ML20038A902

Forwards Containment Sys Branch Draft Evaluation Rept on SEP Topic VI-4 Re Containment Isolation Sys.Requests Info Re Defined Basis Upon Which Specific Isolation Configurations Were Judged to Be Acceptable

ML20038A902
Person / Time
Site:	Oyster Creek
Issue date:	11/18/1981
From:	Crutchfield D Office of Nuclear Reactor Regulation
To:	Finfrock I JERSEY CENTRAL POWER & LIGHT CO.
References
TASK-06-04, TASK-6-4, TASK-RR LSO5-81-11-042, LSO5-81-11-42, NUDOCS 8111240419
Download: ML20038A902 (43)

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November 18, 1981 Docket No. 50-219 LS05-81 11-042 e d,. Mytc lT 3198V*' 2 O

uMMoW Mr. I. R. Finfrock, Jr.

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Vice President - Jersey Central Power & Light Company Post Office Box 388 Forked River, New Jersey 08731 w

SUBJECT:

FORWARDING DRAFT EVALUATION REPORT OF SEP TOPIC VI-4, CONTAINMENT ISOLATION SYSTEM FOR THE OYSTER CREEK NUCLEAR POWER PLANT Enclosed is a copy of our draft evaluation of SEP Topic VI-4, Containment Isolation System. This assessment compares your facility, as described in Docket No. 50-219, with the criteria currettly used by the regulatory staff for licensing new facilities.

Please inform us if your as-built facility differs from the licensing basis assumed in our assessment.

In addition, I would like to draw your attention to two of the more significant issues contained in the conclusion, location of both isolation valves outside containment and use of a simple check as an isolation valve outside containment. Both of these items appear to contradict the explicit wording of the regulations and no other acceptable defined basis could be determined from the information provided.

To enable us to perform our assessment of the deviations identified in this report, we will need the defined basis upon which the specific isolat6on configurations at the Oyster Creek Plant were judged to be accept-able by you.

Please provide this information as a part of your comments on this report.

P Coments are required within 30 days of receipt of this letter so that they may be included in our final report. This evaluation will be a l

basic input to the integrated safety assessment for your facility unless you identify changes needed to reflect the as-built conditions at your l

facility. This assessment may be revised in the future if your facility design is changed or if NRC criteria relating to this subject are modified SAcl Add: A"gg before the integrated assessment is completed.

Sincerely, 1'((

psa aSR M(St) i 8111240419 81111'g7 DR ADOCK 05000219 Dennis M. Crutchfield, Chief j

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November 18, 1981 gv Docket No. 50-219 L505-81-ll-042 Mr. I. R. Finfreck, Jr.

Vice President - Jersey Central Power & Light Company Post Office Box 388 Forked River, New Jersey 08731

SUBJECT:

FORWARDING DRAFT EVALUATION REPORT OF SEP TOPIC VI-4, CONTAINMENT ISOLATION SYSTEM FOR THE OYSTER CREEK NUCLEAR POWER PLANT Enclosed is a copy of our draft evaluation of SEP Topic VI-4, Containment Isolation System.

This assessment compares your facility, as described in Docket No. 50-219, with the criteria currently used by the regulatory staff for licensing new facilities.

Please inform us if your as-built facility differs from the licensing basis assumed in our assessment.

In addition, I would like to draw your attention to two of the more significant issues contained in the conclusion, location of bcth isolation valves outside containment and use of a simple check as an isolation valve outside containment.

Both of tnese items appear to contradict the explicit wording of the regulations and no other acceptable defined basis could be determined from the information provided.

To enable us to perform our assessment of the deviations identified in this report, we will need the defined basis upon which the specific isolation confirgurations at the Oyster Creek Plant were judged to be accept-a bl 0 by you.

Please provide this information as a part of your comments on this report.

Comments are required within 30 days of receipt of this letter so that they may be included in our final report.

This evaluation will be a basic input to the integrated safety ar,sessment for your facility unless you iderlify change: needed to reflect the as-built conditions at your facility.

This assessment my be revised in the future if your facility design is changed or if NRC criteria relating to this subject are modified before the inte-grated assessment is completed.

Sincerely, W

q Dennis M. Crutchfield, Chief

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Operating Reactors Branch No. 5 Division of Licensing

Enclosure:

As stated

Mr. I. R. Finfrock, Jr.

0YSTER CREEK Docket No. 50-219 cc G. F. Trowbridge, Esquire Gene Fisher' Shaw, Pittman, Potts and Trowbridge Bureau Chief 1800 M Street, N. W.

Bureau of Radiation Protection Washington, D. C.

20036 380 Scotts Road Trenton, New Jersey 08628 J. B. Lieberman, Esquire Berlack, Israels & Lieberman Commissioner 26 Broadway New Jersey Department of Energy New York, New York 10004 101 Commerce Street Newark, New Jersey 07102 Natural Resources Defense Council 91715th Street, N. W.

Licensing Supervisor Washington, D. C.

20006 Oyster Creek Nuclear Generating Station J. Knubel P. O. Box 388 BWR Licensing Manager Forked River, New Jersey 08731 GPU Nuclear 100 Interplace Parkway Resident Inspector Parsippany, New Jersey 070$4 c/o U. S. NRC P. O. Box 445 Deputy Attorney General Forked River, New Jersey 08731 State of New Jersey Department of Law and Public Safety 36 West State Street - CN 112 Trenton, New Jersey 08625 Ocean County Library Brick Township Branch 401 Chambers Bridge Road Brick Town, New Jersey 08723 Mayor Lacey Township 818 Lacey Road Forked River, New Jersey 08731 Commissioner Department of Public Utilities State of New Jersey 101 Commerce Street Newark, New Jersey 07102 U. S. Environmental Protection Agency Region II Office l

ATTN: Regional Radiation Representative

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l 26 Federal Plaza New York, New York 10007 l.___.___.

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I TABLE OF CONTENTS 1

I INTRODUCTION II REVIEW CRITERIA III RELATED SAFETY TOPICS I V. '

REVIEW GUIDELINES Y

EVALUATION VI CONCLUSIONS VII REFERENCES l

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i ENCLOSURE l

o Containment Systems Branch Evaluation Report on SEP Topic VI-4 Containment Isolation System for the Oyster Creek Nuclear Power Plant, Unit 1 Docket No. 50-219 I

Introduction The Oyster Creek Nuclear Power Plant, Unit 1 (Oyster Creek) began com-mercial operation in 1969. Since then the safety review criteria have changed and, as part of the Systematic Evaluation Program (SEP), the con-tainment isolation systems at Oyster Creek have been re-evaluated. The y,urpose of this evaluation is to document the deviations from the current safety criteria as they relate to the containment isolation systems. The significance of the identified deviations, and recommended corrective meas-ures to improve safety, will be the subject of a subsequent integrated as-sessment of Oyster Creek.

II Review Criteria The safety review criteria used in the current evaluation of the contain-l ment isolation system for Oyster Creek are contained in the following references:

1-10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants (GDC 54, 55, 56 and 57).

2-NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants (SRP 6.2.4, Containment Isolation System).

3-Regulatory Guide 1.11, Instrument Lines Penetrating Primary Reactor Containment.

4-Regulatory Guide 1.141, Revision 1, Containment Isolation Provisions for Fluid Systems.

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. III Related Safety Topics 1

The review areas identified below are not covered in this report, but are related and essential to the completion of.the re-evaluation of the containment isolation system for Oyster Creek. These review areas are included in other SEP topics or ongoing Generic Reviews, as indi-cated below:

1 III-1, Classification of Structures, Components and Systems (Seismic and Quality) 2 - III-4.C. Internally Generated Missiles 3 - III-5.A, Effects of Pipe Break on Structures, Systems and Cte.ponents Inside Containment 4 - III-5.B, Pipe Break Outside Containment 5 - III-6, Seismic Design Considerations 6

III-12, Environmental Qualification of Safety Related EQuipnent 7

VI-6, Containc. ant Leak Testing 8

VII-2, Engineered Safety Feature System Control Logic and Design 9 - VIII-2, Onside Energency Power Systems - Diesel Generator 10 - VIII a, Electric Penetrations of Reactor Containment 11 - VJREG-0737, Clarification of TMI A: tion Plan Requirements, Item i

II.E.4.2, Containrent Isolation Dependacility 12 - NUREG-C560, NRC Action Plan Developed as a Result of the TMI-2 Acci-dent, Ite: II.E.4.4, Containment Purging and Venting Requi.s nts I

13 - NUREG-0803, Generic Safety Evaluatfion Report Regarding Integrity of B'4R Scram System Piping.

IV Review Guidelines _

The containment isolation syste of a nuclear power olant is an engineere:!

safety feature that functions to allow the norcal or emergency passage of fluids through the containment boundary while preserving the ability of the boundary to prevent or limit the escape of fission products to the environs that may result from postulated accide.nts. General Design criteria 54, 55, 56, and 57 of Appendix A'to 10 CTR Part 50 pertain to the containment iso-lation systr of a nuclear power plant'.

General Design Criterion 54. establishes design and test requirements for the leak detection provisions. the 1' solation function and the containment l

capability of the isolation barriers in lines penetrating the primary re-actorcontainment.

Fram the standpoint of containment isolation, leak de-tection provisions should be capable of quickly detecting and responding l

to a spectrum of postulated pipe break accident conditions. To accomplish

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this, diverse parameters should be sonitored to initiate the containment isolation function. The parameters selected should assure a positive, rapid response. to the developing accident condition. This aspect of the coitainment isolation system review will be addressed during the review of the post-TMI requirements approved for implementation, as stated in NUREG-l l

0737 Item II.E.4.2.

l Leak detectic.) capability should also be provided at the system level to alert the operator of the need to isolate a system train equipped with re-1 mote manual isolation valves. The Standard Review Plan 6.2.4, Item II.B.q, provides guidance in this regard.

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t With respect to the design' requirements for the isolation function, all non-essential systems should be aLtomatically isolated (with manual valves sealed closed), and valve closure times should be selected to assure rapid isolation of the containment in the event of an accident. The review of the classification of systems as essential or non-essential, and the auto-matic isolation provisions for non-essential systems by appropriate signals,.

will be addressed in conjunction with the review of the post-TMI require-ments as stated in NUREG-0737, Item II.E.4.2.

The closure time of the con-tainment ventilation system isolation valves will be evaluated in conjunction with the ongoing generic review of purging practices at operating plants (see NUREG-0660, Item II.E.4.4).

The electrical power supply, instrumentation and control systems should be designed to engineered safety feature criteria to assure accomplishment of the containrent isolation function. This aspect of the review is covered under SEP Topics VII-2 and VIII-2. Also, resetting the isolation signal should not result in the automatic re-opening of containeent isolation valves. This will be addressed in conjunction with the review of the post-TMI requirements approved for implementation, as stated in NUREG-0737, Item II.E.4.2.

With respect to the capabilities of containment isolation barriers in lines penetrating primary containment, the isolation barriers should be designed to engineered safety feature criteria, and protected against missiles, pipe whip and jet impingement. Typical isolation barriers include valves, closed systems and blind flanges.

Furthermore, ymvisions should be made to permit periodic leak testing of the isolation barriers.

. V The adequacy of the missile, pipe whip and jet impingement protection will be covered under SEP Topics III-4.C. III-5. A and III-5.B. :The ac-ceptability of the design criteria originally used in the design of the ccotainment isolation system components will be covered in SEP Topics III-1, III-6 and III-12.

The adequacy of the leak testing. program will be covered under SEP Topic VI-6. The acceptability of electrical penetrations will be covered in SEP Topic VIII-4.

General Design Criterie 55, 56 and 57 establish explicit requirements for isolation valving in lines penetrating the contaiment. Specifically.

they ' address the number and location of isolation valves (e.g., redundant valving with one loc'ated inside containment and the other located outside containment), valve actuation provisions (e.g., automatic or remote manual isolation valves). valve position (e.g., locked closed, or the position of greater safety in the. event of an accident er power failure), and valve tyoe (e.g., a simple check valve is not a pennissable automatic isolation valve out-side containment).

Figures 1 and 2 depict the explicit valve arrangements spe-cified in GDC 55 and 56, and GDC 57, respectively.

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'GDC 55 and 56 also pernit containment isolation provisions for lines pene-trating the primary contaiment boundary that differ from the explicit re-quirements, provided the basis for acceptability is defined. This proviso is typically invoked when establishing the containment isolation require-ments for essential (f.e. safety related) systems, or there is a clear 1s-prevement in safety.

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. The Standard Review Plan 6.2.4, Item II.6 presents guidelines for acceptable alternate containment isolation provisions for certain classes of lines. Con-tainment isolation provisions that are found acceptable on the "other defined basis" represent conformance with the GDC and do not constitute exceptions.

The following evaluation addresses deviations in the containment isolation provisions from the explicit' requirements of the General Design Criteria.

GENERAL DESIGN _ CRITERIA 55 AN) 56 ISO _ATON VA_VE CTER'A MISSILE PROTECTION CONTAINMENT INSIDE OUTSIDE INSIDE OUTSIDE

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f 7-V Evaluation The containnent isolation provisions for the lines penetrating the primary reactor containment of Oyster Creek are listed in Table 1.

This information was obtained from the documents and piping and instru-mentation drawings referenced in Section VII. There was insufficient information to. complete certain elements of Table 1, therefore, the licensee is requested to provide the udssing information and make any necessary corrections.

The containnent isolation provisions, as listed in Table 1, were evalu-ated against the requirement of General Design Criteria (GDC) 54, 55, 56 and 57 ( Appendix A to 10 CFR Part 50), and the supplementary guidance of the Standard Review Plan (SRP), Section 6.2.4, Containment Isolation System, where applicable. Deviations from the explicit requirements of GDC 54, 55, 56 and 57, and the acceptance criteria of SRP 6.2.4 are tabu-lated in Table 2.

The evaluation of containment system deviations from the current licensing criteria is best summarized by listing the areas of non-conformance as follows:

1-Insufficient administrative control; 2-Insufficient leak detection capability on remote manual valves in ESF systems; 3-Use of check valves as isolation valves outside of containment; 4-Used of local menual valves as isolation valves; 5-Use of remote manual isolation valves outside containment on non-ESF systems; and 6-Use of two automatic isolation valves outside containment on non-ESF systems.

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. Administrative Control All test, vent, drain and sampling lines between the inboard and outboard isolation valves shall be sealed closed barriers, which may be used in place of automatic isolation valves in accordance with SRP 6.2.4.11.6.

Sealed closed barriers include blind flanges and sealed closed isolation valves which may be closed manually, closed remote manual valves, and closed automatic valves which remain closed after a LOCA. However, with respect to the test, vent, drain and sampling lines, pipe caps are not suitable isolation barriers; two locked closed isolation valves in series should be provided for these lines.

In any case, sealed closed isolation valves should be under administrative contrcl to assure that they cannot be inad-vertently opened. Administrative control includes mechanical devices to seal or lock the valve closed, or prevent power from being supplied to the valve operator.

The following list of test, vent, drain and sampling lines at Oyster Creek deviate from the explicit requirements of GDC 55 and 56 from the standpoint of valve type. The valves on these lines will be acceptable if they are under adninistrative control as specified in the SRP Section 6.2.4.11.6.

This list was compiled using the availaole piping and instrumentation draw-ings as well as Reference 6. These references indicate that manv of the valves on the test, vent, drain and sampling lines are normally closed, however, there is no indication that they are under administrative control.

Valves Required to be Under Administrative Control venetration No.

Valve Identification X-3B V-14-21 X-3B V-14-11

Valves Required to be Under Administrative control Penetration No.

Yalve Identification X-3A V-14-26 X-3A V-14-27 X-SA V-14-28 X-5B V-14-39 X-7.

V-17-21 X-7 V-17-65 X-7 Y-17-76 X-8 V-17-51 X-8 V-17-52 X-8 V-17-66 X-8 V-17-68 X-8 V-17-83 X-9 Y-16-4 X-lb V-16-65 X-12B V-20-42 X-15 Y-6-394 l

X-16 V-38-7 X-22 Y-38-8 X-61 V-15-31 X-61 Y-15-32 X-66 V-21-66 X-70 V-20-44 X-72 V-1-136 X-72 V-1-137 X-7A V-14-75 X-74 Y-14-76

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Leak Detection SRP Sections 6.2.4.II.6.b and c state that containment isolation provi-sions for lines in engineered safety feature or engineered safety-related systems may include remote-manual valves, but provisions should be made to detect possible leakage from these lines outside containment. Also, isolation provisions for lines in systems needed for the safe shutdown of the plant may include remote-manual valves, but again provisions should be made to detect possible leakage from these lines outside con-tainment.

The containment isolation provisions for the lines identified above nor-mally consist of one isolation valve inside and one isolation valve o':t-side containment. If it is not practical to locate one isolation valve inside containment both isolation valves may be located outside contain-ment. For this type of isolation valve arrangement, the valve nearest the containment and the piping between the containment and the valve should be enclosed in a leak-tight or controlled leakage housing.

If, in lieu of a housing, conservative design of the piping and valve is assumed to preclude a breach of piping integrity, the design should conform to the i

requirements of SRP Section 3.6.2 (SEP Topic III-1). In any event, the design of the valve and/or piping compartment should provide the capa-bility to detect leakage from the valve shaft and/or bonnet seals and terminate the leakage.

The following list of essential systems in which remote manual valves are located outside containment deviates from the explicit requirements of GDC 55 and 56 from the standpoint of location, actuation or both (see Table II), but will be acceptable if leak detection provisions are in accordance with SRP Section 6.2.4.II.b and c.

It appears from the l

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Itcensee's response to our questions that there are no leak detection provisions on these systems at the present time.

Systems Requiring Leak Detection Capabilities System Penetration Yalve Number Emergency Condenser System X-3A V-14-30 X-3B Y-14-33 X-SA V-14-35 X-5B V-14-34 Containment Spray X-22 Y-21-13 X-22 Y-21-17 X-63 V-21-11 Containment Spray to Torus X-51 V-21-18 X-51 V-21-15 Containment and Core Spray X-68A V-21-1 Suction X-69B V-21-3 X-69 V-21-7 V-20-9 V-20-32 Y-20-3 V-20-4 V-20-33 Check Valves as Isolation Valves GCG 55 and 56 specify that one valve should be located inside contain-ment and one valve should be locaed outside containment, and that a simple check valve may not be used as an automatic isolation valve out-side containment.

The following list presents those systems in which the use of a check valve deviates from the explicit requirements of the GDC from the stand-point of location and type of valve in use.

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Systems with Check Valves Outside Containment System Penetration Valve Number Feedwater X-4A V-2-72 X-4B V-2-71 Containment Spray X-22 V-21-19 Pump Test Line X-22 V-21-20 Cleanup Demineralize Relief System X-74 V-16-84 The feedwater penetrations X-4A and X-4B contain simple check valves as isolation valves outside of containment. An acceptable isolation bar-rier for this penetration would consist of the simple check valve outside of containment along with an added remote manual valve.

The containment spray pump test line discharge into the torus through simple check valve located on penetrations X-22, while the cleanup de-mineralizer relief system discharges through a simple check valve into the torus through penetration X-74.

These isolation valves are properly located outside the torus due to the possibility of pool swell loads on these valves if they were inside the torus. However, a judgment regard-ing the acceptability of the simple check valve outside containment t a bonafide containment isolation valve will be made in conjunction with the integrated assessment of the plant.

Local Manual Isolation Valves l

GDC 55 and 56, as it relates to lines that penetrate the primary contain-ment boundary and either are part of the reactor coolant pressure boundary I

or connect directly to the containment atmosphere, requires that isolation

valves be either automatic inside and outside containment, check inside and automatic outside or sealed closed inside and outside containment.

The containment isolation provisions that differ from the explicit re-

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quirements of GDC 55 and 56 may be acceptable on some otner defined basis, if the basis for that difference is justified.

For example, Regulatory Guide 1.11 describes acceptable containment isolation provi-sions for instrument lines. The staff does not believe that local man-ual valves are justifiable alternatives to the provisions of GDC 55 and 56 and, therefore, these systems should be automatically isolated.

The following list presents those systems in which the use of a local manual valve deviates from the explicit requirements of the GDC from the standpoint of type of valve in use.

Systems with Local Manual Valves Outside Containment System Penetration Valve Number Demineralized Water X-23 V-12-60 Torus Level X-49 V-20-63 X-51 V-20-62 Drywell Pressure X-50 V-38-2 X-50 V-38-3 Torus Pressure X-56 V-38-1001 Remote Manual Isolation Valves The use of remote manual valves as containment isolation valves deviates from the explicit requirements of GDC 55 and 56, however, is permitted on engineered safety systems or safety related systems, and on systems

needed for the safe shutdown of the plant, or on closed systems outside of containment.

The staff believes that the use of remote manual vales on the systems listed below deviates from the explicit requirements of the GDC, since these are non-essential systems using remote manual valve actuation outside of containment.

In order to be in conformance with the GDC, these valves should be automatic' ally isolated.

Non-Essential Systems with Remote Manual Valves outside of containment System Penetration Valve Number Reactor Shutdown Cooling X-7 V-17-1 Supply X-7 V-17-2 X-7 Y-17-3 Reactor Shutdown X-8 Y-17-55 X-E Y-17-56 X-8 Y-17-57 Reactor Head Cooling X-62 V-31-2 Twc Automatic Isolation Valves Outside Containment The use of two isolation valves outside containment represents a devi-ation frem the explicit requirements of GDC 55 and 56, but is permitted on ESF or ESF-related systems by SRP Section 6.2.4.II.6.d.

Provisions must be made to detect leakage from these lines outside containment.

The following list present those systems which have two automatic valves outside of containment and which are not classified ESF or ESF-related systems. These non-essential systems, therefore, deviate from the ex-plicit requirements of the GDC with respect to location, and in order to

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be in conformence with the GDC should have one automatic isolation valve inside of containment.

Non-Essential Systems With Two Automatic Valves Outside Containment System Penetration Yalve Number Drywell Oxygen Sample X-16 V-38-9 X-16 Y-38-10 Drywell Oxygen Sagle X-57.

V-38-23 X-57 V-38-22 Containment Particulate X-22 V-38-16 Monitor X-22 V-38-17 VI Conclusions The following sumarizes the deviations from review guidelines that have been identified and described in Section V of this report:

1.

The isolation valving arrangement of the following containment penetrations differ from the explicit requirements of GDC 55 and l

56 from the standpoint of valve type: Penetrations X-3B, X-3A, X-SA, X-5B, X-7, X-8, X-9, X-10, X-12B, X-15, X-16, X-22, X-61 and X-66.

The valves identified with these penetrations in Section V are required to be under administrative control.

2.

The isolation valving arrangement of the following containment penetrations differs from the explicit requirements of GDC 55 and 56 from the standpoint of location, actuation or both:

X-3A, X-38, X-SA, X-5B, X-22, X-63, X-51, X-68A, X-698, and X-69.

The valves identified with these penetrations in Section V are asso-ciated with essential systems and will be acceptable if leak

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3.

The isolation valving arrangement of the following penetrations differs from the explicit requirements of GDC 55 and 56 from the standpoint of location and valve in use:

X-4A, X-48, X-22 and X-74.

The valves identified with these penetrations in Section V are simple check valves outside of containment. A judgment re-garding the acceptability of a simple check valve outside of con-tainment as a bonafide containment isolation valve will be made in cor, Junction with the integrated assessment of the plant.

4.

The isolation valving arrangement of the following penetrations differs from the explicit requirements of GDC 55 and 56 from the standpoint of valve type in use:

X-23, X-49, X-50, X-51 and X-56. The valves identified with these penetrations in Section V are local manual valves, and in order to be in conformance with the GDC should be automatic isolation valves.

5.

The isolation valving arrangement of the following penetrations l

differs from the explicit requirements of GDC 55 and 56 from the I

standpoin t of valve actuation:

X-7, X-8, and X-62.

The valves identified with these non-essential systems identified in Section V are remote manual valves outside of containment, and in order to be in conformance with the GDC should be automatically isolated.

l 6.

The isolation valving arrangement of the folicwing penetrations differs from the explicit requirements of GDC 55 and 55 from the standpcint of location:

X-16, X-57, and X-22.

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. identified with these non-essential systems in Section V are two automatic isolation valves outside of containment; in order to be in conformance with the GDC should have one automatic isola-tion valve inside of containment.

VII References 1.

.0yster Creek Nuclear Power Plant, Unit 1, Safety Analysis Report, Volume 2.

Oyster Creek Nuclear Power Plant, Technical Specification 3.

NRC letter of April 17, 1981 from D. M. Crutchfield to I. R.

Finfrock 4.

JCP&L letter of September 23, 1980 from I. R. Finfrock to Director, NRR, " Technical Specification Change Request No. 90.'

5.

JCP&L letter of April 10, 1980 from I. R. Finfrock to D. G. Eisenhut, NRC, "NUREG-0578 Implementation."

6.

JCP&L letter of January 4,1980 from I. R. Finfrock to D. G. Eisenhut, NRC, "NUREG-0567.Caplementation."

7.

JCP&L letter of December 14, 1979 from I. R. Finfrock tow. Kane, NRC,

" Bulletins and Orders Task Force, Long Term Systems Information."

8.

JCP&L letter of November 20, 1979 from I. R. Finfrock to W. Kane, NRC,

" Bulletins and Orders Task Force, Long Term Systems Information."

9.

JCP&L letter of August 9,1979 from I. R. Finfrock to D. Fremon, NRC, "IE Bulletin 7-08."

i 10.

JCP&L letter of April 25, 1979 from I. R. Finfrock to B. Grier, NRC, j

"IE Bulletin No. 79-08."

l

. 11.

Dyster Creek Plant, Piping and Instrumentation Drawinas a.

Core Spray System - GE 885 0781 b.

Cleanup Demineralizer System - GE 148F444 c.

Reactor Head Cooling System - GE 886D403 d.

Neutron Monitoring System - GE 148F734 e.

Control Rod Drive Hydraulic System - GE 237E487 f.

Reactor Vessel Level / Pressure / Temperature Instruments - GE 148F712 9

Radwaste Treatment Plant - GE 148F437 h.

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