ML20247H474
| ML20247H474 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 08/20/1987 |
| From: | Muller D Office of Nuclear Reactor Regulation |
| To: | Butterfield D COMMONWEALTH EDISON CO. |
| Shared Package | |
| ML20247H448 | List: |
| References | |
| RTR-REGGD-01.097, RTR-REGGD-1.097 GL-82-33, TAC-51102, TAC-56407, NUDOCS 8905310247 | |
| Download: ML20247H474 (48) | |
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V Doc'et Nos.: '50-373 vv 'v
.and 50-374 Mr. L. D. Butterfield..Jr.
Nuclear Licensing Manager Connonwealth Edison Company P. O. Box 767 Chicago, Illinois 60690
Dear Mr. Butterfield:
SUBJECT:
EMERGENCY RESPONSE CAPABILITY, CONFORMANCE TO REGULATORY GUIDE 1.97, REVISION 2, LASALLE COUNTY STATION, UNITS 1 AND 2 (TAC NOS. 51102 AND 56407)
- Generic Letter 82-33 requested Connonwealth Edison Company (CECO) to provide a.
report to the NRC describing how the post-accident monitoring instrumentation meets the guidelines of Regulatory Guide (RG) 1.97, Revision 2, as. applied to emergency response facilities.
CECO respvnded to the generic letter on
-April 14, 1984, referring to a previous letter of June 29, 1982, for response specific to RG 1.97,~ Revision 2.
Additional information was provided by letters dated February 22, 1985 and farcis 10,'1986.
EG&G Idaho, Inc., under contract to the NRC, performed a detailed review 'and-technical. evaluation of CECO's submittals. Technical Evaluation Report (TER-EGG-EA-6770) dated November 1986. was evaluated by the Instrumentation-and Control Systems Branch. Division of~ Engineering and Systems Technology.
The staff evaluation results are contained in.the Safety Evaluation'. Report,.
Based on the staff's review, it has been concluded that the LaSalle County Station. Units 1 and 2. instrumentation complies with the recc e ndations of-R6 1.97_. Revision 2. except for the instrumentation associated with monitoring neutron flux.
.The staff has taken the position that CECO shall install and have operational monitoring instrumentation for neutron flux, which fully meets the i
i B905310247 890 j73 PDR ADOCK O PNV G
Q - - - - _ ______ _____ _. _ _ _
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Mr.. L. D. Butterfield, Jr. August 20, 1987 recommendations of RG 1.97, Revision 2.
Within 45-days of receipt of this_
letter. CECO is reauested to commit to installino the uparaded neutron flu _x l-nonitorino instrumentation system wnen available.
Sincerely, I
I sN.4 m_
Daniel R. Muller, Director Project Directorate III-2 Division of Reactor Projects - III, IV, V and Special Projects
Enclosure:
As stated cc:
See next page o
1 Commonwealth Edison AiWE 2.
One First N%onal Plaza. Cheago. lileos_
Address Revy to Post offce Box 767
(
v' Chcago,laros 60690 0767 October 7, 1987 U.S. Nuclear Regulatory Commission Attn: Document control Desk Washington, DC 20555
Subject:
LaSalle County Station Units 1 and 2 Regulatory Guide 1.97 Revision 2 TAC Nos. 51102 and 56407 NRC Docket Nos, 50-373 and 50-374 Reference (a): Letter dated August 20, 1987 from D.R. Muller to L.D. Butterfield Transmitting the Safety Evaluation Report for LaSalle County Stations Conformance to Regulatory Guide 1.97.
Dear Sir:
Reference (a) documented the conclusion of your staff that LaSalle County Station instrumentation complies with the recommendation of Regulatory Guide 1.97, Revision 2 except for that instrumentation associated with monitoring neutron flux. Your letter requested that Commonwealth Edison comicit to installing an upgraded neutron flux monitoring instrumentation system when available.
Commonwealth Edison is actively working with the BWR Owners Group evaluating potential upgrades of the neutron monitoring system. We will j
follow reecamendations made by the BWR Owners Group for upgrade of the instrumentation when available.
If you have any further questions please contact this office.
C. M. Allen Nuclear Licensing Administrator cc: Regional Administrator - RIII Resident Inspector - LSCS P. Shemanski - WRR M. Parker - IDNS 3670K M7/33lhf4343 If.
408925fcd76.,0ct)888 8:48 P.02 GENE LZC & CONSUL SRVCS TEL No.
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BLA#H OulNERS' GROUP "Wairti e.amemam m a m au.w m. n w. m uom BWROG-8817/BVR1 April 1, 1988 U.S. Nuclear Regulatory Commission Division of BWR Licensing Washin5 ton, D.C.
20535 Attention:
T.E. Murley, Director NRR BWR OWNERS' GROUP LICt.NSING TOPICAL REPORT
- POSITION
SUBJECT:
REGUIATORY GUID81.97, REVISION 3 REQUIREMENTS FOR POST ACCIDENT NEUTRON MONITORING SYSTEM" (GENERAL ELECTRIC REPORT NED0 3155 Gentlemen:
The BWR Owners' Croup (BWROC) has completed -its analysis of requirements The results are for BWR Post-Accident Neutron Monitoring instrumentation.
ured to establish appropriate ncutron monitoring poet cccident functional Deviations from RG 1.97 requirements are justified.
design criteria.
The BWROC approach and overall strategy has been previously discussed with Joyce and others) at a aseating in Bethesda on the NRC (Joseph P.At this meeting the NRC indicated their willingness to January 27, 1988.
evaluate alternate approaches to R0 1.97 usutron monitoring system requirements to resolve this issue.
This letter submits 30 copies of the subject Licensing Topical Report for NRC review and approval.
It is requested that this report receive priority review since several BWROC members have near term license actio the R0 1.97 requirements for post-accident I
associated with resolvingIn the interim, the BWR Ownars' Croup requests that neutron monitoring.
all post-accident NHS implementation requirements be ' deferred until thisTh report has been carefully evaluated by NRC reviewers. cost f selected (total replacement of SRM/IRM million depending upon options specifically for post accident use).
subsystem or addition of equipment Due to the high cost, installation dose significance, and cost benefit uncertainty associated with this plant modification, we do not recommend that installation decisions should be required prior to resolution of this Licensing Topical Report.
b Q
1 4
LGENE LIC & CONSUL SRUCS TEL No.
408.92,5 2476.0c9.18,88_ 8:49 p.03 BWROG 8817/3WR1 April 1, 1988 Page 2 This Licensing Topical Report has been endorsed by a substantial number of tho members of the BWR Owners' Group; however, it should not be interpreted as a commitment of any individual, member to a specific course of action.
Each member must formally endorse the SWROG position in order for that position to become the member's position.
Very truly yours,
- n '"
1 Robert F. Janecek, Chairman BWK Owners' Group
/ta Attachment BWR Owners' Croup Primary Representatives cc:
BWR Owners' Group Executive overview committee M.W. Hod 6cs (NRC)
J.P. Joyce (NRC)
R. Evane (NUMARC)
V.S. Green (INFO)
H. Wyckoff (EPRI)
D.N. Graco (BWROG Vice Chairman)
L.S. Gifford (CE, Bot.hesda) 9 4
e _-_--- ---- - _
i IIKK_HUMDER UNEESQLYED ITEM i's 2 AND 3 REPORT SECT 1Qti 50-373/80027-02 (DRS)
Range for the reactor vessel water 4.a.(1)(a) 50-374/88026-02 (DRS) level instrumentation not in compliance with Regulatory Guide 1.97, Revision 2.
50-373/88027-03 (DRS)
Unquallfled isolation device used 4.a.(1)(b) 50-374/88026-03 (DRS) for vessel level recorder LT-1B21-26BA.
i l
Reactor Water Level Range Issues This issue has been addressed and accepted by the NRC in response to NUREG 0737, Item II.F.2, (which clarified the requirements for post accident reactor water level monitoring.) CECO's original submittal (6/29/87) for R.G.
1.97-stated that the criteria was met for a range of -160"* to
+60",
and that Ceco would comply with the BWR owner's group position addressing detection of inadequate core cooling. The purpose for monitoring reactor water level during accident recovering is to verify that there is accomplishment of core cooling.
(See table 1 of R.G.
1.97 and pgh. 6.2.3 of ANS 4.5).
7n addition it is desired to verify " accomplishment of mitigation". All emergency operating procedure guidelines and LaSalles LGA procedures have been prepared to require operators to maintain RPV level above -161" (top of active fuel).
Provided that there is water covering the core, adequate core cooling is assured.
(Reference NUREG 0626 and NEDO 24708A.) Also, to verify that accident mitigation is being accomplished, the level range from above the top of the active fuel (-161" RWL) to high water level trip (+55"RWL) is monitored to assure that once ECCS systems have been initiated, level is in fact being restored. All operator actions that are required in the Emergency Operating Procedures are based on the operators verification that water level restoration above the top of active fuel is being accomplished.
To monitor the potential of the core being uncovered during a design basis LOCA a water level monitoring system with a range of -150" to +60" RWL has been designed.
This system meets the qualification requirements of R.G.
1.97, for Type B, Category 1.
Its range requirements are to be able to provide the operator with information that the ECCS systems are functioning i
properly.
Further, NUREG 0737 Supplement I and the D. Eisenhut clarifications required that "(6) the indication must cover the full range, from normal operation to complete core uncovery."
In response to this, CECO revised its response to FSAR question 0 31.287 (attachment 3) by defining its post accident water level indication system to include Fuel Zone instruments (Bottons of active fuel to over 4 feet above top of active fuel) as well as the Wide Range Instruments (210" of measurement range).
(See attached copy of page Q31.287-14, dated December 1981 -
Ammendment 59).
Also, CECO committed to participate in and endorse the BWR
{
Owners group position that monitoring for inadequate core cooling by use of the level instrumentation as described in the NEDO document is sufficient for all normal, operational transient, and post-accident recovery activities CECO letter 2/9/82, attachment 2.
In its Supplement 2 (attachment 1) to the LaSalle SER, the NRC concluded "that the existing level instrumentation and the BWR Owners Group emergency operating procedure guidelines" which required that emergency operating procedures will be based on the approach to and detection of inadequate core cooling using level monitoring instrumentation systems "will satisfy the requirements of II.F.2".
Finally, CECO latest revision to the EOP's has addressed inadequate core cooling using level instrumentation, and these procedures are currently in place.
- Note that the value "-160" was a t. typographical in error.
It has been intended to refer to the originally designed BWR 5 wide range level system, which has a low end of -150" RWL.
j
To address a concern regarding the potential for reference leg holl-off (flashing) which could possibly, result in instrument error in the event that the drywell temperature is high and the vessel is suddenly depressurized, CECO committed to installing modifications to minimize the effect of this potential concern. The concern was expressed to licensees in Darrel Eisenhut's letter l
of October 26, 1984 (Generic Letter 84-23, attachment 5) which evaluated the S. Levy, Inc. Report, SLI-8211, regarding this issue and required licensees to-respond within 30 days with the proposed improvements in this area. CECO committed to respond to this concern in their letter of December 4, 1984 (Attachment 6).
A modification was implemented to re-route the reference leg such that its exposure to the potential high drywell air temperature would be minimized. The modification has been implemented on both Units. A description of this' system was contained in CECO's letter of June 10, 1986, (Attachment 7).
The NRC has reviewed this design and has found that it satisfies the concern. Their response is contained in their letter of March l
2, 1987, (Attachment 8).
I ALLAchmants i
1.
LaSalle County Station SER (NUREG 0519) Supplement 2, pp. 22-2, 3,
6, 7, 9,
and 23-1, and A-7, February 1982 i
2.
Letter, C.E. Sargent of CECO to A.
Schwencer of NRC, dated February 9, 1982 regarding CECO commitment to BWR Dwners Group position on i
Inadequate Core Cooling (NUREG 0737, Item II.F.2) 3.
Revised response to FSAR question Q31.287, page 031.287-14, submitted as part of Ammendment 59 (December 1981)
'4.
Figure B 3/4 3-1, page B 3/4 3-7 of Unit 2 Technical Specifications, Ammen6nent 33 (provided for reference only) 5.
NRC Generic Letter 84-23, dated October 26,~1984, regarding review of S. Levy Inc. report SLI-8211 6.
CECO letter, dated December 4, 1984 to H.R. Denton responding to item 5 above.
7.
CECO letter dated June 10, 1986 to H.R. Denton providing status and details regarding modifications for minimizing reference leg flashing.
8.
NRC letter, dated March 2, 1987, to CECO assessing the acceptability of Ceco's modification for minimizing the reference leg flashing errors.
NUREG-0519 Supplement No. 2
- nsa, Safety Evaluation Report related to the operation of LaSalle County Station, Units.1 and 2
)
Docket Nos. 50-373 and 374 Commonwealth. Edison Company
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U.S Mucisar Reguistory Commission Office of Nuclear Reactor Regulation f
February 1982 j
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(2) During th shift, t SCRE shall r,outinely 'nform the s foreman of any signific nt change in plant sta,tus.
)
(3) Duri the shif. the shift eman sh periodi lly check w h the SCRE for a plant s tus update 1
We d' not acce t this pro sal becaus we did n believe it ould be readi y ted.
/au I
ased on'f rther disc ssions of t' issue w' h the appli nt and the N Resident Inspecto at La Sal
, we have dified o requiremen in items (a and (b)
I above stated i our Safety valuatio as follows:
(a) he plant R0 shall p ticipate in the SCRE ift relief t nover.
(.
Durin the shift, e SCRE all inform e plant SRO o any signi icant chan es in plant tatus as soon as prac cal after a cision is eached to ake a chang. The p nt SRO shal, where pract' al, retur to the cgntrolroom o to thr e times per hift for a pl nt status pdate.
Wh e
/
60t practica to retu n to the co rol room, the lant SR0 hall perio cally check with he SCRE or a plant atus update.
In no ca should th plant SRO be re uired t abandon dut" s critical t reactor o erations t return to the ntrol r om unless s cifically ord ed to do so by the ift superv sor.
We conci de tha conformance ith this requ'rement w' I assure e plant SRO '
f aware the o erall plant atus and any volutio that are eing carried ut by th contry room SRO.
ased on disc sions wi the NRC esident Insp ctor, we nelude/that conform nce with this equirem
.t can be adily audit We wi review the La Sal Administrate' e Proced res to as re that th are dified to include these requiremejtts prior to fuel 1 ding.
We wi i conditi the operating license for Unit 1 to assure the imple ntation of this require ent.
/
II.
Siting and Design II.F.2 Instrumentation for Detection of Inadequate Core Cooling In our Safety Evaluation Report Supplement No.1, we specified that prior to issurance of an operating license, we require the applicant to commit to:
(1) incorporate incore thermocouple into the inadequate core cooling monitoring system prior to June 1983 in accordance with Regulatory Guide 1.97, Revision 2 dated December 1980, " Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," and (2) provide documentation required by Item II.F.2 of NUREG-0737, addressing the inclusion of thermocouple in the final inadequate core cooling monitoring system on a schedule acceptable to us.
Recently, representatives from General Electric and the BWR Owners Group met with us several times,. December 17, 1981 and January 27, 1982, to discuss our requirements as specified in SECY 81-582 (October 7, 1981) and the BWR Owners La Salle 55ER2 22-2 l
Group position.
As a result of these meetings, agreement has been reached to f
broaden the issue from the specific. requirement for incore thermocouple to
{
that of monitoring inadequate core cooling.
The BWR Owners Group agreed to actively participate in the analysis of inadequate core cooling instrumentation requirements and will'be submitting a final report for our review in July 1982.
By letter dated February 9,1982, the applicant committed to implement the conclusiosn reached from that evaluation on La Salle since the applicant is a member of the Group.
We conclude, therefore, that the existing level instrumentation and the BWR Owners Group emergency operating procedure guidelines, which will be based on the inadequate core cooling instrumentation requirements, to be developed and implemented upon completion of our review and approval of the applicant's report, to be submitted in July 1982, will satisfy the requirements of 7
Item II.F.2.
Accordingly, we will condition the operating license of Unit No. 1 for the submittal of this report by July 1982.
~
Final Recommendations or tsulletins and Orders Task Force
- n. K. 3 It 25.Effecto[LossofAlternatingCurrintPoweronPumpSeals/
In our Safet valuati Report, we no d that the pplicant w participating in the BWR wners Gr p study on thi tem.
Our equirement s that he applicar shows by nalysis experiment th the la S e recircul ion pum. seals can withst d a loss of cooling wat to the pu seal cool s for 2 urs.
/
B etter d ed October 23 981 from E.
. Swartz ommonweal Edison Company) o 0. G.
'senhut (NRC),
e applican as stated hat the B ' Owners Group supplem (tal input on is Item is plicable t La Salle.
his suppl ' ental input as forwarded o us by lett f dated Sep ember 21, 1 from T.
. Dente (BW Owners) to 0.
. Eisenhut C)..Dat from tests n Bingham mps with s
designs si ~ ar to those sed at La alle were p ovided.
T conditicrns ere represen tive of boili g water re ctor recire ation pum applicat n; and therefore, e results are pplicable o in plant oiling wat r reactor umps.
Observed akages were i s than 5 allons per nute for m e than 5 ours.
/
This is cceptable to III.
Emergency Pr arations d Radiation rotection j
/
I.A NRC and ensee P redness
/
III.A.2 Im ving Lic ee Emergen Preparedn s - Lo Term Discussi and Co sion Sinc the iss nce of our Su lement No, to our Safety Eva ation R ort, the Com onwealth dison Compan conducted exerci e with the resden cility o S tember
, 1981 which neluded fu activa on of the rundy C nty Emer nc.-
an and artial activ ion of the tate of Ilinois P1 As indicat in ourSupsiement,Grun County is so wit 'n the 10-m' e plum exposure mergen Planntog Zone for L Salle, and he find' gs of the ederal mergency anagem Agency specified at the Gru y Count performed oorly exerci s held
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l La Salle SSER2 22-3
(3) Prior to April 1, 1982, the applicant shall include a description of the dose calculational methodology with a Class A transport and diffusion module, and a description of an acceptable meteorological measurement-preventative and corrective maintenance program in the radiological emergency plan.
St.bject te compliance with the above identified license conditions, we conclude that the requirements of Item III.A.2 are satisfied.
- w
~n 22.3 Comaleti_on Dates of TMI Dated Items Not Meeting Scheduled NUR EG-0737 Dates At our request, the applicant by letter dated December 30, 1981 transmitted to-i us those dated items that will or may not be complete'd as specified by NUREG-0737, I
" Clarification of TMI Action Plan Requirements," its reasons for not meeting the specified date, and the new completion date.
The following is our assessment of this letter for those dated items, and Table 22.1 gives an overview of these I
dated items' s
I.
Operational Safety _
I.C Operating Procedures I. D.1 Control Room Design Review In our Safety Evaluation Report Supplemer.t No. 1, we indicated that there were some deficiencies in the control room design of La Salle that were required to be corrected prior to fuel load, some prior to full power operation, and that there were some deficiencies needing further consideration that will be required to be addressed in the applicant's detailed control room design review.
For those deficiencies required by fuel load date of which there were 68, at this time 55 have been corrected and verified.
By letter detec' December 30, 1981, the applicant proposes to correct the remaining 13 deficiencies prior to fuel.
load.
The prior to full power deficiencies have been corrected.
With this action, we conclwh that the applicant's actions to comply with the require-ments of NUREG-0737 for a preliminary design assessment of the control room design are acceptable.
I.G.1 Training During Low-Power Testing In our Safety Evaluation Report, we stated that the applicant described an operator training program to be used in conjunction with the initial testihg program.
The applicant also indicated that the La Salle initial test program will be conducted on an aruund-the-clock basis, such that, shift rotation and test cepetition will be provided to assure that all operators will participate in or observe all tests which have important training value.
The applicant asserted that the tests to be performed during the initial test program will provide sufficient training and technical data for normal and off-normal operations.
We agreed with the applicant that an adequate basis for licensing was provided.
In addition, we proposed an additional test, subject to the applicant's. safety analysis of this test, which would be a simulated loss of I
onsite and offsite alternating current power test.
Performing this test, using fission heat at a power no greater that 5 percent as required by the Action Plan, would require the operators to devote much of their attention to reactivity control at a time when reactivity control would be very sensitive, La Salle SSER2 22-6
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I I. F.' 1 Accident Monitoring Instrumentation 30, 1981 letter, the noble gas and As indicated by' the applicant's December iodine / particulate monitoring systems (Attachments 1 and 2 of Item II.F.1; the However, the final
.other Attachments 3 through 6 are completed) are installed.
testing and calibration for Attachments 1 and 2 will be delayed until February Based on our review 1982 beyond the January 1, 1982 requirement'.of NUREG-0737.
of the information provided in the Final Safety Analysis Report through Amend-ment 59, we require that the noble gas and iodine / particulate monitoring f.
-systems are operational prior to criticality.
.F$2 Instrumentation for Detection of Inadequate Core Cooling l
As we noted in Section 22.2, Item.II.F.2 of this report, we have had meetings with the applicant to discuss this Item.
As a result of these meetings,.an-agreement was reached with the BWR Owners Group to broaden the issue from only A final
(
incore thermocouple'to that of monitoring inadequate core cooling.
report will be submitted to'us by the BWR Owners Group addressing their findings in July 1982.
We, therefore, conclude that with emergency procedures, existing k
level instrumentation, the July 1982' report, and implementation cf the staff's-conclusion as a result of the NRC staff's review of this report should. satisfy our concern with this item in a timely manner.
y II.K.1 IE Bulletins on Measures to Mitigate Small-Break Loss-of-Coolant Accidents and Loss-of-Feedwater Accidents Item 22 Proper Functioning of Heat Removal ~ Systems In our. Safety Evaluation Report, we noted that the applicant has not completed.
modificatiMs to provide for automatic restart (af ter a high water trip) of the reactor core isolation system within the schedule required by Action Plan Item II.K.3.13.
In the applicant's letter dated December 30, 1981, the applicant denotes that the delay is necessary to insure the equipment is installed prior to implementing the operating procedure.
Our acceptance of the schedule slip-page for this modification is discussed under Item II.K.3.13 below.
II.K.3 Final Recommendations of Bulletins and Orders Task Force Item 13 Separation of High Pressure Coolant Injection and Reactor Core Isolation Coolina system Initiation Levels In our Safety Evaluation Report, we required that the applicant provide for automatic restart of the reactor core isolation cooling system on subsequent low water level signal af ter the system had been tripped by a high water level Modifications were to be implemented by July 1,1981.
The applicant signal.
committed to pr' ovide this modification and submitted a satisfactory conceptual design.
In addition, we stated in our Safety Evaluation Report that these modifications would not be completed on the implementation date because of the We accepted the applicant's commitment unavailability of qualified equipment.
to complete the modification prior to startup af ter the first refueling outage.
In the interim, we consider the operation during this short period to be agcep-t table because of the automatic restart capability of the high pressure core spray system and the emphasis on water level control in the plant emergency procedures.
22-9 La Salle SSER2
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23 CONCLUSIONS' Based on our evaluation of the application as set forth in our Safety Evalution Report issued in March 1981, Supplement No. 1 issued June 1981, and our evaluatio as set forth in this supplement, we conclude that the operating license for La Salle County Station, Unit 1 can be issued to allow power operations.at full rated power (3323 megawatts thermal) subject to license conditions which will require futher Commission approval.
We conclude that the construction of La Salle County Station, Unit I has been completed-in accordance with the requirements of Section 50.57(a)(1) of 10 CFR Part 50 and that construction of the plant has been monitored in accordance The construction and with the inspection program cf the Commission's staff.
We will address our review of La Salle County Station Unit 2 is not complete.
these issues in another supplement to this report.
. Subsequent to,the issuance of the operating license for full rated power for
-La St.lle County Station, Unit 1, the plant may then be operated only in accordance with the Commission's regulations and the conditions of the operating license under the continuing surveillance c.f the Commission's staff.
We conclude that the activities authorized by the operating license can be con-ducted without endangering the health and safety of the public, and we affirm our conclusions as stated in our Safety Evaluation Report and its supplements.
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23-1 Salle SSER2
January 7, 1982 Letter from applicant concerning Unit 2 Target Fuel Load Date.
l January 10, 1982 Letter to applicant concerning, Keys, Locks, and Combinations to Control Access of Protected Areas.
January 13, 1982 Letter to applicant concerning Rod Block Monitors on La Salle Q-List.
January 13, 1982 Letter from applicant concerning Process Control Program for the Radioactive Waste Solidification System.
January 21, 1982 Letter from applicant concerning Response to NUREG-0803.
January 22, 1982 We met with representatives from Commonwealth Edison and Sargent and tundy in Bethesda, Maryland to dicuss the status of La Salle, Unit 1 for licensing.
(Summary of meeting issued January 29, 1982).
January 25, 1982 Letter from applicant concerning La Salle, Unit 1 Target Fuel Load Date.
January 27, 1982*
Letter to applicant concerning Request for Additional Information on Item II.K.3.18.
February 4, 1982 Letter from applicant concerning Unit 1 Startup Test Schedule.
February 4, 1982 Letter from applicant concerning Unit 2 Response to Questions on Reactor Pressure Vessel Pressure /
erature Curves.
February 9, 1982 Letter from applicant concerning Inadequate' Core i
Cooling, NUREG-0737, Item II.F.2.
February 9, 1982 Letter to applicant concerning Delay of Preopera-tional Tests at La Salle.
La Salle SSER2 A-7
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