ML17256B067

Improvements in Training & Requalification Programs as Required by TMI Action Items I.A.2.1 & II.B.4

ML17256B067
Person / Time
Site:	05000000, Ginna
Issue date:	06/24/1982
From:	SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY
To:	NRC
Shared Package
ML17256B066	List: ML17256B065 ML17256B067 ML17256B068
References
CON-NRC-03-82-096, CON-NRC-3-82-96, RTR-NUREG-0737, RTR-NUREG-737, TASK-1.A.2.1, TASK-2.B.4, TASK-TM SAI-186-029-11R, SAI-186-29-11R, NUDOCS 8206280425
Download: ML17256B067 (20)

Text

SAI-186-029-11R TECHNICAL EVALUATION REPORT IMPROVEMENTS IN TRAINING AND REQUALIFICATION PROGRAMS AS REQUIRED BY TMI ACTION ITEMS I.A.2.1 AND II.B.4 for the R. E. Ginna Nuclear Power Plant (Docket 50-244)

June 24, 1982 Prepared By:

Science Applications, Inc.

1710 Goodridge Drive McLean, Virginia 22102 Prepared for:

U.S. Nuclear Regulatory Commission Washington, D.C.

20555 820b280425 820b24

,'DR ADOCK 05000244

,P PDR Contract NRC-03-82-096 Xi Science Applications, Inc

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TABLE OF CONTENTS Section III

~Pa e

INTRODUCTION..

1 SCOPE AND CONTENT OF THE EVALUATION.

1 A..

I.A.2.1:

Immediate Upgrading of RO and SRO Training and gualifications......

1 B.

II.B.4:

Training for Mitigating Core Damage..

6 LICENSEE SUBMITTALS.................

7 IV EVALUATION.

A.

I.A.2.1:

Immed iate Training

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8 Upgrading of RO and SRO and gualifications......

8 V.

B.

II.B.4:

Training for Mitigating Core Damage.

11 CONCLUSIONS 11 VI.

REFERENCES.

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I.

INTRODUCTION Science Applications, Inc. (SAI),

as technical assistance contractor to the U.S. Nuclear Regulatory Commission, has evaluated the response by Rochester Gas and Electric Corporation (RGEE) for the R.

E.

Ginna Nuclear Power Plant (Docket 50-244) to certain requirements contained in post-TMI Action Items I.A.2.1, Immediate Upgrading of Reactor Operator and Senior Reactor Operator Training and gualification, and II.B.4, Training for Mitigating Core Damage.

These requirements were set forth in NUREG-0660 (Reference 1) and were subsequently clarified in NUREG-0737 (Reference 2).*

The purpose of the evaluation was to determine whether the licensee's operator training and requalification programs satisfy the requirements.

The evaluation pertains to Technical Assignment Control (TAC)

System numbers 44164 (NUREG-0737, I.A.2.1.4) and 44514 (NUREG-0737, II.B.4.1).

As delineated

below, the evaluation covers only some aspects of item I.A.2.1.4.

The detailed evaluation of the licensee's submittals is presented in Section IV; the conclusions are in Section V.

II.

SCOPE AND CONTENT OF THE EVALUATION A.

I.A.2.1:

Immediate Upgrading of Reactor Operator and Senior Reactor Operator Training and gualifications The clarification of TMI Action Item I.A.2.1 in NUREG-0737 incor-porates a letter and four enclosures, dated March 28, 1980, from Harold R.

Denton, Director, Office of Nuclear Reactor Regulation, USNRC, to all power reactor applicants and licensees, concerning qualifications of reactor operators (hereafter referred to as Denton's letter).

This letter and enclosures imposes a number of training requirements on power reactor licensees.

This evaluation specifically addressed a subset of the require-ments stated in Enclosure 1 of Denton's letter, namely:

Item A.2.c, which relates to operator training requirements; item A.2.e, which concerns instructor requalification; and Section C, which addresses operator requali-fication.

Some of these requirements are elaborated in Enclosures 2,

3, and 4 of Denton's letter.

The training requirements under evaluation are sum-marized in Figure 1.

The elaborations of these requirements in Enclosures 2,

3 and 4 of Denton's letter are shown respectively in Figures 2,

3 and 4.

As noted in Figure 1, Enclosures 2 and 3 indicate minimum require-ments concerning course content in their respective areas.

In addition, the Operator Licensing Branch in NRC has taken the position (Reference

3) that

• Enclosure 1 of NUREG-0737 and NRC's Technical Assistance Control System distinguish four sub-actions within I.A.2.1 and two sub-actions within I I.B.4.

These subdi v is i ons are not carri ed for ward to the actual presentation of the requirements in Enclosure 3 of NUREG-0737. If they had been, the items of concern here would be contained in I.A.2.1.4 and II.B.4.1.

Figure 1.

Training Requirements from THI Action Item I.A.2.1*

Program Element NRC Requirements'*

OPERATIONS PERSONNEL TRAINING Enclosure I. Item A.2.c(l)

Training programs shall be modified. as necessary, to provide training in heat transfer, fluid flow and thermodynamics.

(Enclosure 2 provides guidelines for the mininum content of such training.)

Enclosure I, Item A.2.c(2)

Training programs shall be modified, as necessary to provide training in the use of 1nstalled plant systems to control or mitigate an accident in which the core is severely damaged.

(Enclosure 3 provides guidelines for the minimum content of such training.)

Enclosure I, Item A.2.c. (3)

Training programs shall be modified, as necessary to provide increased emphasis on reactor and plant trans1ents.

INSTRUCTOR REQUALIFICATION Enclosure I, Item A.2.e Instructor s shall be enrolled in appropr1ate requalification programs to assure they are cognizant of current operating history, problems, and changes to pro-cedures and administrat1ve limitations.

PERSONNEL REQUALIFICATION Enclosure I, Item C. I Content of the licensed operator requalif5cation programs shall be modif5ed to in'elude instruction 5n heat transfer, fluid flow, thermodynamics, and m5tiga-tion of accidents involving a degraded core.

(Enclosures 2 and 3 provide guide-11nes for the minimum content of such training.)

Enclosure I, Item C.2 The criteria for requiring a licensed individual to participate in accelerated requalification shall be mod1fied to be consistent with the new passing grade for issuance of a license:

80g overall and 70'>> each category.

Enclosure I, Item C.3 Programs should be mod1f5ed to require the control manipulations listed 5n.

Normal control manipulations, such as plant or reactor startups, must be performed.

Control manipulations during abnormal or emergency opera-tions must be walked through with, and evaluated by, a member of the training staff at a minimum.

An appropriate simulator may be used to satisfy the requirements for control manipulations.

'The requirements shown are a subset of those contained 1n Item I.A.2.l.

• 'References to Enclosures are to Oenton's letter of Harch 28, 1980, which is contained in the clarifi-cation of Item I.A.2.I in NUREG-0737.

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Figure 3.

Enclosure 3 from Denton's Letter A.

Incore Instrumentation TRAINING CRITERIA FOR MITIGATING CORE DAMAGE l.

Use of fixed or movab'Ie incore detectors to determine extent of core damage and geometry changes.

2.

Use of thermocouples in determining peak temperatures; methods for extended range readings; methods for direct readings at terminal 5unctions.

3.

Methods for calling up (printing) incore data from the plant computer.

8.

Excore Nuclear Instrumentation NIS l.

Use of NIS for determination of void formation; void location basis for NIS response as a function of core temperatures and density changes.

C.

Vital Instrumentation I.

Instrumentation response in an accident environment; failure sequence (time to failure, method of failure); indication reliability (actual vs indicated level).

2.

Alternative methods for measuring flows, pressures,

levels, and temperatures.

a.

Oetermination of pressurizer level if all level transmitters fail.

b.

Oetermination of letdown flow with a clogged filter (low flow).

c.

Determination of other Reactor Coolant System parameters if the primary method of measurement has failed.

I.

Expected chemistry results with severe core damage; consequences of transferring small quantities of liquid outside containment; importance 'of using leak tight systems.

2.

Expected isotopic breakdown for core damage; for clad damage.

3.

Corrosion effects of extended iamersion in primary water; time to failure.

E.

Radiation Monitorin 1.

Response

of process and Area Monitors to severe damages; behavior of detectors when saturated; method for detecting radiation readings by direct measurement at detector output (overranged detector);

expected accuracy of detectors at different locations; use of detectors to determine extent of core damage.

2.

Methods of determining dose rate inside containment from measurements taken outside containment.

F.

Gas Generation I.

Hethods of H2 generation during an accident; other sources of gas (Ke, Ke); techniques for venting or disposal of non-condensibles.

2.

H2 flammability and explosive limit; sources of 02 in containment or Reactor Coolant System.

Figure 4.

Control Manipulations Listed in Enclosure 4.

CONTROL MANIPULATIONS

<<1.

Plant or reactor startups to include a range that reactivity feedback from nuclear heat addition is noticeable and heatup rate is established.

2.

Plant shutdown.

• 3.

Manual control of steam generators and/or feedwater during startup and shutdown.

4.

Boration and or dilution during power operation.

<<5.

Any significant (greater than 10') power changes in manual rod control or recirculation flow.

6.

Any reactor power change of 10'r greater where load change is performed with load limit control or where flux, temperature, or speed control is on manual (for HTGR).

• 7.

Loss of coolant including:

1.

significant PNR steam generator leaks 2.

inside and outside primary containment 3.

large and small, including leak-rate determination 4.

saturated Reactor coolant response (PNR).

8.

Loss of instrvxent air (if simulated plant specific).

9.

Loss of electrical power (and/or degraded power sources)

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Loss of core coolant flow/natural circulation.

11.

Loss of condenser vacuum.

12.

Loss of service water if required for safety.

13.

Loss of shutdown c'ooling.

)4.

Loss of component cooling system or cooling to an individual component.

15.

Loss of normal feedwater or normal feedwater syste~ failure.

• 16.

Loss of all feedwater (normal and emergency)

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Loss of protective system channel.

18.

Mispositioned control rod or rods (or rod drops).

19.

Inability to drive control rods.

20.

Conditions requiring use of emergency boration or standby liquid control system.

21.

Fuel cladding failure or high activity in reactor coolant or offgas.

22'urbine or generator trip.

23.

Halfunction of automatic control system(s) which affect reactivity.

24.

Malfunction of reactor coolant pressure/volume control system.

25.

Reactor trip.

26.

Hain steam line break (inside or outside containment).

27.

Nuclear instruaentation failure(s).

'tarred items to be performed annually, all others biennially.

the training in mitigating core damage and related subjects should consist of at least 80 contact hours* in both the initial training and the requali-fication programs.

The NRC considers thermodynamics, fluid flow and heat transfer to be related

subjects, so the 80-hour requirement applies to the combined subject areas of Enclosures 2

and 3.

The 80 contact hour criterion is not intended to be applied rigidly; rather, its purpose is to provide greater assurance of adequate course content when the licensee's training courses are not described in detail.

Since the licensees generally have their own unique course out-

lines, adequacy of response to these requirements necessarily depends only on whether it'is at a level of detail comparable to that specified in the enclosures (and consistent with the 80 contact hour requirement) and whether it can reasonably be concluded from the licensee's description of his train-ing material that the items in the enclosures are covered.

The Institute of Nuclear Power Operations (INPO) has developed its own guidelines for training in the subject areas of Enclosures 2 and 3.

These guidelines, given in References 4 and 5, were developed in response to the same requirements and are more than adequate, i.e., training programs based specifically on the complete INPO documents are expected to satisfy all the requirements pertaining to training material which are addressed in this evaluation.

The licensee's'esponse concerning increased emphasis on tran-sients is considered by SAI to be acceptable if it makes explicit reference to increased emphasis on transients and gives some indication of the nature of the increase, or, if it addresses both normal and abnormal transients (without necessarily indicating an increase in emphasis) and the requalifi-cation program satisfies the requirements for control manipulations, Enclo-sure 1, Item C.3.

The latter requirement calls for all the manipulations listed in Enclosure 4 (Figure 4 in this report) to be performed, at the frequency indicated, unless they are specifically not applicable to the licensee's type of reactor(s).

Some of these manipulations may be performed on a simulator.

Personnel with senior. licenses may be credited with these activities if they direct or evaluate control manipulations as they are performed by others.

Although these manipulations are acceptable for meet-ing the reactivity control manipulations required by Appendix A paragraph 3.a of 10 CFR 55, the requirements of Enclosure 4 are more demanding.

Enclosure 4 requires about 32 specific manipulations over a two-year cycle while 10 CFR 55 Appendix A requires only 10 manipulations over a two-year cycle.

B.

II.B.4:

Training for Mitigating Core Damage Item II.B.4 in NUREG-0737 requires that "shift technical advisors and operating personnel from the plant manager through the operations chain to the licensed operators".receive training on the use of installed systems to control or mitigate accidents in which the core is severely damaged.

• A contact hour is a one-hour period in which the course instructor is present or available for instructing or assisting students;

lectures, seminars, discussions, problem-solving sessions, and examinations are considered contact periods.

This definition is taken from Reference 4.

Enclosure 3 of Denton's letter provides 'guidance on the content of this training.

"Plant Manager" is here taken to mean the highest ranking manager at the plant site.

For licensed personnel, this training would be redundant in that it is also required, by I.A.2.1, in the operator requalification program.

However, II.B.4 applies also to operations personnel who are not licensed and are not candidates for licenses.

This may include one or more of the highest levels of management at the plant.

These non-licensed personnel are not explicitly required to have training in heat transfer, fluid flow and thermodynamics and are therefore not obligated for the full 80 contact hours of training in mitigating core damage and related subjects.

Some non-operating personnel, notably managers and technicians in instrumentation and control, heal th physics and chemistry depar tments, are supposed to receive those portions of the training which are commensurate with their responsibilities.

Since this imposes no additional demands on the program itself, we do not address it in this evaluation.

It would be appropriate for resident inspectors to verify that non-operating personnel receive the proper training.

The required implementation dates for all items have passed.

Hence, this evaluation did not address the dates of implementation.

Moreover, the evaluation does not cover training program modifications that might have been made for other reasons subsequent to the response to Denton's letter.

III.

LICENSEE SUBMITTALS The licensee (RGRE) has submitted to NRC a

number of items (letters and various attachments) which explain their training and requalification programs.

These submittals, made in response to Denton's

letter, form the information base for this evaluation.

For the Ginna plant, there were four submittals with attachments, for a total of eight items, which are listed below.

The last three items were in a combined submittal in response to a request for additional information prepared by SAI, dated February 24,

1982, and transmitted by NRC to the licensee in a letter dated March 23,,

1982.

1.

Letter from L.D. White, Jr.,

Yice President Rochester Gas E Electric Corporation, to D.M.

Crutchfield, Chief of Operating Reactors Branch k5, NRC. August 25, 1980.

(1 pg, with enclosure:

item 2).

NRC Acc.

No: 8009020086.

(re:

05 additional TMI-2rel ated requirements).

2.

"Response to NRC letter dated March 28, 80 (guali-fications of Reactor Operators)".

R.E.

Ginna Power Plant, Unit No. 1, Rochester Gas 5 Electric Corp.

(3 pp, attached to item 1).

3.

Letter from L.D. White, Jr.,

Vice President Rochester Gas 8 Electric Corporation, to P.F.

Collins, Chief of Operator Licensing Branch, NRC.

August 22, 1980.

(1 pg, with enclosure:

item 4).

NRC Acc No:

8009050274 (Transmittal) 4.

"R.E.

Ginna Operator Requalification Program",

Rochester Gas 8 Elec. Corp.,

Ginna Station, Pro-cedure No: A-102.14, Rev.

No: 3. Approved for use by the Plant Superintendent, May 21, 80.

(13 pp, attached to item 3).

NRC Acc No:8009050278 5.

Letter from J.E. Maier, Rochester Gas E Electric Corporation, to D.M. Crutchfield, Chief of Operat-ing Reactors Branch 5'5, NRC.'arch 13,81.

(1 pg).

NRC Acc No:

8103250242.

(re: Training for mitigating, core

damage, NUREG-0737).

6.

Letter from J.E. Maier, Vice President, Electric and Steam Production, Rochester Gas 5 Elect.

Corp.,

to D. M. Crutchfield, Chief of Operating Reactors Branch k5, NRC.

April 6, 82.

(8 pp, with enclo-sures:

item 7

E 8).

NRC Acc No: 8204190066.

(re:

Response

to RAI letter dated 03/23/82, concerning NUREG-0737, items I.A.2.1, and II.B.4).

7.

"Westinghouse Mitigating Core Damage" R.E. Ginna Power Plant, Unit No.l.

Undated (10 pp, attached to item 6).

8. 'Introduction to Physics, Thermodynamics, Fluid 5 Fluid Flow Principles" Ginna,

1980, Revision 3,

06/25/81.

(4 pp, attached to item 6).

IV. EVALUATION SAI's evaluation of the training programs at Rochester Gas and Electric Corporation's R.

E.

Ginna"Nuclear Power Plant is presented below.

Section A addresses TMI Action Item I.A.2.1 and presents the assessment organized in the manner of Figure 1.

Section B addresses TMI Action Item II.B.4.

A.

I.A.2.1:

Immediate Upgrading of Reactor Operator and Senior Reactor Operator Training and qualifications.

Enclosure 1

Item A.2.c 1

The basic requirements are that the training programs given to reactor operator and senior reactor operator candidates cover the subjects of heat transfer, fluid flow and thermodynamics at. the level of detail specified in Enclosure 2 of Dehton's letter.

The submittal of RGhE which addressed item A.2.c.(l) (submittal item

2) stated that the licensing training program (Administrative Procedure A-102.13) was revised on May 2, 1980, to include the necessary topics.

No further details were avail able

until RGEE submitted additional information (submittal items 6

and 8).

In these submittals RGEE stated that their level of instruction was comparable with Enclosure 2 of Denton's letter.

They also provided a table of contents for their instruction "Introduction to Physics, Thermodynamics, Fluid and Fluid Flow Principles."

The table of contents does not have one-for-one correspondence with the items of Denton's Enclosure

2. It does,

however, have a moderate level of detail and outlines a program which would appear to contain all of the required material.

Enclosure 1

Item A.2.c 2

The requirements are that the training programs for reactor and senior reactor operator candidates cover the subject of accident mitigation at the level of detail specified in Enclosure 3 of Denton's letter (see Figure 3 of this report).

SAI has examined the submittals of RGEE and has found that most of the elements identified in Enclosure 3 of Denton's letter are explicitly identified in the response to SAI's request for additional information.

A few items are not explicitly identified, these being "alter-nate measurement methods,"

"gas generation,"

and "corrosion." It seems reasonable to expect that these unidentified items are covered in the training program because (1) they are explicitly identified in the Westing-house submittal to NRC of July 23, 1980 (Reference 6), dealing with accident mitigation, and (2)

RGEE has arranged for Westinghouse to teach an accident mitigation training course for Ginna personnel through August 1981.

This analysis suggests that the requirements were met through August 1981.

Assuming this or a similar program has been used since August 1981, the Ginna Training program still meets the NRC requirements.

RGKE does not describe the extent of their training programs in terms of "contact hours," but rather in terms of days.

We estimate that the training in mitigating core damage and related subjects (including heat transfer, fluid flow and thermodynamics) involves in excess of 100 contact hours.

In view of NRC's criterion for 80 contact hours, we take this as further evidence that the training programs at Ginna satisfy NRC's require-ments regarding course content and level of detail.

Enclosure 1

Item A.2.c 3

The requirement is that there be an incr eased emphasis in the training program on dealing with reactor transients.

The submittal of RGEE dated August 25, 1980, (submittal item 2) addressed this requirement by saying that the training program was revised to meet the requirement.

No additional details were provided in this particular submittal.

In a later submittal (submittal item 6),

RGRE stated that two additional weeks of training are involved.

This increase in program length is associated with an increase in scope, the most significant change being an increase in simulator use.

Enclosure 1

Item A.2.e The requirement is that instructors for reactor operator training programs be enrolled in appropriate requalification programs to assure they are cognizant of current operating

history, problems and changes to procedures and administrative limitations.

The RGEE submittal of August 25, 1980 stated that the administrative procedure A-102.14 was revised to

accommodate the requirement.

The procedure itself stated that instructors shall participate in a program to keep them current in plant changes which include procedure

changes, current operating history, current R.E. Ginna LERs and other relevant LERs.

Enclosure 1

Item C.l The primary requirement is that the requalification programs have instruction in the areas of heat transfer, fluid flow, thermodynamics and accident mitigation.

The level of detail required in the requalification program is that of Enclosures 2 and 3 of Denton's letter.

In addition, these instructions must involve an adequate number of contact hours.

RG5E's submittal of August 25, 1980 (submittal item 2) stated that the requirement for including these materials in the requalification program was met by modifying the requalification program (Administrative Procedure A-102.14).

The program defined in A-102.14 listed lectures on "heat transfer, fluid flow and thermodynamics" and "mitigating core damage during accidents."

No further details were available until RGEE submitted a

response on April 6, 1982 (submittal items 6 and 8).

In this response, RG&E attached the table of contents for the instructions given in these two general areas.

The details were the same as discussed and analyzed previously in items A.2.c.(1) and A.2.c.(2).

Any judgment made about technical adequacy for those items would be applicable for this item.

Esti-mates of the contact hours involved with this item have been made based on information supplied by RG&E.

The estimate is that 88 contact hours are involved which is greater than the necessary number of hours according to this NRC criterion.

Enclosure 1

Item C.2 The requirement for licensed operators to participate in the accelerated requalification program must be based on passing scores of 805

overall, 70K in each category.

According to the submittal of August 25, 1980 (submittal item 2),

the licensee has been judged to meet the require-ment.

Enclosure 1

Item C.3 TMI Action item 1.A.2.1 calls for the licensed operator requalifi-cation program to include performance of control manipulations involving both normal and abnormal situations.

The specific manipulations required and their, performance frequency are identified in Enclosure 4 of the Denton letter (see Figure 4 of this report).

The RGRE submittal of August 25, 1980 (submittal item 2) stated that Administrative Procedure A-102.14 had been modified to meet the requirement.

In procedure A-102.14, all of the appropriate Enclosure 4

manipulations are included with control manipulation titles similar or identical to those of Enclosure 4.

The frequency of the manipulation performance is also compatible with the requirements of Enclosure 4.

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B.

II.B.4 Training for Mitigating Core Damage Item II.B.4 requires that training for mitigating core

damage, as indicated in Enclosure 3 of Denton's letter, be given to shift technical advisors and operating personnel from the plant manager to the licensed operators.

This

~ includes both licensed and non-licensed personnel.

With regard to content, training for mitigating core damage as required by II.B.4 is the same as that required by I.A.2.1 under Enclosure 1

items A.2.c(2) and C.l.

The licensee meets this aspect of the requirements.

They also satisfy the 80 contact hour requirement for licensed personnel.

Another requirement relative to accident mitigation training is that it be given to all operating personnel from the plant manager down to the licensed operators and also to the plant technical advisors.

Again, based on information supplied by RGEE in their response to SAI's request for information, it appears that this requirement is satisfied at the Ginna plant.

Specifically, this training is given to personnel holding the following positions:

plant superintendent, assistant superintendent, opera-tions engineer, operations. supervisor, shift supervisor, head control opera-tor, control operator, technical assistant for operational assessment, and shift technical advisor.

V.

CONCLUSIONS Based on SAI's evaluation as discussed

above, we conclude there is reasonable assurance that the training programs at the R.E. Ginna Nuclear Power Plant meet the requirements of NUREG-0737:I.A.2.1 as delineated in Section II of this report; and of NUREG-0737:II.B.4.

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VI.

REFERENCES 1.

"NRC Action Plan Developed as a Result of the TMI-2 Accident."

NUREG-

0660, United States Nuclear Regulatory Commission.

May 1980.

2.

"Clarification of TMI Action Plan Requirements,"

NUREG-0737, United States Nuclear Regulatory Commission.

November 1980.

3.

The NRC position regarding the requirement for 80 contact hours is an informal one.

It was included with the acceptance criteria provided by NRC to SAI for use in the present evaluation.

See letter, Harley Silver, Technical Assistance Program Management Group, Division of Licensing, USNRC to Bryce.Johnson, Program

Manager, Science Applic'ations, Inc.,

Subject:

Contract No. NRC-03-82-096, Final Work Assignment 2,

December 23, 1981.

4.

"Guidelines for Heat Transfer, Fluid Flow and Thermodynamics Instruction," STG-02, The Institute of Nuclear Power Operations.

December 12, 1980.

5.

"Guidelines for Training to Recognize and Mitigate the Consequences of Core Damage,"

STG-01, The Institute of Nuclear Power Operations.

January 15, 1981.

6.

Letter, J. J.

Evans, Manager, Westinghouse Electric Corporation Nuclear Training Services, to Paul Collins,

Chief, Operator Licensing

Branch, Division of Reactor Licensing, Nuclear Regulatory Commission, with attachments A through I, July 23, 1980.

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