ML17326A329

Requests Deletion of License Condition (3)(c),per Encl Revised Response to Question 212.40 in App Q of Fsar. Revision Due to Util Misinterpretation of Requirements Re Check Valve Leak Testing.W/Fee & Affidavit

ML17326A329
Person / Time
Site:	Cook
Issue date:	10/29/1979
From:	Dolan J INDIANA MICHIGAN POWER CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
AEP:NRC:00259, AEP:NRC:259, NUDOCS 7911090472
Download: ML17326A329 (27)

Text

REGULA'TORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESS I ON NBR 0 79 1 1 090472. DOC ~ DATE ~ 79/ 1 0/2 1 NOTARIZED YES ~ DOCKET FACIAL:50 316'Donald C ~ , Cook Nuclear Power Plenty Unit 2i Indiana L 05000316 AOTHBNAME",, AUTHOR AFFILIATION Dal AA J;E. Indiana 8 Rich,igan Power Cos-

'ECIPiNAME>> RECIPIENTT AFFlLIATION Office of Nuclear Reactor Regulation

'EATONgH.R',

SUBJECT:

Requests deletion of- License Condition (3)(c)iper encl revised response to. Question 212,40 in App Q of FSAR, Rev'is.ion due to 'util'misinterpietation of- requirements re check valve leak testing,H/fee 8 affidavits DISTRIBUTION CODE: A001S COPIES RECEIVED:LTR J 'NCL: J SIZE:

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INDIANA II MICHIGAN POWER COMPANY Pi 6. BOX 18 BOWLING GREEN STATION NEw YoRK, N. Y. 10054 October 29, 1979 REP;NgC;Q0259.

Donald C..Cook Nuclear Plant Unit No. 2 Docket No. 50-316 License No. DPR-74 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S, Nuclear Regulatory Commission Washington, D.C. 20555 Dear Mr.

Denton:'urther review of Question 212.40 as contained in Appendix Q to the Donald C. Cook Nuclear Plant Final Safety Analysis Report' (FSAR) has led us to conclude that some of the testing described in the response is not necessary to satisfy the stated staff con-cerns and that the lists of valves need to be revised. The response to Question 212.40 was previously revised in our letter to Mr. Edson G. Case dated February 17, 1978. The intent of Question 212.40 is that we leak test the check valves which perform an isolation function

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of protecting low pressure safety systems from full reactor pressure.

The staff required that each check valve which performs this isolation function be identified and classified ASME IWV-2000 category AC with the leak testing being performed to code specifications. License condition (3) (c) was included in our Unit No. 2 operating license in accordance with the'ommitments made in our response to Question 212.40.

Our review has indicated that in the cases where low pressure systems are'isolated from full reactor pressure by check valves, the over-pressure protection of the low pressure system piping is provided by ASME code safety relief valves. As such, the check valve performs an isolation function but does not protect low pressure systems from full reactor pressure. Our misinterpretation of the staff position contained in Question 212.40 resulted in the commitments made in the response which became license condition (3) (c). The results of our review are con-tained in a revised response to Question 212.40 which is attached for your review. We request tha't operating license condition '(3) (c) be de-leted in accordance with the attached revision to Question 212.40.

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Nr. Harold R. Denton, Di'rector AEP;NRC:00259 This revision to the question 212.40 response does not inyolve an unreviewed safety question or Technical Specification change, nor it will endanger the health or safety of the public. We intend to formally incorporate this revised response into the FSAR as part of a future Amendment.,

Our review indicates that this revision constitutes a fee Class III Amendment to the facility license. In accordance with 10 CFR 170.22, we therefore enclose a check for $ 4,000,00.

Very truly yours, John E. Dol an Vice President cc:. R. C. Callen G. Charnoff D. V. Shaller-Bridgman R. S. Hunter RE W. Jurgensen

0 Res onse to uestion 212.40 There are no check valves which protect low pressure piping from full reactor pressure. This overpressure protection is provMed by safety relief valves on the low pressure piping systems as described below.

This response addresses the staff concern system by system. The design pressure of the boron injection system is higher than the design pressure of the Reactor Coolant System (RCS). Therefore the check valves in the boron injection system do not perform the function of protecting a low pressure system from full reactor pressure.

The function of protecting the Emergency Core Cooling Systems (ECCS) from fully reactor pressure is performed by safety relief valves. The ECCS lines to the RCS hot legs are isolated by normally closed valves.

The Residual Heat Removal normal cooldown line is isolated by normally closed valves. The check valves in the other ECCS lines perform an isolation function only to the extent that any leakage should not exceed the capacity of the associated safety valves. In each case, there are either two or three check valves in series between the RCS and the ECCS components with a lower pressure rating. These series check valves are listed in Table 212.40-1 along with the associated safety valves which protect the lower pressure systems. For each check valve, the

. rated capacity and pressuro setting of the associated safety valve(s) are adequate to protect the low pressure piping system. The allowable leakage rate for each listed check valve was determined, very con-servatively, based on the lowest relief capacity of the associated safety valve(s) and under the assumptions that all the other check valves in series are fully open and that all the other check valves in parallel leak at the maximum allowable rate.

The performance of the check valves in isolating the ECCS from full reactor pressure is tested at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during operational modes 1, 2, 3 and 4 by Technical Specification surveillance requirement 4.4.6.2d. to demonstrate that unidentified leakage from the RCS is limited to 1 gpm. Because this limit is well below the allowable leakage rate through any check valve, the adequacy of these check valves to perform thei'r isolation function is continuously verified by satisfaction of this survei'llance requirement. Because of this requirement, any gradual de-teri'oration of the check valve seats will be recognized and remedied.

These valves are located in systems that are normally maintained full of liquid, with either high pressure on the downstream side of the disc or no differential pressure across the disc. In this application, where th'e check valve is normally closed, any sudden, severe damage to the seating surface is very unlikely.

212.40-2

The test frequency for exercising the valves identified'n Table 212.40-1 is in accordance with ASI1E Section XI paragraph IW-3520 of the 1974 edition with addenda through the summer of 1975. These valves are normally closed during plant operation and cannot be exercised without initiating conditions similar to a safety injection. These valves will be exercised during cold shutdowns as stated in our Inservice Inspection Program submittals dated September 29, 1977 and September 22, 1978 (the latter resubmitted September ll, 1979.)

The design pressure of the Chemical and Volume Control System (CVCS) on the discharge side of the charging pumps is higher than the design pressure of the RCS. Therefore the discharge side of the CVCS does not require pro-tection from full reactor pressure. The suction side of the charging pumps is protected by the suction header safety relief valve. The CVCS reciprocating charging pump discharge check valve i s not required to perform a pressure isolation function because the construction of a multi-piston, positive dis-

. placement pump precludes pressure propagation in the reverse direction. The centrifugal charging pump discharge valves perform an isolation function only to the extent that any leakage should not exceed the capacity of the suction header safety relief valve. These check valves are listed in Table 212.40-2 along with the associated safety valve which protects the low pressure portion of the system. The pressure setpoints and relief flow capacity ratings for the safety valves are adequate to protect the low pressure piping system.

The allowable leakage rate was determined assuming that all four check valves leak at the maximum allowable rate and that there is no recirculation. However, during all modes of plant operation with the Reactor Coolant System above 220 psi, normal practice. is to have one charging pump running. Therefore, any leakage through the discharge check valve of a non-operating centrifugal charging pump is recirculated by the operating pump and does not cause a significant in-crease in the suction side pressure.

The testing for "exercising" will be performed for the check valves in Table 212.40-2 in the same manner and at the same frequency as described above for those in Table 212.40-1.

212,40-3

I TABLE 212.40 - 1 ECCS SERIES CHECK YALVES Alloivable Check Valve Leakage Rate:

Protecting Check Valve Nomenclature Safet Val ve s

• GPM SI151E ECCS Low Head Safety Injection SV-104E 400 SI151W ECCS Low Head Safety Injection SV-104W 400 SI152N ECCS Safety Injection SV-98A 20 SI152S ECCS Safety Injection SV-98B 20 SI161L1 SI Hot To Cold Leg Crosstie SV-98A 5 SV-104E 10 SI161 L2 SI Hot To Cold Leg Crosstie SV-98B 5 10 SY-104W SI161L3 SI Hot To Cold Leg Crosstie SY-98B 5 10 SV-104W SI161L4 SI Hot To Cold Leg Crosstie SV-98A 8 10 SV-104E SI166-1 Accumulator Discharge SV-100-1 47 SI166-2 Accumulator Discharge SY-100-2 47 SI166-3 Accumulator Discharge SV-100-3 47 SI166-4 Accumulator Discharge SV-100-4 47 SI170L1 ECCS Cold Leg Loop SV-98A, 10 SV-100-1 8(

SV-104E SI170L2 ECCS Cold Leg Loop SV-98B, 10 SV-100-2 &

SV-104M S I170L3 ECCS Cold Leg Loop SY-98B, 10 SV-100-3 5 SV-104W SI170L4 ECCS Cold Leg Loop SV-98A, 10 SY-100-4 8 SV-104E

".The Safety Valve designations are the same as those used in the Unit 2 ISI Program.

e TABLE 212.40 - 2 CVCS CENTRIFUGAL CHARGING PUMPS DISCHARGE CHECK VALVES Allowable Check Valve Protecting Leakage Rate Check Valve Nomenclature Safet Val ve s GPM CS299E Discharge SV-56 CS299M Discharge SV-56 CS297E Recirculation SV-56 CS297iI Recirculation SV-56

. Mr . Harold R. Denton, Director AEP:NRC:00259 STATE OF NEW YORK )

) ss.

COUNTY OF NEW YORK)

~ohn E Dolan, being duly sworn, deposes and says that he is the Vice President of licensees Indiana 8 Michigan Electric Company and Indiana 5 Michigan Power Company; that he has read the foregoing request and justificati'on for deletion of Condition (3) (c) on License No. DPR-74 and knows the contents thereof; and that said contents are true to the best of his knowledge and belief.

Subscribed and sworn to before me this 29th day of October, 1979.

Notary Public NOTA.,Y'yUobLcC, 5~co~ Ie ot liow Yock No. 4c-~~i" Gi 92 Queiifieo in 4ueens Courcy Cociiiicsiu fi!ed in iisw Ycck enoicoi cnecoh 30, 198 County'vccuno5con i

~y,fl REMI Wp UNITED STATES o NUCLEAR REGULATORY COMMISSION

I < WASHINGTON, D. C. 20555 0

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OCT 17 1979 IIEMN08It 8fgKg I[ ~opy All Power Reactor Licensees All Applicants With Applications for a License Gentlemen:

This past March, the NRC transmitted to you a copy of Volume 3 of NUREG-0460, "Anticipated Transients Without Scram for Light Water Reactors" (ATWS) and a copy of an NRC letter that was sent this past February to each of the four nuclear reactor vendors. The letters to the vendors contained requests for information needed to perform generic analyses related to ATWS.

As we pointed out in our March letters, the generic analyses we requested were intended to confirm that the modifications proposed by the NRC staff for. various classes of LWR designs would in fact accomplish the degree of ATWS prevention and mitigation described by the staff in its report. We also pointed out that we had chosen to work'directly with the vendors in obtaining this information in an effort to conserve both NRC and industry resources. We requested that utilities cooperate with the vendors in per-forming the requested analyses. II Shortly after sending the letters to the vendors, the NRC Staff met with representatives of each of the NSSS vendors and many Utility representa-tives in Bethesda on March 1, 1979. The meeting was called to discuss the "early verification" approach in which we planned to use generic analyses as the basis for rulemaking. We hoped thereby to avoid costly a~d unneces-sary repetitive analysis for individual plants. At the meeting, a, tenta-tive schedule was agreed to for generic analyses for each class pf plants to be provided in three separate packages to be submitted May l,ISeptember 1, and December 1, 1979.

I og yv 1979

<<y o o g ch 1 meeting, the NRC staff met separately with of h NSSS vendors and agreement i upplied in the May 1 package. Also, as note a ATWS stqff report and the generic analyses questions w Utilities.

• n I sl and acci dent occurre d . Because e of the heavy e uired for Three Ml i e Is 1 d ltd ed to the ATWS issue for three months o r t t 1 b t ti 1 d tio n effort on the part of the PWR industry during that perio,d an d o d tio fo BWR of Nuclear Reactor Regulation was temporarily reorganized. Within thi s interim organization a gr d d th direction of S. Hanauer to work on t e nre i nat-ed by the Commission and reported to Congress Con ress this past January in NUREG-0 - 510.

ATWS is one of these 19 issues.

'he A reliminary pre NRR Staff review su ested that, suggeste a , for PWRs,, the Three Mile Island land accident raised new questions w g

t. the technical impact of Three Mile s corn letion and review o e 1 for BWRs as specified in March should procee roceed as eexpeditiously as possible.

da on July 25, 1979 to discuss, with representa-v s o i ie e ig r considerations arising from the Three a copy of the staff minute tes of that meeting is a tt h d E lo 1 A can be seen from ro the minutes, at the meeting tin the staff:

is still believed by the staff to be a serious sa ey 0 " e p rotection should be provided. We stated that we are unwilling to wait anot er year on ATWS.

1 and s ecific technical concerns raised by the Three Mile Island accident with regard to thee ATWS resolution pro-posed in Volume 3 of NUREG-0460.

rovide in writing, within 30 days of the meeting ent of the Three Mile Island impact on ATWS, to resolve TMI issues, and a realistic roviding the needed ATW ATWS in orm both the March request an d thee TMI-related analyses.

QCT 17 1979 gsubsequent to the July 25 meeting, we have met with representatives of the four g NSSS vendors and of some Utility/Owners. We have met with GE to discuss the scope qf the remaining generic analysis information to be supplied for BWR 4/5/6's. We have also met with representatives of the GE BWR/3 Owners, B8W, BEW ATWS Owners Group, W, W ATWS Owners Group, and CE. At all these meetings, we considered further the required information and the schedule for its sub- .

mittal.

We have now received letters (see the list in Enclosure 2, attached) from the various groups describing the information to be furnished and projected schedules.

On the basis of our review of these letters and meetings with the industry representatives, we perceive that the projected responses in several cases would not address several questions in our ~February 15 letter. In particular, several items are lacking that we will need to justify acceptance of the hardware approaches of NUREG 0460 Vol 3 rather than using the design basis accident approach.

I am determined to submit a proposed ATWS rule to the Commission for both PWRs and BWRs early in 1980. The type and content of the rule we will propose will depend critically upon the types and content of the information available to the staff. This will, of course, include whatever responses are actually pro-vided by the industry in response to the questions attached to the February 15 staff 'letter, the March meetings, and the Three Mile Island related concerns as discussed in the July 25 and subsequent meetings.

I still believe that it is possible for the early verification generic analysis program to provide an acceptable resolution of'he ATWS issue and that this is the way to achieve resolution with the least possible expenditure of NRC and industry resources. However, I want to reiterate that the success of this approach depends on whether or not all of the information necessary for the staff to confirm that its proposed ATWS modifications provide an acceptable level of protection, for all plants, is provided by the industry.

I strongly encourage you to join or form Utility/Owners Groups, if you have not already done so, and provide the resources necessary to supply the needed tech-nical information pertaining to your plants, either operating or under construc-tion. It would further reduce the impact on the industry as well as the staff resources if the ATWS effort coordination and the review role is performed by one industry group, If you haye additional questions on the generic analysis early verification program discussed in this letter, please contact Mr. Ashok Thadani, (301-492-7341).

Sin ly, s

~

H. R, 'Denton,

~

~ Director Office of Nuclear Reactor Regulation

Enclosures:

1. NRC-Industry ATWS Meeting Summary dtd 7/25/79

2. List of letters from Industry on Content of Report Submittals

' 4, ENCLOSURE 1 g RECg

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tp 0 UNITED STATES NUCLEAR REGULATORY COMhlISSION WASHINGTON, D. C. 20555 y~ ~*4 JUL 2P 1379 Task Action Plan A-9 MENORANOUtl FOR: S. H. Hanauer FROM: A. Thadani

SUBJECT:

NRC-INDUSTRY ATWS tlEETING'UIlMARY Th e sataff met- with the PWR vendors, the Atomic Industrial Forum (AIF) and utility representatives to discuss the impac t of TMI-2 events me-w'everal on the ATWS resolution plan described in Volume 3 of NUREG-0460.

The staff'ade the following initial remarks:

1) ATWS is still a safety concern and protec ion from these events must be d d Alth h plants need not be shutdown immediately because of relatively low likelihood of a severe ATWS in a PWR in the nex p of years, ATMS resolution with suitable speed is necessary to permit an implementation plan which would assure an acceptably low risk from ATWS over the life of nuclear plants.

2) The staff would like to recei,ve industry views on the impact of TflI-2 on ATWS and how to proceed from now on to resolve ATWS. The staff noted that they intend to propose an ATMS solution to the Commission preferably with but if necessary without the industry input.

3) In view of TMI-2 accident, the staff expressed the following general con-cerns with the Vol. 3 proposed resolution and asked for industry comments.

a) What assurance do we have that the excessive calculated pressures for some designs modified per Alternative 83 would not result in loss of integrity of reactor coolant pressure boundary. (Note - Some designs may experience peak pressures - 4000 psi).

b) Would increasing the number of safety valves as per Alternative 84 result in insufficient overall risk reductionf Would the primary system integrity be maintained? Would it be better to have larger capacity valves'

S. H. Hanauer I

I c)) In v i ew 0 f que stions a and b above, the pressurizer relief and safety

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valves must be qualified for water relief to assure that th e nozzles, valve body and the support s .ructure integrity will be maintained

'he and to estimate discharge flow rate and the likelihood and effects of valve chatter.

d) I nve i w of significant plant differences in the designs of auxiliary feedwater system, Emergency Core Cooling Systems and other s ystems how would the industry provide assurances that plant specific

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f ea tures have been adequately addressed in the "Early Verification" approach for resolving ATMS as described in NUREG-0460, Vol. 3.

e) Other Lessons Learned from TNI-2.

Following prelim>nary comments from the NRC staff members, G. Sorensen of WPPS who is. also the Chairman of the AIF ATMS committee, made the following comments.

1) ATMS is not a safety issue but rather it is a licensing issue which needs resolution.

2) AIF in concert with the industry had reviewed ATMS in light of TMI-2 and had concluded that the Alternative 84 fix {mitigation) in Vol. 3 of NUREG-0460 is not the correct solution to ATMS. The industry believes that the alternative 82 fix {Prevention - Electrical Portion of RPS) is the appropriate ATMS solution.

3) Industry recognizes the THI-2 impact on the role of the operator, his training aids and other lessons learned from this event. The industry believes that there is no need to rush to resolve ATWS because of the low probability of ATMS and because some of the anticipated changes to plants as a result of TMI-2 accident review would direct resources to other issues.

Following the AIF presentation, the staff raised their concerns that the ATWS resolution {not yet achieved) gas been anything but hasty, that the NUREG docu-t ATHS have been out for sufficiently long time period, that protection from ATWS is necessary, that THI-2 H event has raised ""oncerns with the ana 1 y ses assumptions and therefore the htaff needs industry technical assessment of the TMI-2 impact on ATWS. The staff suggested that the THI-2 event indicates a need to answer at least the following specific questions.

S. H. Hanauer 1) Analyses indicate the sensitivity of peak pressure to AFWS design and actuation time for some plants.

Mhy should auxiliary feedwater actuation not be delayed beyond technical spehification values? What bases are available to assume AFWS actua-tion earlier than the technical specification value? How do the analyses take into consideration the limits on AFWS injection rate due to water

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hammer considerations? How is the impact of flow restrictors on some AFMS designs considered in the ATWS analyses? How are the significant plant specific features of AFWS treated in the analyses?

2) As in question 1 above how are the differences in ECCS designs evaluated?

For example, for some ATMS events, the pressure and the pressurizer level remain hiqh enough such that either the HPSI cannot be actuated (because of shut off head considerations) or the operator may fail to actuate HPSI because of insufficient available information.

3) Would single failure cause all PORVs to fail to open? If so, then analyses must be based on all PORVs failing to open. Further, several plants are operating today with PORVs isolated. For these plants credit cannot be taken for relieving capability of these valves.

4) What assurance do we have that the ATWS events with a stuck open safety

. valve have been correctly analyzed? What is the potential for core un-covering under this scenario? What is the importance of ECCS actuation, reactor coolant pumps operation, and the pressurizer safety/relief valve discharge model on the potential for uncovering of the core? Further, why should more valves not be assumed to stick open following discharge of subcooled water.

5) For long term shutdown, discuss the following:

a) available equipment, instrumentation and their qualification. (Must consider the effect of water discharged to the containment via ruptured quench tank).

b) impact of loss of offsite power c) continued operation 'of reactor coolant pumps. Also consider tripping of reactor coolant pumps.

d) Describe natural circulation, including effects of non-condensables.

Is reflux boiling mode of operation anticipated? If so, justify.

S. H. Hanauer 4-e) Would one anticipate Boron precipitatton problem? Also consider TMI-2 type problems with possible letdown line plugging from Boron precipitation.

f) How are leakage problems from equipment outside containment considered?

6) Why should credit be given for operator action even after ten minutes fallowing an ATWS event injtiation7 TMI-2 experience does not provide enoughconfidence in the ability of the operator to perform correct actions only in this short time period under high stress conditions.

In response.to the staff concerns the industry made the following comnents.

AIF

1) The industry is frustrated because the staff concerns imply consideration of multiple failures and small LOCA which are beyond the credible events to be considered under ATWS. (Note - safety valve stuck open (small LOCA) is considered an anticipated transient).

2) Industry would like to wait for approximately six months before consider-ing ATWS evaluations to minimize duplicate expenditures.

l) W has submitted responses to the 2/15/79 Mattson letter.

2) Calculated peak pressure of 2800 > 2900 psi (for Alt, 83) and proposed modifications in turbine trip and auxiliary feedwater system actuation ci rcui try.

3) EPRI expects to issue a request for proposal to conduct tests on PORVs and safety valves and some results should be available by end of CY 79.

4) Recommended that "Early Verification" approach should be continued.

CE - Ed Shearer speakin for himself

1) TMI raises few questions like the behavior of S/R valves and the operator action. Further, prevention is better than mitigation and that mitigation would mean more and more analyses.

2) Continue with early verification.

a S. H. Hanauer BlkW

1) Basically agrees with the staff concerns. Industry has longer list of items that could impact ATWS.

2) Stress analyses should be completed.

3) Likelihood of additional failures beyond ATWS should be considered.

4) Prevention is better than mitigation.

~BLfl 0 6

1) ATWS is not a safety pr.oblem.

2) Even if ATWS occurs, no significant risk to public health and safety.

3) TMI-2 suggests a desirability for realistic analyses. TMI-2 suggests a need to assure that analyses bound the facilities.

4)- Wait until "Lessons Learned" and "Bulletins and Orders" issues are resolved before pushing ahead with ATWS.

After the above industry comments, the staff made the following concluding remarks.

1) We don't intend to go too fast on ATWS.

2) If Early Verification is to be pursued then there is a need to assure that

- earlier ATWS analyses are correct and review the industry TMI-2 related list. In this regard the industry was invited to meet with the staff to discuss the technical issues which impact ATWS. The staff asked the indus-try to provide their assessment of TMI-2 impact on ATWS, the scope of I effort to resolve these issues, and the schedule for performing this effort within 30 days. I

3) We cannot wait another year to make progress in ATWS.

A. Thadani

Enclosure:

As stated cc: See next page

S. Hanauer cc: Meet)ng Attendees ATWS Distrfbut)on PDR RSB Files T. Spels

ENCLOSURE ATWS Meetin with Vendors & AIF July 25, 1979 Ashok Thadani NRC/DSS Arthur McBride B&W

.Alan Hosier WPPSS Samir K. Sarkar FP&L Alan E. Ladieu YAEC Fred T. Stetson A!F Richard G. Rateick DECO Andrew J. Rushnok OEC M. Srinivasan NRC/DSS F. Akstulewicz NRC/DSE G. Sorensen WPPSS/A IF T. Speis NRC/DSS F. C. Cherny NRC/DSS J. A. Norberg NRC/OSD Stuart Thickman TVA - EN DES Karl 0. Layer BBR J. Ted Enos AP&L Ted Myers TECo Robert Dieterick SMUD Michael J. Salerno CPCo S. Hardy Duerson B&W Bob Steither W Gary Augustine P. M. Abraham Duke Power Mark Wisenburg USTVA - Office of Power Michael Tokar NRC/DSS Paul Boehnert NRC/ACRS David Bessette NRC/ACRS Steven Traisman Pacific Gas & Electric Sam Miranda W Pat Loftus W Fred Mosby Wyl e Laboratory Roger Newton Wisconsin Electric Power Craig Grochmal Stone & Webster Charles A. Daverid Long Island Lighting Co.

Robert L. Stright SNUPPS Joseph M. Weiss GE Joseph A. Gonyeau Northern States Power

Seth M. Coplan NRC/OSE Clayton L. Pittiglio NRC/OSE

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Kulin D. Desai NRC/OSS Fuat Odar NRC/OSS Kris Par czewski NRC/DOR Roy Hoods NRC/DOR Harold Vander Molen ~ NRC/DOR Gururajarao Rangarao PASNY Frank McPhatter B&H Steve Banwarth BGH William R. Murray Virginia Electric 5 Power Co.

Ben Rodell VEPCO Don Swanson PGE Co.

Paul Y, Holton Bechtel Tommy Errington Mississippi Power E Light Ron Clauson Florida Power Corporation Charles B. Brinkman CE C. L. Kling CE William Benjamin Commonwealth Edison Co.

Denny Kreps CE Villiam E. Burchill CE A. E. Scherer CE Richard C. L. Qlson Baltimore Gas It Electric Co.

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ENCLOSURE 2 from R. H. Bucholz (GE) to S. Hanauer, "ATWS Generic Analyses-I'etter Content of December 1979 Submittal", dated September 5, 1979.

Letter from D. H. Taylor (BLM) to S. Hanauer, "B8W Commitments for ATWS", dated September 13, 1979.

Letter A. E. Scherer (CE) to S. Hanauer, "NRC Request for Generic ,

ATWS Information", dated August 31, 1979.

Letter L. 0. DelGeorge (BMR 3 Owners representative) to S. Hanauer, "ATWS BMR/3 Plants and Vermont Yankee - Generic Analysis Supplement",

dated August 28, 1979.

Letter T. N. Anderson (W) to S. Hanauer, "ATWS", dated August 24, 1979.

50-317 altinore Gas 6 Electric Company 50-318 CC:

Jan s A. Biddison, Jr. ttr. R. M. Douglass, Hanaoer G neral Counsel gual ity Assurance Depart>rent G and E Building Room 923 Gas 5 Electric Building Charles Center P. 0. Box 1475 Bal timore, tlaryl and 21203 Bal tirrare, Maryland .21203

. George F. Trowbridge, Esquire Shaw,. Pittman, Potts and Trowbridge 1800 N Street, tt,tt.

Washington, D. C. 20036

. ttr. R. C. L. Olson Baltimore Gas and'Electric Company Room 922 - G and E Building Post Office Box 1475 Bal timore, Maryland 21203 fir. Leon B. Russell, Chief Enqineer Calvert Cliffs I!uclear Power Plant Gas and Electric Company altinore Lusoy, ttaryland 20657

" Bechtel Power Corporation ATTH: tlr..J. C. Judd Chief Nuclear Engineer 15740 Shady Grove Road Gaithersburg, ttaryland 20760 Combustion Engineering, Inc.

ATTN: Nr. P. W. Kruse, Hanager Engineering Services Post Of fice Box 500 t"indsor, Connec ticut 06095 Cal vert County Library Prince Frederick, Maryland 20678