opening of the LACBWR valves to no more than 50, later reduced to 25, cf full open (DPC letters dated December 7, 18 and 28, 1979).
The licensee also reports that until the maximum valve opening can be limited to 25, the 20-inch isolation valves will remain closed except to prevent deterioration of the containment atmosphere.
In addition, for the long term solution of this issue the licensee and Allis-Chalmers, the valve manufacturer, have conducted scale model tests of the valves simulating the LOCA accident environment. Guidelines for long term demonstration of purge and vent valve operability have been developed and are being used to assess the valves installed in operating plants (NRC letter dated September 27,1979).
It is projected that this effort will be completed near the beginning of FY 80 and that some operating and systems modifications will be required.
The licensee will assess the long term operability of the plant's valves utilizing these guidelines and the infomation gained through testing recently completed.
Based on the aforementioned data there is reasonable assurance, during the interim, that these valves will operate during a design basis accident.
NRC is continuing to evaluate the information provided by the licensee regarding reactor containment purging and ventilation as well as the infomation provided by the licensees of all other operating reactors to determine what, if any, changes may be necessary to further reduce the risk of accidental release of radioactivity that could affect the health and safety of the public.
Based on our review to date, I would not order suspension of License No. DPR-45 because of containment venting.
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- l LIQUEFACTION ANALYSIS (ITEM 2)
Tne NRC staff and its consultant, the U. S. Army Corps of Engineers Waterways Experiment Station (WES) have canpleted their review of the licensee's inves-
.l tigation of the liquefaction potential at the LACBWR site. The reviewed report is entitled " Liquefaction Potential at Lacrosse Boiling Hater Reactor (LACBWR)
{
Site near Genoa, Vernon County, Wisconsin" by Dames and Moore dated August 10, 1979.
During a meeting with the NRC staff on October 17, 1979, the licensee submitted the final report, dated September 28, 1979, which contains only minor modifications to the August 10 draft. The licensee's consultant, Dames and Moore, has concluded "that the threshold liquefaction resistance at the LACBWR site occurs for a design SSE which yields a maximum ground surface acceleration greater than 0.189 and less than 0.20g".
Based on review of this report, we conclude that if sustained strong ground motion with peak accelerations of.12g or higher occurs (normally associated with a magnitude 5 or greater earthquake) liquefaction can occur down to a depth of Below.08, we conclude that there is little potential for liquefaction.
40 feet.
9 These conclusions are based on our comparison of this site with other sites where liouefaction has occurred and on the use of laboratory strength data as interpreted by the staff and our consultant Dr. William Marcuson, a WES geotechnical engineer.
WES has provided a letter dated October 19, 1979, which further defines the basis for this conclusion.
In sunmary, based on judgment concerning the density and strength data and on empirical correlations WES concludes that the foundation material below the water table down to a depth of 40 feet is not safe against liquefaction if the licensee designated safe shutdown earthquake with a peak acceleration of 0.129 occurs.
l
. s In our opinion, the more recent investigations, report dated August 10, 1979, undertaken by the licensee's consultant Dames and Moore, Inc. confirm the previous conclusion that the soils at the La Crosse site could strain badly during a postulated earthquake producing a surface level peak acceleration of 0.12g as noted by WES in " Liquefaction Analysis for Lacrosse Nuclear Power Station," dated December,1978.
A final version of the WES report was issued as Paper GL-79-ll, da ted June,1979. Although the staff's evaluations to date indicate that there is a relatively low seismic hazard at the LACBWR site (discussed infra), our current evaluations suggest that soil liquefaction could occur if ground motion at the.129 level occurred during an earthquake at the site.
The staff further discussed the liquefaction issue with the licensee in a meeting on November 2, 1979. At that meeting the licensee agreed to consider remedial measures to preclude liquefaction at the site. On November 29, 1979, the licensee submitted for the NRC staff's review its conceptual design for a dewatering system to preclude liquefaction. The staff's preliminary review of the proposed dewatering system indicates that the system is a feasible soletion to the potential liquefaction problem at the LACBWR site. The staff is ur.able to determine conclusively at this time, however, that the proposed system will preclude liquefaction with reasonable certainty during potential earthquakes with peak accelerations of.12g cr less, because the final design of the system has not yet been developed by the licensee and submitted to the NRC for review.
I have, therefore, issued the attached Order, which requires the licensee to show cause why it should not submit by May 27, 1980, a detailed design
. proposal for a site dewatering system and why it should not implement such system, after the NRC approves it, or shut down the LACBWR facility by February 25, 1981.
Because the seismic hazard associated with the LACBWR site is relatively low, the Order does not require shutdown of the LACBWR during the development and implementation of the site dewatering system.
As discussed in the Order, the staff has made an initial estimate of the probability of exceeding a range of peak accelerations at the La Crosse site in order to make an estimate of the hazard associated with the liquefaction potential.
In doing so, we utilized all readily available estimates of earth-quake probability that included the site region.
These were estimates taken from Milne and Davenport (1969), Algermissen and Perkins (1976), the Applied Technology Council (1978), the Haven Site Preliminary Safety Analysis Report (1978), and preliminary results from the Systematic Evaluation Program (SEP) probabilistic study of the La Cross site.
The Safe Shutdown Earthquake (SSE) free field ground motion designated by the licensee in the full term license application is.12g anchored to a Regulatory Guide 1.60 spectrum.
Based on our review of probabilistic studies listed above, the return period for.12g would be at least 1,000 years. This peak acceleration (.129) is equivalent to Intensity VII when utilizing the relationship of Trifunac and Brady (1975). The return period for.089 would be at least 400 years. These values are based upon the minimum return period calculated in the above studies. While these values should not be interpreted
7, as absolute minimums, the actual return period could be an order of mr.gnitude larger. As mentioned above, these estimates are preliminary and only serve j
to indicate the general level of seismic hazard at the site.
As part of the SEP Program, we are currently evaluating the SSE seismic design at La Crosse.
Based upon limited consideration of current Standard Review Plan procedures, the La Crosse site lies in an area of low seismicity in the Central Stable Region Tectonic Province. The highest intensity near the i
site historically was estimated to be Intensity V due to the 1811-1812 New Madrid earthquakes, 800 kilometers from the La Crosse site. The 1909 Beloit earthquake on the Wisconsin-Illinois border probably produced Intensity II to IV at the site. The site is not located near any known localizers of seismicity. Based on a recent staff decision for the Tyrone construction permit application, the SSE intensity could be VI: or VII-VIII for the general region including the La Crosse site.
Using the Trifunac and Brady (1975) relationship, the free field ground motion corresponding to Intensity VII would be.13g and Intensity VII-VIII would be.20g, which would be used as the high frequency anchor to the Regulatory Guide 1.60 response spectrum.
Based on the estimates of return periods of earthquakes with potential
.129 ground acceleration the staff has concluded that the general level of seismic hazard at the LACBWR site is sufficiently lcw that operation of the plant for the next twelve months would not endanger the health and safety of the public. To the extent that Ms. Morse's petition requests suspension of operation of the LACBWR plant while the liquefaction issue is being resolved, her petition is denied.
- 11 APPENDIX J REQUIREMENTS (ITEM 3).
10 CFR 650.54(0) of the Commission's regulations requires that primary reactor containments for water cooled reactors shall be subject to the require-ments set forth in Appendix J.
Appsndix J to 10 CFR Part 50 specifies the requirements for testing procedures, testing frequency and testing method.
Appendix J also specifies the leakage limits for determining test failures, and associated reporting procedures for such failures including corrective action plans to effect the repairs. The intent of periodic containment leak tests is to detect the leaks so that prompt corrective action can be taken to restore leak tightness and prevent gradual deterioration.
If significant containment leaks are f,ound, the plant will be shutdown until containment integrity is restored and demonstrated by successful and more frequent tests.
Contrary to Ms. Morse's suggestion, the licensee has complied with the intent of Appendix J.
The licensee has been conducting its containment leak test in accordance with the approved requirements contained in the Technical Specifications (Section 5.2.1) and has also been reporting all test failures t
and associated corrective actions for repairs as required by Appendix J.
P The fact that some electrical penetrations have failed the leak tests does
- Moreover, not provide a basis for suspending operation of the Lacrosse facility.
the licensee has taken appropriate corrective action where necessary upon f ailure of any leak test. There is no indication that the licensee will not continue to take such corrective action or that the result of the licensee's tests thus far indicate a safety problem that justifies suspension of operation of the LACBWR facility.
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- 12 FORCED CIRCULATION (ITEM 4)
LACBWR has had several instances during its commercial operation where a reactor coolant forced circulation pump had to be removed from service because of TheshaftsealingarrangementconsIstsof problems with the pump shaf t seals.
four sections comprised of mechanical seals, seal rings, a buffer-type seal, and a labryrinth bushing to backup the mechanical seal. Tolerances between the pump shaf t and the seals exist to permit seal injection water to cool and lubricate the rotating seal members. The seal water injection system operates at a higher pressure than the reactor coolant system, maintaining a pressure differential between the twc that ensures reactor coolant will not pass out through the seals. Normal seal leak-off water is returned to the seal water injection pump reservoir. The seal water injection system has several alarms and protective circuits which alert the operator of abnormal conditions and trip the reactor's forced circulation pump if the anomalies are not corrected.
These protective features are designed to minimize the loss of primary :oolant and to protect the pump from excessive damage due to improper seal cooling and lubrication. The alarms and associated pump trips are actuated by diverse parameters such as low differential pressure, high differential pressure, high temperature in the leak-off lines, low flow in the inlet lines, high or low flow in the leak-off lines. The seal water injection system's protective features mitigate the severity of damage and potential radiological consequences associated with a malfunction of a seal or the seal water injection system.
l LACBWR's forced circulation pump seals have caused the seal water injection system to actuate a pump trip due to excessive flow created by seal degradation.
In every instance, the system performed as intended without loss of primary coolant or damage to the pump. The latest pump seal failure in December 1978
caused the licensee to operate the facility for a period of approximately two months with one recirculation pump and its associated loop out of service before repairs could be performed. On previous occasions, the pump was placed out of service, repairs made, and the pump returned to service without the need to operate in the single loop configuration since this repair was made during a scheduled reactor shutdown for refueling or maintenance and there was no benefit in ~ continuing operation with only one loop in service.
Dairyland Power Cooperative has investigated the cause of the seal degradation and has attributed it to a change in the seal material which was not as durable as that used by the original pump vendor.
The change in the hardness of the seal material lead to faster degradation.
Dairyland Power Cooperative has replaced the seals in both forced circulation pumps with an improved seal material. The pumps have since operated for two months without failures.
Past seal failures have typically occurred in a matter of days after pump startup. This operating l
problem appears to have been resolved.
The NRC approved LACBWR Operating Technical Specification 4.2.2.9 permits reactor operation with only one of the two forced circulation loops in service at power levels up to 82.5 MW.
By letter dated April 19, 1979, the NRC staff t
transmitted to Dairyland Power Cooperative a safety evaluation perfonned by the Systematic Evaluation Program staff supporting operation with less than all loops in service. The NRC review considers such things as impact on normal operation, the potential for accidents not previously evaluated, and the calculated effect on previously analyzed accidents and transients.
Based on our review, we conclude that operation with less than all loops in service at LACBWR continues j
to be acceptable in accordance with the limits of Technical Specification 4.2.2.9.
- 14 OPERATIONAL RESTRICTIONS (ITEMS 5 AND 6)
Operational restrictions were placed on LACBWRl/ as a result of gross fuel failures experienced in Cycle 4.
These restrictions were intended to provide a means of monitoring and limiting the progression of fuel failures. The limits were primarily set te preclude the severity and number of fuel failures to levels lower than experienced in Cycle 4.
These limits are well below fuel damage limits which would be inimical to the health and safety of the public.2]
The LACBWR Technical Specification implemented for Cycle 5 limited offgas ratio factor to activity to a maximum specified value which included a account for changes in power level. Also included was a term allowing for offgas activity which m,ay be generated from the residual activity leftover from the gross failures of Cycle 4.
The allowable offgas due to the residual activity was measured by the licensee during startup for Cycle 5.
The licensee did not exceed the offgas Technical Specification limits at any time during Cycle 5 operation. However, a high residual activity term allowed LACBWR to operate with a higher offgas activity limit than the limit set for Cycle 6.
This occurred because the reactor water cleanup system effectively removed the residual uranium which resulted in less offgas due to the residual activity, thus allowing more offgas activity from damaged fuel. To better account for the effectiveness of the reactor water cleanup system, we requested and the licensee agreed to change the LACBWR Technical Specifications to reduce the allowable residual activity term to 10% of the initial value in 50 daysE. (This was based on alpha activity graphs from Cycle 5 which provide a good indication of removal rates of the residual uranium).
The i
1/ Amendment No.11 to License No. DPR-45, dated March 3,1978.
2] Enclosure 2 to the Commission's transmittal letter for Amendment No.11, dated March 3,1978, titled " Analysis of LACBWR Fuel Failures," dated February 1978.
3.] Amendment No.16 to License No. DPR-45, dated May 25, 1979.
- 15 '
new limits make the offgas technical specifications more restrictive in that permitted fuel damage during Cycle 6 will be well below that allowed for Cycle 5 because the total offgas limit will be reduced due to the reduction in the residual offgas limit.
The reactor coolant water exceeded the gross alpha activity limit on 5 occasions during the early days of Cycle 5.
In each case reactor power level was reduced as required by the Technical Specifications. These incidents were the result of residual activity from Cycle 4 and not as a result of new failures.
On each occasion the activity returned to within the Technical Specification limits (as verified by. samples) in approximately 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. After these early incidents, the licensee operated the remainder of Cycle 5 without exceeding Technical Specification limits.
Whereas fuel damage was predicted for Cycle 5, the severity and numbers of damaged fuel was maintained well below that experirnced in Cycle 4l!. The end-of-Cycle 5 (E0C-5) fuel inspections indicated that less then 0.3% of the core sustained damage. This was in good agreement with the predicted results. As it is impossible to assure 100% fuel integrity, the 99.7% fuel integrity obtained during Cycle 5 is considered acceptable, and represents a damage limit that does not pose a threat to the health and safety of the public. 2/
If DPC letter dated May 9, 1979 (Attachment 2)
-2/ Amendment No.11 to License No. DPR-45, dated March 3,1978; to the Connission's transmittal for Amendment No.11, dated March 3,1978, entitled " Analysis of LACBWR Fuel Failures,"
dated February 1978.
The NRC expects that fuel failures during Cycle 6 operation will be signi-ficantly less than the failures experienced during Cycle 5.
This is due to (1) the reduced number of Allis Chalmers fuel assemblies that remain in the core, (2) the remaining Allis Chalmers fuel assemblies being positioned at low power locations, and (3) local power peaking in the fuel assemblies as a result of control rod movements being minimal for Cycle 6 locations of the Allis Chalmers fuel assemblies.
Based on our evaluations and the revised Technical Specifications we have concluded that there will be fewer fuel rod failures in Cycle 6 than Cycle 5 and that there is no increased risk to the health and safety of the public that would justify an order to shut down the LACBWR plant.
SPENT FUEL STORAGE P0OL (ITEM 7)
There is no NRC requirement for licensees to maintain space in the spent fuel pool for a full core offload.
It is the RC's position that the health and safety of the public is not impaired by leaving a core in the reactor vessel.
Thus, if some time in the future the Dairyland Power Cooperative does not have i
l storage capacity in the spent fuel storage pool, the spent fuel assemblies may be stored in the reactor vessel. Therefore, I find no basis to order the suspension of operation of the plant because of an impending shortage of storage space for spent fuel.
CONCLUSION Based on the foregoing discussio1 and the provisions of 10 CFR 52.206, I have determined that there is no adequale basis for suspending Dairyland Power Cooperative's License No. DPR-45 for the LACBWR plant However, as discussed in this decision, the NRC staff does support Ms. Morse's concern about the
. liquefaction issa involving LACBWR and has issued to the licensee an
" Order to Show Cause," dated February 25, 1980, regarding this matter.
The request of Ms. Anne K. Morse is, therefore, granted in part and denied in part.
A copy of this decision will be placed in the Commission's Public Document Room at 1717 H Street, N.
W., Washington, D. C.
20555 and the Local Public Doct. ment Room for the LACBWR Plant, located at the La Crosse Public Library, 800 Main Street, La Crosse, Wisconsin 54601.
A copy cf this decision will also be filed
..a the Secretary for the Comission's review in accordance with 10 CFR 2.206(c) of the Conunission's regulations.
As provided in 10 CFR 2.206(c) of the Commission's regulations, this decision will ccastitute the final action of the Comission twenty (20) days after the date of issuance, unless the Commission, on its own motion, institutes a review of this decision within that time.
Harold R. Denton, Director Office of Nuclear Reactor Regulation Dated at Bethesda, Maryland this 29th day of February, 1980.
1 Attachments:
1.
Order to Show Cause Dtd. 2/25/80 2.
DPC Ltr. Dtd. 5/9/79 i
E [ c(7,p, UNITED STAT"ES NUCLEAR REGULATORY COMMISSION a
waswincrow, o. c.rosss
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February 25, 1980 Docket No. 50-409
?
i Mr. Frank Linder General Manager Dairyland Power Cooperative 2615 East Avenue South La Crosse, Wisconsin 54601
Dear Mr. Linder:
The Comission has issued the enclosed " Order to Show Cause," which is being filed with the Office of the Federal Register for publication.
This Order relates to the liquef. action issue involving the La Crosse Boiling Water Reactor.
Si ncerely,
Harold R. Denton, Director
~
Office of Nuclear Reactor Regulation
Enclosure:
Order to Show Cause cc w/ enclosure:
See next page
$ 603betof
February ' 25,1980 Mr. Frank Linder
,2 -
cc w/ enclosure:
Director, Technical Assessment Fritz Schubert, Esquire Division Staff Attorney Office of Radiation Programs Dairyland Power Cooperative 2615 East Avenue South
( AW-459)
U. S. Environmental Protection La Crosse, Wisconsin 54601 Agency O. S. Heistand, Jr., Esquire Crystal Mall #2 Morgan, Lewis & Bockius Arlington, Virginia 20460 1800 M Street, N. W.
U. S. Envi anmental Protection Washington, D. C.
20036 Agency Federal f tivities Branch Mr. R. E. Shimshak La Crosse Boiling Water Reactor Region V Office ATTN:
EIS COORDINATOR Dairyland Power Codperative 230 South Dearborn Street P. O. Box 135 Genoa,' Wisconsin 54632 Chicago, Illinois 60604 Coulee Region Energy Coalition Charles Bechhoefer, Esq., Chairman Atomic Safety and Licensing Board ATTN: George R. Nygaard U. S. Nuclear Regulatory Commission P. O. Box 1583' La Crosse, Wisconsin 54601 Washington, D. C.
20555 La' Crosse Public Library Dr. George C. Anderson Department of Oceanography 500 Main Street La Crosse, Wisconsin 54601 University of Washington Seattle, Washington 98195 Mrs. Ellen Sabelko Society Against Nuclear Energy Mr. Ralph S. Decker 929 Cameron Trail Route 4, Box 190D Eau Claire, Wisconsin 54701 Cambridge, Maryland 21613 l
Town Chairman Town of Genoa Route 1 Genoa, Wisconsin 54632 Chairman, Public Service Commission of Wisconsin Hill Farns State Office Building l'adison, Wisconsin 53702 l
i l
I t
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION l
t i
In the Matter of DAIRYLAND POWER COOPERATIVE
)
)
(La Crosse Boiling Water Reactor)
)
Docket No. 50-409
)
ORDER TO SHOW CAUSE I.
Dairyland Power Cooperative (DPC or licensee), La Crosse, Wisconsin, is the holder of Provisional Operating License No. DPR-45, issued on August 28, 1973, which authorizes the operation of the " a Crosse Boiling Water Reactor (LACBWR), located in Vernon County, Wisconsin.
LACBWR is a direct-cycle, variable-flow forced circulation boiling water reactor, which is designed to operate at a rated power not in excess of 165 megawatts themal.
DUPLICATE DOCUMENT
'p Entire document previously entered into system under:
Bods 940/o 2(
nO No. of pages:
/
-QO3
i
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i DA11tYLAND A*0 Weft C00E*EECATIVE Da Groue, Othconsta 54601 May 9, 1979 In reply, please refer to LAC-6274 DOCKET NO. 50-409 Director of Nuclear Reactor Regulation ATTN:
Mr. Dennis L.
Ziemann, Chief Operating Reactors Branch #2 Division of Operating Reactors U. S. Nuclear Regulatory Commission Washington, D. C.
20555
SUBJECT:
DAIRYLAND POWER COOPERATIVE LA 'ROSSE BOILING WATER REACTOR (LACBWR)
PROVISIONAL OPERATING LICENSE NO. DPR-45 REFUELING PLAN FOR CYCLE 6
Reference:
(1)
DPC Letter, LAC-6130, Linder to Ziemann, dated February 26, 1979.
(2)
NRC Letter, Ziemann to Linder, dated April 24, 1979.
Gentlemen:
Enclosed with this letter are 40 copies of the LACBWR report LAC-TR-068, "LACBWR Cycle 5 Fuel Performance and Finalized Refuel-ing Plan for Cycle 6",
dated April 1979.
There is a supplement to the report LAC-TR-067 submitted with Reference 1.
Also included with this letter as Enclosure 1 are DPC's answers to NRC questions (Reference 2) concerning the submittal in Refer-ence 1.
The report LAC-TR-068 and the answers to the NRC questions have been reviewed and approved by the LACBWR Safety Review Committee.
If there are any questions concerning this submittal, please contact us.
Very truly yours,
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