ML20055A241

Safety Evaluation Report Related to the Operation of Midland Plant,Units 1 and 2.Docket Nos.50-329 and 50-330.(Consumers Power Company)

ML20055A241
Person / Time
Site:	Midland
Issue date:	06/30/1982
From:	Office of Nuclear Reactor Regulation
To:
References
NUREG-0793, NUREG-0793-S01, NUREG-793, NUREG-793-S1, NUDOCS 8207150635
Download: ML20055A241 (50)

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i NUREG-0793 Supplement No.1 l

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Safety Evaluation Report related to the operation of Midland Plant,

} Units 1 and 2 1

Docket Nos. 50-329 and 50-330 Consumers Power Company U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation June 1982 p s "cas,,

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NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the following sources:

1. The NRC Public Document Room,1717 H Street, N.W.

Washington, DC 20555

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Referenced documents available for inspection a :d copying for a fee from the NRC Public Docu-ment Room include NRC correspondence and internal NRC memoranda; NRC Office of Inspection and Enforcement bulletins, circulars, information notices, inspection and investigation notices; Licensee Event Reports; vendor reports and correspondence; Commission papers; and applicant and licensee documents and correspondence. {

The following documents in the NUREG series are available for purchase from the NRC/GPO Sales Program: formal NRC staff and contractor reports, NRC-sponsored conference proceedings, and .

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GPO Ponted copy price. 54.75

1 NUREG4793 Supplement No.1

Safety Evaluation Report related to the operation of Midland Plant, Units 1 and 2 Docket Nos. 50-329 and 50-330 l

Consumers Power Company U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation l June 1982

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l ABSTRACT This report supplements the Safety Evaluation Report, NUREG-0793, issued May 1982 by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission with respect to the application filed by Consumers Power Company, as i applicant and owner, for licenses to operate the Midland Plant, Units 1 and 2 (Docket Nos. 50-329 and 50-330). The facility is located in the city of I Midland in Midland County, Michigan. This supplement provides recent infor-mation regarding resolution of some of the open items identified in the Safety Evaluation Report and discusses recommendations of the Advisory Committee on Reactor Safeguards in its interim report dated June 8, 1982.

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l TABLE OF CONTENTS Page ABSTRACT ............................................................. iii l

1 INTRODUCTION AND GENERAL DISCUSSION ............................. 1-1 l

( 1.1 Introduction ............................................... 1-1 1.8 Confirmatory Issues ........................................ 1-1 3 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS .......... 3-1 1

3.5 Missile Protection ......................................... 3-1 3.5.1 Missile Selection and Description ................... 3-1 3.5.1.3 Turbine Missiles ........................... 3-1 3.5.2 Structures, Systems, and Components To Be Protected From Externally Generated Missiles ................ 3-1

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4 REACTOR ........................................................... 4-1 4.2 Fuel System Design ......................................... 4-1 4.2.3 Design Evaluation ................................... 4-1 ,

l 4.2.3.3 Fuel Coolability Evaluation ................ 4-1 5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS ...................... 5-1 5.4 Component and Subsystem Analysis ........................... 5-1 5.4.4 Decay Heat Removal System ........................... 5-1 5.4.4.2 Cold Shutdown Capability ................... 5-1 5.5 Design Sensitivity of B&W Reactors ......................... 5-1 5.5.6 Main Feedwater Overfill ............................. 5-1 6 ENGINEERED SAFETY FEATURES ............. ........................ 6-1 6.3 Emergency Core Cooling System .............................. 6-1 6.3.4 Performance Evaluation .............................. 6-1 6.3.4.1 Natural Circulation During Small-Break LOCA.. 6-1 Midland SSER 1 v

TABLE OF CONTENTS (continued)

Page 13 CONDUCT OF OPERATIONS ........................................... 13-1 13.3 Emergency Planning ......................................... 13-1 13.3.1 Introduction ....................................... 13-1 13.3.2 Emergency Plan Evaluation........................... 13-1 13.3.3 Conclusion.......................................... 13-8 15 ACCIDENT ANALYSIS ............................................... 15-1 15.2 Secondary System Transients and Accidents .................. 15-1 15.2.3 Decrease in Heat Removal ............................ 15-1 15.3 Reactor Coolant Transients and Accidents ................... 15-1 15.3.1 Decrease in Reactor Coolant Flow Rate ............... 15-1 19 REVIEW BY ADVISORY COMMITTEE ON REACTOR SAFEGUARDS............... 19-1 APPENDIX A CONTINUATION OF CHRONOLOGY APPENDIX B BIBLIOGRAPHY  :

APPENDIX D ABBREVIATIONS APPENDIX E NRC STAFF CONTRIBUTORS (

APPENDIX G ACRS INTERIM REPORT ON MIDLAND PLANT, UNITS 1 and 2 APPENDIX H ERRATA TO MIDLAND PLANT SAFETY EVALUATION REPORT l

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Midland SSER 1 vi 1

l 1 INTRODUCTION AND GENERAL DISCUSSION 1.1 Introduction On May 11, 1982, the Nuclear Regulatory Commission staff (NRC staff or staff)

( issued a Safety Evaluation Report, NUREG-0793, regarding the application by Consumers Power Company (the applicant) for licenses to operate the Midland Plant, Units 1 and 2. This report is the first supplement to that Safety Eval.uation Report (SER). -

l This supplement provides more recent information regarding resolution of some of the open items identified in the SER. This supplement also provides and discusses the recommendations of the Advisory Committee on Reactor Safeguards (ACRS) in its interim report on the Midland Plant, dated June 8, 1982.

Each of the following sections or appendices of this supplement is numbered the same as the section or appendix of the SER that is being updated, and the dis-cussions are supplementary to and not in lieu of the discussion in the SER un-less otherwise noted. Accordingly, Appendix A is a continuation of the chro-nology of the safety review. Appendix B is an updated bibliography.* Appendix 0 is a list of abbreviations used in this supplement. Appendix E is a list of t principal contributors to this supplement. Appendix G is a copy of the ACRS f

report. Appendix H is a list of errata for the SER. No changes in SER Appen-dices C and F have been made by this supplement.

The Project Manager is Darl S. Hood; he may be reached on (301) 492-8474. Mr.

Ronald W. Hernan also serves as Project Manager, and may be reached on (301) 492-8395.

1.8 Confirmatory Issues Section 1.8 of the SER noted that certain confirmatory information had not yet been provided by the applicant for several identified items. This supplement updates some of those items for which the confirmatory information has subse-quently been provided by the applicant and for which review has been completed by the staff. These items, and the sections of this supplement discussing its review conclusions, are (5) Supplemental ECCS Calculations (4.2.3.3)

(9) Adequacy of BWST To Provide Boric Acid to RCS (5.4.4.2)

(10) Main Feedwater Overfill Protection (5.5.6)

(29) Steam Generator Water Inventory as a Function of Power Level (15.2.3)

(30) Loss-of-Flow Transients (15.3.1)

Additionally, SER confirmatory item (28), " Applicability of Power Train Code,"

is deleted by the errata page provided in this supplement (Appendix H). As

• Availability of all material cited is described on the inside front cover of this report.

Midland SSER 1 1-1

discussed in SER Section 15.1.2, the staff concluded that reasonable assuratice exists regarding the results from the POWERTRAIN computer code for the Midland Plant on the basis of audit calculations performed for the staff by Argonne National Laboratory.

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3 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS 3.5 Missile Protection 3.5.1 Missile Selection and Protection 3.5.1.3 Turbine Missiles Section 3.5.1.3 of the SER noted that the General Electric turbine generators are unfavorably oriented , and that additional information on the turbine-missile hazards was under review.

l The information provided by the applicant includes an evaluation of the turbine-missile risk for Hidland Plant, Units"1 and 2. Based on this analysis, which uses General Electric calculated probabilities for the generation of missiles from desi per year and 5.0 x 10 gn J andyear per destructive overspeed respectively, failure of of the probability 8.7unacceptable x 10 9 damage for Unit 1 is 1.4 x 10 5 per year; for Unit 2 it is 1.5 x 10 9 per year.

The applicant contends that his turbine inspection and test programs are either explicitly or implicitly incorporated in the evaluation and justify his use of j the General Electric missile generation probabilities. The staff has requested that the relevant General Electric analyses be submitted for its review to verify the adequacy of the applicant's turbine inspection and test programs.

This open item remains unresolved and will be discussed in later supplements pending receipt and review of the General Electric analyses.

3.5.2 Structures, Systems, and Components To Be Protected From Externally ,

Generated Missiles SER Section 3.5.2 stated that the applicant was not proposing to provide spe-cific tornado-missile protection for certain areas of the plant because of the low probability of missile damage resulting from the inherent protection by other plant structures. The SER identified certain items in these areas that were not protected. These items are portions of cabling for the component cooling water system and the chemical addition system, the auxiliary building fuel-handling bridge, certain ventilation systems, and specific penetrations in the south end of the auxiliary building.

Since issuance of the SER, the staff has made a site visit on May 20, 1982 and also has met with the applicant on June 11, 1982 to discuss the unprotected items in more detail. At the June 11 meeting the applicant provided the details of a deterministic study of the inherent tornado-missile protection afforded the identified items by virtue of their location and protection by other plant structures. The applicant also provided the results of a preliminary probabi-listic study based on Electric Power Research Institute reports NP-768 and NP-769 dated May 1978.

Midland SSER 1 3-1

The staf f will review the applicant's documentation of the information discussed at the June 11, 1982 meeting once it is received. The staff anticipates that its review will be completed by early 1983. The staff has investigated the risk to the public of allowing operation for a limited time without tornado-missile protection to these limited areas. Based on the details provided at the meet-ing, and on observations made during the site visit which confirmed the inherent tornado-missile protection provided for the items listed above, the staff con-cludes that specific tornado-missile protection need not be provided before the first refueling outage, but shall be provided no later than January 1,1985.

In addition to the specific items identified above, the diesel fuel oil lines between the underground diesel fuel oil storage tanks and the diesel generator building are buried only 2 to 3 ft under ground. These depths do not provide adequate missile protection. The applicant has verbally committed to provide concrete tornado-missile shielding. The staff requires the applicant to for-mally document the missile protection to be furnished for these li~nes. This protection should be in place before the operating license is issued.

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4 REACTOR 4.2 Fuel System Design 4.2.3 Design Evaluation 4.2.3.3 Fuel Coolability Evaluation (3) Cladding, Ballooning, and Flow Blockage I

I Section 4.2.3.3(3) of the Midland SER discusses the models for cladding rupture, cladding ballooning, and flow blockage integrally employed in the Babcock &

Wilcox (B&W) emergency core cooling system (ECCS) evaluation model and used to demonstrate conformance to the ECCS acceptance criteria of 10 CFR 50.46. This SER section noted that because of deficiencies in these models, B&W had pro-vided by report 12-1132424 a supplemental ECCS calculation using the cladding l models of NUREG-0630. The calculation also drew on compensating margins from  ;

the use of models of TACO-2, REFLOD 3, and THETAl-B, all three of which are currently undergoing NRC review. The supplemental ECCS calculation demon-strated that the Midland Plant would meet the ECCS acceptance criteria if a penalty of 1 kW/ft were applied to the linear heat generation rate at rupture-node-affected elevations. Consequently, the SER specified new peak linear heat I rate limits for the Midland Technical Specifications that reflected this L penalty.

The SER considered this issue confirmatory pending endorsement and formal sub-mittal by the applicant of B&W report 12-1132424. The applicant has submitted and endorsed this report. The staff considers this issue to be closed.

Midland SSER 1 4-1

t 5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.4 Component and Subsystem Analysis 5.4.4 Decay Heat Removal System

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5.4.4.2 Cold Shutdown Capability Section 5.4.4.2 of the SER discussed manual action outside the control room associated with achievement of cold shutdown in accordance with Branch Tech-nical Position RSB 5-1. The SER noted the applicant's statement that in the .

absence of a stuck control rod, adequate boric acid could be provided to the  !

reactor coolant system using water from the borated water storage tank supplied by a high pressure injection pump aligned from the control room. The SER indicated, as a confirmatory issue, that the staff had asked tne applicant to provide an analysis of this capability. This analysis has subsequently been provided and indicates that adequate shutdown margin is maintained. This confirmatory issue, identified as Item 9 in SER Section 1.8, is, therefore, closed.

5.5 Design Sensitivity of B&W Reactors 5.5.6 Main Feedwater Overfill Section 5.5.6 of the SER discussed the proposed main feedwater overfill protec-tion that would reduce the severity of overcooling from main feedwater and pro-tect the steamline and turbine from potential damage resulting from the admis-sion of water into the steamline. For the Midland design, the overfill protec-tion signal would close the main feedwater isolation valves but would not trip the main feedwater pumps or the reactor. Thus, a main feedwater overfill signal would terminate the overfill transient, but at the same time would initiate a loss-of-feedwater transient. The steam generators would begin to boil dry and the reactor would trip on high pressure. The auxiliary feedwater system would be initiated automatically on low steam generator level. Conser-vative analyses for loss of feedwater, discussed in SER Section 15.2.3, demon-strate that both the core and the reactor coolant pressure boundary would be adequately protected, although the safety valves are predicted to be challenged in the conservative analyses. The SER noted, as a confirmatory issue, that the staff had requested an evaluation demonstrating that the primary system safety valves would not be challenged by the !oss of feedwater resulting from main feedwater isolation valve closure. This issue also was identified as Item 10 in SER Section 1.8.

The additional analysis requested by the staff has been provided and indicates that if best-estimate conditions are present, the power-operated relief valve (PORV) would be challenged, but the safety valves would not. The Midland Plant is protected against the occurrence of a stuck open PORV by a safety grade iso-lation system. Failure of the PORV to open would probably produce a challenge to the safety valves, which are not provided with isolation protection.

Midland SSER 1 5-1

On the basis of the safety grade design of the PORV and its isolation block valves, and the best-estimate calculations, the staff concludes that the occurrence of a loss-of-coolant accident resulting from a stuck-open PORV or safety valve resulting from initiation of the main feedwater overfill protec-tion system is an unlikely event at Midland. The staff will consider the operating history of PORV and safety-valve challenges during the first fuel cycle to determine if additional protection such as an anticipatory reactor trip is warranted. Challenges to the PORV and safety valves will be reported by the applicant as required by Item II.K.3.3 of NUREG-0737. This confirmatory issue is, therefore, closed.

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6 ENGINEERED SAFETY FEATURES 6.3 Emergency Core Cooling System 6.3.4 Performance Evaluation 1

6.3.4.1 Natural Circulation During Small-Break LOCA The staff's review of the applicant's calculations for postulated small-break LOCA events was presented in SER Section 6.3 and concluded that the Midland design conforms to the fuel-damage criteria of 10 CFR 50.46 for these events.

It was also stated that the staff previously concluded that the B&W small-break l LOCA evaluation model used in these calculations conforms to the requirements of Appendix K to 10 CFR 50.

In the high point vent discussions of SER Section 5.4.7, the staff noted that for certain small-break sizes of 0.01 f t2 and less, natural circulation would be required to remove decay heat from the primary system. It was further noted that natural circulation could be temporarily lost for these break sizes but that it would be reestablished before the core could become uncovered. These calculations are documented in B&W report, " Evaluation of Transient Behavior l and Small Reactor Coolant System Breaks in the 177-Fuel Assembly Plant," May 7, 1979. These calculations demonstrate that the core will be adequately cooled as long as it remains covered by water or a two phase mixture, whether or not natural circulation occurs as a means of removing heat from the primary system.

Natural circulation can exist at B&W plants in three basic modes:

(1) single phase natural circulation in which water would flow through the coolant loops, driven by density differences (2) a continuous flow of a bubbly mixture, driven by density differences (3) boiler-condenser in the steam generators As discussed in SER Section 5.5, natural circulation in modes 1 and 2 could be blocked by steam formation at the top of the hot legs. This volume of steam was stated to be about 320 ft3 per loop or approximately 7% of the primary system volume. For natural circulation to occur in the boiler-condenser mode, a steam-condensing surface must exist. Emergency operating procedures, developed after the accident at Three Mile Island Unit 2, require that the steam generator water level be raised to 95% of the operating range following a small-break LOCA. If the primary system water level were above the 95% water level on the secondary side, the condensing surface would be lost and natural circulation would not occur. The core would be protected, however, because the 95% water level is well above the elevation of the top of the core and con-tinued water loss from the break would establish natural circulation in the boiler-condenser mode before the core could be uncovered. The 95% level Midland SSER 1 6-1

corresponds to a primary system void fraction of approximately 0.20. There-fore, for void fractions between approximately 0.07 and 0.20, natural circu-lation is not expected to occur in B&W-designed reactors.

Although the staff believes that the core will be adequately protected at Midland in the event of a small-break LOCA, it is concerned that (1) un-certainties in the symptoms of a small-break LOCA could result in incorrect operator action and (2) the long-term recovery process by which the reactor system will be cooled and depressurized is not well understood at this time. A better understanding of the recovery process is required for the development of small-break LOCA procedures. These procedures will be developed from emergency guidelines, which will be reviewed by the staff under Itcm I.C.1 of NUREG-0737 as discussed in SER Section 13.5.2.3.

The staff also is conducting a re-review of B&W's small-break LOCA methods under Item II.K.3.30 of NUREG-0737 as discussed in SER Section 15.7.9. The staff is concerned that the CRAFT-2 computer code, which assumes thermodynamic equilibrium, may not be adequate for prediction of primary system refill and bubble collapse that may occur in the recovery process. The staff has determined that integral system experimental data are needed to confirm the predicted behavior of the B&W-designed nuclear steam supply system. The staff currently is examining how best to obtain these data with the B&W Owners Group and NRC's j Office of Regulatory Research. The staff also is evaluating B&W small-break recovery using the TRAC code under contract with the Los Alamos National Laboratory. TRAC is an advanced LOCA code that includes nonequilibrium effects.

The staff currently is evaluating the results from two sets of analyses: (1) a  !

0.0167-ft2 break in a cold leg that was assumed to be subsequently isolated and (2) a 0.0129-ft 2break in a cold leg that was not isolated. The core was predicted to be cooled adequately for both conditions. For the isolated break i case, TRAC calculated that the primary system would be refilled and that the '

hot-leg voids would be condensed sufficiently by incoming high pressure injec-tion (HPI) water so that single phase natural circulation would be reestablished.

For the unisolated-break case, TRAC calculated that condensation of steam by the HPI water and emergency removal by the break flow would be sufficient to depressurize the plant without the establishment of natural circulation.

Continued break .ow prevented the primary system from refilling but no core uncovery was calculated. Although the analyses included various assumed operator actions to restore natural circulation, these actions were not effec-tive (primary system refill was preempted by the continued break flow). The staff has concluded that these calculations are not inconsistent with those previously performed by B&W for the Midland Plant and do not present new information. Furthermore, refill of the primary system beyond the top of the core is not required for small-break LOCA recovery.

In summary, the staff is continuing its evaluation of small-break LOCA models including those used for Midland under Items I.C.1 and II.K.3.30 of NUREG-0737.

The staff is working with the B&W owners and NRC's Office of Research to obtain confirmatory experimental data to address issues discussed in previous paragraphs.

This area will continue as an outstanding issue on the Midland docket until an experimental program to obtain the necessary data is funded and established to further confirm the staff's understanding of portions of the B&W system dynamics, and to provide additional verification of existing analytical methods.

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13 CONDUCT OF OPERATIONS 13.3 Emergency Preparedness 13.3.1 Introduction The evaluation of the state of emergency preparedness associated with the Midland Plant, Units 1 and 2, involves review of the applicant's onsite emer-gency preparedness plus review of the Federal Emergency Management Agency (FEMA) findings and determinations pertaining to offsite emergency prepared-ness. Consumers Power Company submitted Revision 42 of the FSAR in February 1982, which encompasses a revised Site Emergency Plan. This plan was reviewed against the 16 planning standards in 10 CFR 50.47; the requirements of Appendix E to 10 CFR 50; Regulatory Guide 1.101, Revision 2; and the specific ,

criteria of NUREG-0654/ FEMA-REP-1, Revision 1, entitled " Criteria for Prepara- l tion and Evaluation of Radiological Emergency Response Plans and Preparedness l in Support of Nuclear Power Plants," November 1980. The review was done in accordance with the Standard Review Plan (SRP), Section 13.3, " Emergency Plan-ning" (NUREG-0800).

The findings of FEMA on offsite emergency preparedness must be provided before operation above 5% of rated power.

This evaluation report follows the format of Part II of NUREG-0654 in that each of the planning standards is listed followed by a summary of applicable portions of the plan that relate principally to that specific standard. The conclusions of the staff are provided in Section 13.3.3 of this supplement.

13.3.2 Emergency Plan Evaluation Assignment of Responsibility (Organization Control)

The Federal, State, and local organizations that are intended to be part of the overall response organization for the emergency planning zones (EPZs) are iden-tified. The role of the State of Michigan is described, with reference to the State Emergency Preparedness Plan. Midland Caunty Department of Emergency Serv-ices is a primary figure in preemergency planning for the Tri-County area of Midland, Bay, and Saginaw Counties. The responsibilities of Midland, Bay, and Saginaw Counties are outlined and reference is made to the Midland, Bay, and Saginaw County plans. Federal support from the NRC and Department of Energy is described. Working relationships with other organizations are listed, including the Institute for Nuclear Power Operations, Dow Chemical, Bechtel Power Corpor-ation, and Babcock & Wilcox.

The concept of onsite operations and its relationship to the total effort is described and block diagrams showing the interfaces between and among the principal response organizations are provided.

The Site Emergency Director is responsible for all onsite emergency activities and will coordinate assistance outside the plant through the Offsite Emergency Midland SSER 1 13-1

l Coordinator. Before full activation of the emergency operations facility, offsite assistance will be provided from the General Office Emergency Control Center in Jackson, Michigan.

Written agreements are included to verify assistance arrangements between the plant and other support organizations.

The Administrative Supervisor is the individual responsible for ensuring conti-nuity of technical, aaministrative, and material resources.

Onsite Emergency Organization The onsite emergency organization of plant staff personnel is indicated and the normal operations title of each emergency organization role is given.

1 The Site Emergency Director is designated as the individual with authority to direct and coordinate emergency actions. The line of succession for this posi-tion as shown on an organizational chart is the General Manager, the Operations and Maintenance Superintendent, and the Operations Superintendent. The Plant Supervisor will act as the Site Emergency Director until relieved by the General 1 Manager or designated alternate. The duty station for this individual is the Technical Support Center (TSC). .

The interfaces are identified between and among the onsite functional areas of emergency activity, licensee headquarters support, local services support, and State and local government response organizations.

Logistical support for the emergency response personnel activated in an emer-gency is the responsibility of the Internal Services Superintendent of the Bay City Service Center. Technical support is provided by the Technical Super-intendent, and interface with government authority is provided by the Offsite Emergency Coordinator. Public information is coordinated by the Media Coordinator.

Contractors and private organizations who may be requested to provide technical assistance to and augmentation of the emergency organizations are specified.

Police, ambulance, medical, hospital, and fire fighting support which can be provided by local agencies is identified.

The applicant recently submitted a concept of operations for the emergency operations facility. This is currently under review by NRC.

Emergency Response Support and Resources The Executive Vice President of Nuclear Operations, Vice President of Nuclear Operaticas, and Executive Director of Nuclear Operations are specified as the persons authorized to request Federal assistance. Such assistance is to be provided by the Department of Energy (D0E) and would be on site in about 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />. Provisions are made for the dispatch by the Site Emergency Director of a company representative to any of the State or local emergency operations centers.

Midland SSER 1 13-2 1

l l Provisions are made for backup radiological analysis capabilities if the normal laboratory is unavailable. The first backup is the process steam system labora-tory, which has such capabilities. A second backup is the Palisades nuclear plant laboratory. In addition, the Michigan State mobile van is available in about 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and the DOE mobile laboratory is available in about 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />.

Other assistance is available from Detroit Edison Company, Toledo Edison Company, Cincinnati Gas and Electric Company, and Cleveland Electric Illuminating Company through a formal mutual assistance agreement.

l Emergency Classification System 1

The applicant established an emergency classification scheme in accordance with i that set forth in Appendix 1 to NUREG-0654. The four classes are unusual event, 1

alert, site area emergency, and general emergency.

Emergency action levels (EALs) are used to describe each of the four emergency classes. These EALs are composed of a combination of plant parameters (such as instrument readings and system status) that can be used to give a relatively quick indication of the severity of the accident. Specific EALs will be described in the appropriate plant emergency operating procedures and site emergency plan implementing procedures. The initiating conditions that are to form the basis for establishing the specific plant instrumentation readings are given in the plan in an accident classification guide. The guide separates the initiating conditions into four groups: (1) radiological, (2) operational, )

(3) fire / natural / security, and (4) miscellaneous. These initiating and asso-ciated observable conditions Jirect the operators to the emergency procedures l

that delineate actions to be taken for each of the four accident classifications.

The guide includes an initiating condition based on onsite toxic or flammable gases. One of the EALs associated with this initiating condition is a warning of a toxic gas release received from the Dow Emergency Communications and Control Center over the Midland hotline. Additional information is needed to describe the actions to be taken by the plant staff when a warning is received from Dow.

Notification Methods and Procedures Procedures that describe mutually agreed bases for notification of response organizations consistent with the emergency classification scheme, including means for verification, are established. Provisions also are established for alerting, notifying, and mobilizing emergency response personnel. Communica-tions off site may be made by one or more of the following means: conventional )

telephone system / radio system, private telephone lines, and dedicated telephone  ;

lines.

The contents of the initial emergency message to be sent from the plant are preestablished. Initial notification will be made within 15 minutes of emer-gency recognition regardless of the classification.

The applicant states that a public prompt notification system will be installed throughout the 10-mi EPZ before fuel load. The description of this system was submitted recently to the NRC for review. This review is not complete.

Midland SSER 1 13-3

Emergency Communications Communication capabilities with contiguous State / local governments within the EPZs are provided in Table 7.1 of the Site Emergency Plan, which specifies both primary and backup communication resources. These resources consist of direct dedicated telephones, intraplant telephones, company radios, Bell telephones, and State Police radios. Communication links are provided to various onsite and offsite locations including the control room, technical support center, operational support center, emergency operations facility, General Office Con-trol Center in Jackson, Power Controller (Central Region), NRC, State emergency operations center (EOC), Midland County E0C, radiation monitoring teams, and Dow Chemical Company.

The control room, county E0C, State E0C, NRC, and Consumers Power offsite emer-gency support organizations maintain 24-hour coverage of their communication systems. Communication links between the site, State, and local emergency operations centers and field assessment teams will be tested annually.

Public Education and Information The Media Coordinator is responsible for disseminating information to the public by various methods such as direct mailing of literature, providing information brochures in billing statements, telephone book inserts, and posting information documents in public areas. This information pertains to topics associated with plant activities and will be updated where required and disseminated on an annual basis to members of the public within the 10-mi emergency planning zone.

The joint public information center (JPIC) is located in a lecture theater and adjacent classrooms at Delta Community College approximately 13.5 mi east of the plant in Bay County, Michigan. The JPIC is staffed by technical and public affairs personnel from Consumers Power Company and public affairs personnel from Federal, State, and local agencies. The Governor's press secretary is  !

designated as the principal spokesperson during a crisis. {

l Coordinated arrangements for dealing with rumors are established. )

Emergency Facilities and Equipment 1

The applicant has provided a technical support center (TSC) in the protected area within the plant security perimeter. It is separate from but near the (

control room (within about a 2-minute walk), and contains more than 4,000 f t2 of floor area. It is designed to have similar radiological habitability as the i control room and will be fully functional within 30 minutes after activation.

The operations support center (OSC) at the Midland Plant is located in the machine shop / lunchroom area of the administration and services building. The Maintenance Superintendent is the OSC Supervisor and is designated to make assignments as necessary. The primary OSC communication link with the control room and TSC is the plant telephone system. Direct voice radio communication by walkie-talkie supplements the telephone links.

The emergency operations facility (EOF) is located in the Consumers Power Bay City Service Center approximately 18 mi east of the plant in Bay City, Michigan.

At the onset of an emergency situation, before full activation of the EOF, off-site assistance is provided from the general office emergency control center Midland SSER 1 13-4 l

l in Jackson, Michigan. Before full activation of the EOF, personnel from Midland Plant designated by the Site Emergency Director report to the E0F to initiate those actions necessary to augment the onsite emergency organization.

The applicant recently has submitted to the staff for review a more detailed description of the E0F concept of operations. This review is in progress.

Most of the onsite monitoring systems and instrumentation used to initiate emergency measures and/or provide continuing assessment are identified. These systems include a meteorology system with wind speed and direction and tempera-l ture capabilities, seismic instrumentation to measure ground acceleration levels, inprocess lines that actually or potentially contain radioactive effluents, installed area radiation monitors to measure increases in radiation levels in specific locations in the station, fire and smoke detection instruments placed i in strategic plant locations, portable dose rate and radiation detection instru-ments, and laboratory counting and analysis facilities.

{ The applicant recently submitted review copies of the site-specific atmospheric diffusion and dose projection models. Staff review is in progress.

Procedures are developed for emergency preparedness including those for the inventory and operational readiness of emergercy equipment and supplies on a quarterly basis. Calibration of equipment is conducted at intervals recommended by equipment suppliers.

j i

Offsite radiation monitoring results are reported to the TSC until the EOF is operational. Once the E0F is functional, results will be reported there by l I

portable radios. Onsite radiation monitoring results will be reported to the t

TSC by the health physics network or portable radios.

Accident Assessment Provisions are made to estimate integrated doses from projected and actual dose rates using a computer-based dose projection model with input from a meteoro-logical model. These doses are then compared with referenced Environmental Protection Agency and State of Michigan Protective Action Guides (PAGs). To determine the source term, primary reliance is placed on direct measurements using inplant instrumentation for which procedures have been developed that relate radiation levels to radioactive materials. If the instrumentation is inoperable or off scale, backup procedures are developed for estimating radia-tion doses. These procedures assume site-specific source terms for different accident scenarios and meteorological conditions. Backup for the computer l model is provided by calculator programs and transparent overlays depicting l dose isopleths.

In addition to the use of onsite instrumentation, field monitoring teams are dispatched to make direct field measurements and to collect air, soil, and water samples. They also will " change out" the environmental thermoluminescent dosimeters (TLDs). These teams have the capability and resources to measure radioiodine levels in the plume EPZ as low as 10 7 pCi/cc using special silver zeolite cartridges that can be placed in portable air samplers. The onsite teams are deployed and report initial information to the plant by portable radio within 20 to 30 minutes. Two offsite teams are available for deployment in about 60 minutes.

Midland SSER 1 13-5

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Protective Response The plan describes the protective actions to be taken by onsite personnel including the assembling of all personnel not having an emergency response assignment in predetermined areas. This includes visitors and contractor per-sonnel. The notice to assemble is given by a plant alarm system. Alphabetical lists of personnel in each assembly area are posted to help in verifying their presence. The lists contain the names of all persons assigned to that partic-ular area. Personnel are checked off as they arrive at their designated loca-tion. The visitor sign-in sheet is brought to the visitor assembly area to' account for those individuals temporarily on site. Personnel are monitored, decontaminated, and evacuated as necessary.

The plan provides for respiratory protection, use of protective clothing, and use of radioprotective drugs for onsite workers. It also states that the appli-cant will promptly notify offsite authorities in the event of an emergency and '

provides a mechanism for recommending protective actions to them based on the Environmental Protection Agency PAGs.

Population distribution by sector and distance within a 50-mi radius is compiled -

and is included in the plan. Maps indicating major evacuation routes for the-public and site personnel are also provided. In the evacuation time estimates the relationship of the Midland Plant to Dow Chemical is considered. The Dow 1 Chemical facility has had an emergency plan for more than 30 years. Its evac-uation plan is activated immediately on notification by the Midland Plant (over the control room hotline) that evacuation is necessary. The Dow plan provides for evacuating the Dow Complex in 1 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> under the worst conditions and within 45 minutes under the best conditions.

Arrangements are also made to take protective actions for Midland personnel l based on an accident at Dow. As stated previously in this supplement, the applicant provides an EAL for a protective response based on a toxic gas release at the Dow Chemical plant. Communications are established between the Dow emergency center and the Midland control room. The staff also is aware of a general integration of response plans between Dow and Midland but notes that a description of such an integration is not provided in the Midland Emergency Plan. The staff also supports the comment by the Advisory' Committee on Reactor Safeguards (Appendix G of this supplement) relative to this' subject that encour-ages participation by Midland Plant personnel in emergency procedures' developed on the basis of an assumed failure at the Dow Chemical plant.

Radiological Exposure Control Onsite exposure guidelines consistent with the Environmental Protection Agency PAGs are established for removing injured personnel, and for corrective and ' I assessment action, first aid, personnel decontamination, medical t"ransport, and J medical treatment services. Provisions exist for 24-hour per-day capability to '

determine the exposures of emergency personnel involved in a nuclear accident.'

Emergency response personnel may receive exposures greater than 10 CFR 20 limits but consistent with those in the Environmental Protection Agency PAGs for emer-gency workers. Planned axposures are on a voluntary basis and must he author-ized by the Site Emergency Director on the basis of a recommendation by the x Radiological Assessment Coordinator. If these personnel are unavailable, the Midland SSER 1 13-6 .

. . - - . - m_

plant Health Physicist or Senior Health Physicist makes the authorization I

decision.

The station supples clothing and decontamination materials to onsite perscnnel requiring relocation and found to be contaminated.

In the event of a nuclear accident, procedures are in effect to provide for distribution of self reading dosimeters and TLDs to emergency workers. Per-sonnel are trained in proper procedures to follow when they are working in a radiation field and are instructed to return to the health physics area when self reading dosimeters exceed 3/4 full scale.

Medical and Public Health Support The station provides for onsite first aid capability. Radiation protection  ;

personnel and at least one person on each operating shift are required to have first aid training.

The applicant has made arrangements by written agreement with the Midland Hospital Center and Bay City Hospital Center to previde medical assistance to personnel injured or exposed to radiation and/or radioactive material.

Transportation of victims is provided by the Midland Hospital Center Emergency Medical Services Department. This department is equipped and manned for trans-portation of both contaminated and noncontaminated patients.

( Recovery and Reentry Planning and Postaccident Operations i

The plan describes the applicant's general plans for recovery and reentry. Any decision on the applicant's part to relax protective measures is made by the General Manager / Site Emergency Director with the mutual agreement of the NRC and the Michigan Department of Public Health, Radiological Health Services Division. Whenever a recovery operation is initiated or any change is made in the organizational structure, the offsite Emergency Coordinator notifies repre-sentatives of the response organizations.

Computer models and other data are used to calculate estimates of total popula-tion dose on a periodic basis for comparison with PAGs. Terminals for using these models are located in the TSC and EOF.

Exercises and Drills Annual exercises are conducted to test the integrated capabilities and a major portion of the basic elements existing within the plan. Offsite organizations as well as applicant response organizations are involved. At least once every 6 years, an exercise is started between 6:00 p.m. and midnight and another between midnight and 6:00 a.m. The scenarios used for the various exercises contain the essential elements set forth in NUREG-0654 and are designed to allow flexibility in decisionmaking. Provisions exist for unannounced exercises.

In addition to the exercises, various drills are conducted covering communica-tions, fires, medical emergencies, health physics, and radiological monitoring.

Drills are supervised instruction periods aimed at testing, developing, and Midland SSER 1 13-7

maintaining skills in these areas. Management control is established so that necessary corrective actions are implemented.

The onsite Emergency Coordinator and plant management are responsible for the planning, scheduling, and coordinating of all emergency planning related to drills and exercises. All drills and exercises are approved by the General Manager. In addition, the Vice President of Nuclear Operations approves the annual radiation emergency drill.

Radiological Emergency Response Training The Nuclear Training Coordinator is responsible for coordinating the training of all station personnel, and the plan provides for training and qualifying all personnel in the emergency tasks for which they are responsible. Satisfactory completion of the training program elements for such personnel is verified by direct observation of performance in drills or exercises and/or a written examination.

Those personnel assigned to the plant emergency organization with specific Emergency Plan duties and responsibilities receive specialized training for  ;

their respective assignments. All personnel badged for unescorted access to l the protected area receive a general indoctrination in the Site Emergency Plan l and the Site Emergency Plan Implementing Procedures. Selected station personnel l on each shift attend the multimedia National Red Cross First Aid Course.

Responsibility for the Planning Effort: Development, Periodic Review, and Distribution of Emergency Plans The Emergency Planning Coordinator has the overall responsiblity for the Midland Site Emergency Plan and Implementing Procedures and for reviewing the offsite county and supporting emergency plans for compatability with the Site Emergency Plan.

The plan, as well as any changes thereto, is provided to the organizations and individuals having a responsibility to implement it. Provisions exist for an annual review of the plan and for the incorporation of necessary revisions. An independent review of the overall emergency preparedness program is conducted at least annually by the Institute for Nuclear Power Operations (INPO), but in the absence of the INP0 review it is conducted by the applicant's Quality Assurance Nuclear Operation Department.

13.3.3 Conclusion Based on its review, the staff concludes that the Midland Site Emergency Plan, )

on satisfactory compler, ion of the items listed below, will meet the planning ]

standards of 10 CFR 50.47(b) and the requirements of 10 CFR 50, Appendix E, and will conform with the guidance in Regulatory Guide 1.101, Revision 2. The items are summarized as follows:

(1) The applicant must provide a brief description of the interfaces between the emergency plan for the Midland Plant and for the Dow plant emphasizing the actions that Midland Plant personnel will take after being notified of an accident at Dow.

Midland SSER 1 13-8

(2) The staff now has under review the applicant's (a) meteorological and dose assessment proposals (b) concept of operations and method for meeting the staffing guidelines of NUREG-0654 for the EOFs (c) description of the prompt notification system The applicant must resolve satisfactorily any deficiencies that result from this review.

l Midland SSER 1 13-9

1 15 ACCIDENT ANALYSIS 15.2 Secondary System Transients and Accidents 15.2.3 Decrease in Heat Removal The SER noted, as a confirmatory issue, that the staff had requested the basis for the applicant's steam generator inventory predictions used in calculations for transients involving decrease in heat removal by the secondary system. ,

This issue also was identified as Item 29 in SER Section 1.8. The applicant I subsequently has confirmed that these analyses were based on test data from the Alliance 19-tube steam generator tests. The predicted inventories were decreased by 10% for conservatism. The staff finds this to be a sufficient basis, and considers this confirmatory issue closed.

15.3 Reactor Coolant Transients and Accidents 15.3.1 Decrease in Reactor Coolant Flow Rate In SER Section 15.3.1, the staff requested, as a confirmatory issue, that the applicant perform analyses of partial loss-of-flow transients assuming tripping of one reactor coolant pump during initial four pump and three pump operation.

l This issue also was identified as Item 30 in SER Section 1.8.

The analyses subsequently have been performed at the naxi.num power levels for each mode of operation (100% power for initial four pump operation, and 77%

power for the initial three pump case). Of the two cases, the initial four-pump case resulted in the minimum calculated departure from nucleate boiling ratio, and had a value of 1.43. The staff finds that no fuel damage results from this event. The plant design, therefore, meets the acceptance criteria of SRP Section 15.3.1 for loss-of-flow events, and is acceptable. This confirma-tory issue is closed.

i 1

Midland SSER 1 15-1

19 REVIEW BY ADVISORY COMMITTEE ON REACTOR SAFEGUARDS l During its 266th meeting, June 3-5, 1982, the Advisory Committee on Reactor Safeguards (ACRS) reviewed the application of Consumers Power Company for licenses to operate the Midland Plant, Units 1 and 2. This application was also considered at Subcommittee meetings hele on April 29, 1982, in Washington, D.C.; on May 20-21, 1982, in Midland, Michigan; and on June 2, 1982, in Washington, D.C. On May 20, 1982, members of the Subcommittee toured the plant. Transcripts of each of these meetings are available from Alderson I

Reporting, 400 Virginia Avenue, S.W., Washington, D.C. 20024. Copies of transcripts are also available for review at the NRC Public Document Room at 1717 H Street, N.W., Washington, D.C. 20555, and at the Grace Dow Memorial Library at 1710 W St. , Andrews Road, Midland, Michigan 48640.

A copy of the ACRS Interim Report on Midland Plant, Units 1 and 2, dated June 8, 1982, is included as Appendix G to this supplement. The interim report indicates the ACRS's belief that subject to satisfactory completion of construc-tion and staffing and if due regard is given to comments therein, the Midland Plant, Units 1 and 2, can be operated at power levels up to 5% of full power with reasonable assurance that there is no undue risk to the health and safety of the public. The ACRS defers its recommendation regarding operation at full power pending review of a plan for an audit of plant quality and resolution of the question of natural circulation in the presence of a small-break LOCA.

These and other comments from the June 8, 1982 letter are discussed below. A further supplement to the Midland SER will be issued on receipt of ACRS recom-mendations regarding operation at full power.

(1) Design Adequacy and Construction Quality The Committee noted its overall concern about Midland quality assurance and recommended that the NRC arrange for a broader assessm nt of Midland's design adequacy and construction quality with emphasis on installed electrical, control, and mechanical equipment as well as piping and foundations. The Committee also noted its desire to receive a report discussing design and construction problems, their disposition, and the overall effectiveness of the effort to ensure appropriate quality.

The NRC, through its Region III office, is preparing a report discussing Midland construction problems, their disposition, and the overall effectiveness of the Consumers Power Company's effort to ensure appropriate quality. This report will cover the period starting with the beginning of construction up to June 30, 1982, and will be issued by October 1, 1982. Before the NRC's finding relative to Midland readiness for operation, a final report will be issued on the above subjects for the period from July 1, 1982, through the completion of construction discussing the overall quality of plant construction. At that time the preoperational and hot functional test program should be essentially complete, providing an additional insight into the overall construction and design quality. Any new staff inspection program activities, and their results, Midland SSER 1 19-1

\

i directed toward evaluating design quality will be included in this final report.

In addition to the above NRC reports, the staff is considering the need for the applicant to acquire an independent design verification to evaluate the adequacy of Midland design and construction for NRC review. This matter will be resolved in the near future.

(2) Decay Heat Removal Following More Severe Earthquakes The Committee recommended that all systems and components important to decay heat removal be carefully evaluated for their ability to accomplish necessary functions in the unlikely event of Icwer probability, more severe earthquakes.

The Committee noted its wish to be kept informed about resolution and suggested that any plant changes from this evaluation be implemented by the end of the second refueling outage.

Under current requirements, applicants qualify safety-related equipment to five operating-basis earthquake events followed by a safe shutdown earthquake (SSE) in combination with relevant dynamic and static loads created by postulated accidents. Within a given plant a certain equipment type is qualified to the most severe conditions that exist at the various equipment locations. Further-more, to qualify equipment for use in many different plants, the qualification is usually performed to even higher excitations than the limiting condition in a given plant. During qualification, equipment is tested to predetermined conditions. The failure level of the equipment, which could be significantly j higher than the test condition, is not established. Thus, a certain amount of  ;

margin exists between the failure condition of the equipment and the expected loads during an SSE. The margin, however, is not quantified.

NRC has no requirement to design equipment to more severe earthquakes than the SSE. Nevertheless, because of the margin inherent in the design of individual components and the fact that redundant components of a safety system typically are not exposed to the same loads because of different locations, it is likely that systems functions would not be lost in case of earthquakes somewhat more severe than the SSE.

(3) Reactor Vessel Head Vent The Committee noted that the applicant has not committed to supply a high point vent on the reactor vessel head and concluded that this matter should be re-solved in a manner satisfactory to the NRC staff.

The NRC staff's review of the Midland high point vent design is discussed in SER Section 5.4.7. It is noted that the applicant has not complied with 10 CFR 50.44.C.3.iii, which requires that a high point vent be installed on the reactor vessel head. The applicant has not committed to meet this requirement at this time. This issue remains oper (4) Reactor Vessel Level Indication The Committee recommended that the applicant review further the potential for providing indications of water content or level within the reactor vessel.

Midland SSER 1 19-2 Y

The NRC staff's review of the inadequate core cooling instrumentation proposed for the Midland Plant is discussed in SER Section 4.4.4.1. The SER notes that the information submitted by the applicant in response to Item II.F.2. of NUREG-0737 is incomplete and that submittals by the applicant should include additional differential pressure instrumentation setween the vessel head and l

the lower level of the hot leg. The SER further indicates that this issue will be the subject of a license condition.

In addition to the above information in the SER, it is the staff's preliminary position that further submittals by the applicant also should include a reactor coolant pump current monitor, or equivalent, to trend void content of reactor coolant during forced circulation. An acceptable equivalent would be a dif-ferential pressure transmitter sensing pressure change from a tap at the bottom of the vessel, and designed to trend voiding with the pumps running.

This matter will be discussed in subsequent supplements pending submittal and review of Item II.F.2 information.

(5) Operating Shift Experience The Committee endorsed the NRC staff position concerning experience on each operating shift.

l The NRC staff position, as given in SER Section 13.1.2.1 for the Operations Department, is that each operating shift should have at least one person who has commercial pressurized-water reactor (PWR) operating experience in the startup of a similar PWR. The shifts should be so supplemented at fuel load and for 1 year or until the attainment of nominal 100% power, whichever occurs later. A copy of the Committee's letter has been forwarded to the applicant.

By letter dated February 11, 1982, the applicant has stated that he intends to meet the staff position and is exploring the availability of experienced personnel through several agencies. The NRC staff will condition the appli-cant's license to require appropriate experience on each shift.

(6) Augmented NRC Audit of Operations The Committee recommended that the NRC staff institute an augmented audit of operations at Midland, at least during the early years of operation at power.

NRC Region III will implement an augmented audit of operations at Midland dur-ing the early years of operation. The scope of this program will be developed before an operating license is issued and will be based on the following:

(a) Resident Inspector Experience - A construction Senior Resident Inspector has been at the Midland site since July 1978. An opera-tions Resident Inspector will be assigned to Midland starting August 3, 1982.

(b) Operating experience with other Consumers Power Company nuclear power plants.

(c) Regional inspection experience with the applicant's performance in conducting the preoperational and hot functional test programs - This Midland SSER 1 19-3

i will be an augmented inspection effort going beyond the current inspection program.

(d) Results of the regional based inspection team audit to evaluate the operational quality assurance program - This audit begins approxi-mately 6 months before an operating license is issued.

It should be recognized that the routine inspection program provides extensive coverage of reactor operations during the first year of operation. Experience from the routine program will be used to " fine tune" the augmented audit of operations.

(7) Probabilistic Risk Assessment The Committee noted that the applicant is having a probabilistic risk assessment (PRA) performed for the Midland Plant. The Committee indicated its desire to review the Midland PRA with assistance from the NRC staff and to offer comments or recommendations as appropriate.

The applicant has advised the NRC staff that its PRA will be submitted for review in early 1983. The NRC staff will interact with the Committee during the course of its review.

(8) Natural Circulation During Small-Break LOCA The Committee noted recent questions regarding availability of natural cir-culation for B&W plants in the presence of an interrupted or continuing small-break LOCA and indicated its wish to see a proposed NRC staff resolution of this issue.

This matter is discussed in Section 6.3.4.1 of this supplement.

(9) Systems Interactions The Committee noted that the applicant had described a systems interactions i study being undertaken for the Midland Plant. The Committee indicated its desire to be informed of the results of this study.

During a meeting with the ACRS Subcommittee on June 2, 1982, the applicant discussed his initial steps toward the performance of a separate analysis from a multidisciplinary point of view. The applicant plans a program to reduce common-cause failures by addressing spatially coupled, functionally coupled, and humanly coupled systems interactions. He plans to use walkdowns for l seismic, flooding, and high-energy-line-break hazards. He stated that humanly coupled interactions are reduced by operator training, control room design, and the review of operating experience that is reported by the Institute for Nuclear Power Operations. The applicant plans to analyze functional systems interactions by a risk assessment.

The staff's systems interaction program is discussed in Appendix C of the SER, Unresolved Safety Issue (US1) A-17. The staff considers that the program outlined by the applicant provides experience that will be a significant addition to the considerations of systems interactions at the Midland Plant.

l Midland SSER 1 19-4

i By letter of June 25, 1982, the staff has requested that the applicant provide a summary report of his plans and the results of the program for staff review.

The staff will report the results of its review of the Midland study.

(10) Population Density Considerations The Committee stated its belief that in view of the population density near this plant, additional prudence is appropriate for the Midland Plant in the resolution of the issue of anticipated transients without scram and other unresolved safety issues.

The technical resolution and implementation of an unresolved safety issue are done on a generic basis, rather than on an individual plant basis, and usually take one of the following two forms:

(a) A prescribed resolution involving design and hardware changes and operational procedure changes to address specifically defined new criteria or requirements that are applicable to all plants.

(b) A prescribed resolution as above, but for several different groups of plants where the prescription is somewhat different for each group of plants. Groupings can be defined in terms of chronology of design (based typically on application or construction permit date); com-monality of design (as for example, PWRs and BWRs); or commonality of other particular design features (for example, dry containment, ice condenser containment, or Mark II BWR containment).

For resolutions patterned after Item (a), the resolution is reasonably well defined for each plant and there is little latitude in applying such a resolu-tion to individual plants. For resolutions patterned after Item (b), there are a number of different resolutions that apply to different plant groups, and there is, therefore, some flexibility in deciding on the implementation of a resolution for a particular plant.

The staff will specifically consider the ACRS recommendation on Midland when grouping the plants for implementation of a technical resolution of the type described in Item (b). For this type of technical resolution, this considera-tion may be explicitly included in the implementation decisionmaking process.

The staff will also consider the ACRS comment, to the extent practical, in implementing technical resolution of the type described in Item (a). In this latter case, however, it should be recognized that there is normally little flexibility in implementing the prescribed generic resolution on an individual plant basis.

(11) Emergency Procedures The Committee believes there should be active participation by Midland Plant personnel in emergency procedures developed on the basis of an assumed accident at the Dow Chemical Plant. The NRC staff endorses this recommendation as discussed in Section 13 of this supplement. The staff also has requested the applicant to provide a brief description of the interfaces between the emergency plan for the Midland Plant and that for the Dow Plant emphasizing the actions that Midland Plant personnel would take in the event of an accident at Dow.

Midland SSER 1 19-5

n' (12) Turbine Missiles The Committee stated that it wished to be kept advised concerning resolution of the turbine missile issue.

This matter is discussed in Section 3.5.1.3 of this supplement.

J Midland SSER 1 19-6

APPENDIX A CONTINUATION OF CHRONOLOGY April 30, 1982 Letter from applicant concerning unresolved safety issues.

Makes correction to letter dated February 22, 1982.

April 30, 1982 Letter from applicant concerning effects of cracks on serviceability of concrete structures and repair of cracks.

May 3, 1982 Letter from applicant concerning test results for remain-ing soil test specimens taken for various plant structures.

May 3, 1982 Letter from applicant forwarding photographic documents of soil samples.

May 3, 1982 Letter from applicant forwarding underground piping information information requested during April 16, 1982 meeting.

May 4, 1982 Letter from applicant concerning challeng'e to safety valves caused by steam generator overfill protection system.

May 4, 1982 Board Notification 82 Systematic Assessment of Licensee Performance (SALP) Report on Midland Plant for period July 1, 1980 to June 30, 1981.

May 5, 1982 Atomic Safety and Licensing Appeal Board issues Memoran-dum and Order affirming Atomic Safety Licensing Board's (ASLB) ruling denying intervenor Wendell Marshall's request to halt further construction of the Midland facility because of electromagnetic pulse issue.

May 7, 1982 ASLB issues Memorandum and Order (telephone conference call of May 5, 1981).

May 7, 1982 Telephone conference call on auxiliary building phase 2 underpinning issues (summary issued May 17, 1982).

May 7, 1982 Letter from applicant concerning limit analyses to evaluate service water pump structure east and west wall capacities.

May 7, 1982 Letter from applicant concerning soil impedance functions of the auxiliary building's electrical penetration wings.

Midland SSER 1 A-1

I May 10, 1982 Letter from applicant requesting clarification of staff approvals regarding specific soils construction activities.

May 10, 1982 Letter from applicant concerning ASLB April 30, 1982 Order imposing certain interim conditions on the remedial soils and related work.

May 10, 1982 Letter from applicant on borated water storage tank load-ing evaluation and releveling procedure.

May 11, 1982 SER issued.

May 14, 1982 Letter from applicant concerning underpinning of the auxiliary building.

May 17, 1982 Letter from applicant responding to draft SALP Report.

May 19, 1982 Letter from applicant transmitting the Weston Geophysical Report on the probability of exceeding the operating-basis earthquake.

May 20-21, 1982 ACRS Subcommittee meeting and site tour in Midland, Michigan.

May 24, 1982 Meeting with applicant to discuss the status of emergency planning.

May 25, 1982 Letter to applicant concerning the completion of soils remedial activities review.

May 26, 1982 Letter to applicant forwarding Amendment 3 to construc-tion permits in response to ASLB April 30, 1982 Order imposing certain interim conditions pending issuance of a partial initial decision.

June 1, 1982 Meeting with applicant to discuss plant security plan (summary issued June 23, 1982)

June 1, 1982 Letter from applicant concerning settlement-related analyses for the diesel generator building.

June 1, 1982 Letter from applicant concerning schedule for response to request for additional information on soils remedial activities.

June 1, 1982 Letter to applicant concerning review comments for the security and contingency plans.

June 1, 1982 Letter from applicant concerning issues related to the Site Emergency Plan.

June 2, 1982 ACRS Subcommittee meeting.

Midland SSER 1 A-2

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June 3, 1982 ASLB issues order denying Wendell Marshall petition to halt construction of the facility.

June 4, 1982 ACRS meeting.

June 7, 1982 Letter from applicant with technical report on dewatering recharge tests and responding to requests on auxiliary building underpinning.

June 7, 1982 Letter from applicant concerning postaccident access to the decay heat removal manual valves.

June 8, 1982 Telephone conservation with applicant on quality plan for underpinning (summary issued June 18, 1982).

June 8, 1982 Letter to applicant requesting information regarding masonry walls and ultimate containment capacity.

June 8, 1982 Letter from ACRS providing Interim Report on Midland Plant.

June 9 1982 e Generic Letter 82 Transmittal of NUREG-0916 relative to the restart of R. E. Ginna Nuclear Power Plant.

June 11, 1982 Letter to applicant transmitting ACRS Interim Report.

June 11, 1982 Meeting with applicant to discuss additional information required to complete staff review of soils remedial work (summary issued June 18, 1982).

June 11, 1982 Meeting with applicant to discuss tornado-missile protec-tion.

June 14, 1982 Letter from applicant responding to request for additional information on soils remedial actions from staff's letter of May 25, 1982.

June 14, 1982 Letter from applicant on relationship of crack width and spacing to stress.

June 15, 1982 Meeting with applicant to discuss elements of staff review schedule being developed for soils remedial activities (summary issued June 23, 1982)

June 15, 1982 Letter from applicant describing preliminary design of Midland prompt public notification system.

June 16, 1982 Meeting with applicant to discuss additional information needed for the equipment environmental qualification audit.

June 16, 1982 Letter from applicant submitting reliability analysis for three pump auxiliary feedwater system.

Midland SSER 1 A-3

3 June 17, 1982 Generic letter 82 changes in examination used to license reactor and senior reactor operators.

June 18, 1982 Letter from applicant on feedwater isolation valve pit load verification.

June 21, 1982 Meeting with applicant to discuss response to draft SALP Report.

June 21, 1982 Letter from applicant announcing design change from internal to external auxiliary feedwater distribution header.

June 23, 1982 Letter to applicant regarding problems with certain Midland correspondence.

June 25, 1982 Letter to applicant requesting additional information on locked-rotor analysis.

June 25, 1982 Letter to applicant requesting turbine-missile generation probability data.

June 25, 1982 Letter to applicant on systems interactions.

June 25, 1982 Meeting with applicant on staff's requests for additional information on soils remedial actions.

June 28, 1982 Meeting with applicant to discuss fire protection open items and review drawings of safety-related electrical wiring.

June 30, 1982 Letter to applicant concerning feedwater isolation valve pit proof test.

Midland SSER 1 A-4

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i APPENDIX B BIBLIOGRAPHY Babcock & Wilcox report 12-1132424, " Bounding Analysis Impact Study of l NUREG-0630." l

-- , " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177-Fuel Assembly Plant," May 7, 1979.

Code of Federal Regulations, Title 10, " Energy" (includes General Design Criteria).

Electric Power Research Institute, NP-768, " Tornado Missile Risk Analysis:

Probability Modeling, Simulation Methodology, and Case Studies,"

May 1978.

-- , NP-769, " Tornado Missile Risk Analysis Appendixes - Analytical Models and Data Bases," May 1978.

U.S. Nuclear Regulatory Commission, NUREG-0630, " Cladding Swelling and Rupture Models for LOCA Analysis," April 1980.

-- , NUREG-0654/ FEMA-REP-1, " Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants," Rev. 1, November 1980.

-- , NUREG-0737, " Clarification of TMI Action Plan Requirements," November 1980.

-- , NUREG-0793, " Safety Evaluation Report Related to the Operation of Midland Plant, Units 1 and 2, May 1982.

-- , NUREG-0800 (formerly NUREG-75/087), " Standard Review Plan for Review of Safety Analysis Reports for Nuclear Power Plants," July 1981 (includes Branch Technical Positions).

-- , Regulatory Guide 1.101, " Emergency Planning for Nuclear Power Plants,"

Rev. 2.

Midland SSER 1 B-1

l APPENDIX D ABBREVIATIONS ACRS Advisory Committee on Reactor Safeguards ASLB Atomic Safety Licensing Board B&W Babock & Wilcox CFR Code of Federal Regulations DNBR departure from nucleate boiling ratio DOE U.S. Department of Energy EAL emergency action level ECCS emergency core cooling system EOC emergency operations center EOF emergency operations facility EPZ emergency planning zone FEMA Federal Emergency Management Agency HPI high pressure injection INPO Institute of Nuclear Power Operations JPIC joint public information center LOCA loss-of-coolant accident NRC U.S. Nuclear Regulatory Commission OSC operational support center PAG protective action guide PRA probabilistic risk assessment PWR pressurized-water reactor PORV power-operated relief valve SALP systematic assessment of licensee performance SER safety evaluation report SRP Standard Review Plan SSE safe shutdown earthquake TLD thermoluminescent dosimeter TMI-2 Three Mile Island Unit 2 TSC technical support center USI unresolved safety issue Midland SSER 1 D-1

i APPENDIX E NRC STAFF CONTRIBUTORS This supplement is a product of the NRC staff. The staff members listed below were principal contributors to this report.

Name Title Review Branch Franklin D. Cof fman, Jr. Section Leader, Systems Reliability and Risk Interaction Assessment Robert W. DeFayette Sr. Emergency Incident Response and Planning Specialist Development Walton L. Jensen Sr. Nuclear Engineer Reactor Systems Karl Kniel Chief Generic Issues Arnold Jen-Hsu Lee Sr. Mechanical Engineer Equipment Qualification William T. LeFave Sr. Auxiliary Systems Auxiliary Systems Engineer William S. Little Chief, Engineering Region III Inspection Branch Johnny L. Mathis Reactor Safety Engineer Emergency Preparedness Licensing James M. Peschel Inspector, Management Licensee Qualifications, Programs Region III Laurence E. Phillips Section Leader, Core Performance Thermal Hydraulics Dale A. Powers Metallurgist, Core Performance Fuel Systems John 0. Schiffgens Materials Engineer Materials Engineering Midland SSER 1 E-1

APPENDIX G ACRS INTERIM REPORT ON MIDLAND PLANT, UNITS 1 AND 2 Midland SSER 1

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, , - "g gl NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS l 0, WASHINGTON, D. C. 20555

\ * * " * $g June 8, 1982 Honorable Nunzio J. Palladino Chairman U. S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

ACRS INTERIM REPORT ON MIDLAND PLANT, UNITS 1 AND 2 During its 266th meeting, June 3-5, 1982, the Advisory Committee on Reactor Safeguards reviewed the application of Consumers Power Company for a li-cense to operate the Midland Plant, Units 1 and 2. This application was also considered at Subcommittee meetings held on April 29, 1982 in Washing-ton , D. C. , on May 20-21,1982 in Midland, Michigan and on June 2,1982 in Washington, D. C. On May 20,1982 members of the Subcommittee toured the plant. In the course of these meetings the Committee had the benefit of n discussions with representatives and consultants of Consumers Power Company,

! Babcock and Wilcox Company, Bechtel Corporation, the Nuclear Regulatory Commission Staff, and members of the public. The Committee also had the benefit of the documents listed below.

The ACRS reported on June 18,1970 regarding the construction permit ap-plication for the Midland Plant; on September 23, 1970 regarding several amendments to the application; and on November 18, 1976 regarding applica-ble generic matters.

. The Midland Plant site is located on the south bank of the Tittabawassee River adjacent to the southern city limits of Midland. The main industrial

'j complex of the Dow Chemical Company lies within the city limits directly across the river from the site. There are about 2000 industrial workers i within one mile of the site, and the estimated 1980 population was about 51,400 residents within five miles of the site. This makes the Midland site one of the more densely populated sites at distances close to the

, Plant.

Each of the two Midland units employs a Babcock and Wilcox desigaed nuclear steam supply system cated at 2468 MWt with a stretch power rating of 2552 MWt. The Midland Plant is unique in that the heat generated will be used not only to produce electricity but also to produce process steam for the Dow Chemical Company plant via a tertiary system.

Tne Midland Plant has been the subject of several major problems related to quality assurance during plant construction. One of these problems relates to the soil fill under several safety-related structures. The Midland SSER 1 G-1

Honorable Nunzio J. Palladino June 8, 1982  !

deficiencies relating to soil fill have led to excessive settlement and some cracking of these structures, and have also introduced questions concerning the adequacy of protection against liquefaction of the granular f portions of the fill in the event of strong vibratory motion accompanying an earthquake.

)

The Applicant has proposed and is implementing, under close surveillance by the NRC Staff, remedial measures with regard to the foundation deficiencies.

We are generally satisfied with the approach being taken, subject to confir-mation of the overall quality assurance program and the seismic design basis. Both of these items are discussed below.

With regard to quality control of design and construction, the report of the NRC Staff's Systematic Assessment of Licensee Performance (SALP) review for the period July 1,1980 to June 30, 1981 revealed deficiencies in the instal-lation of piping and piping suspension systems, in the pulling of electrical cables, and in the handling of problems relating to soils and foundation.

Deficiencies by the Applicant in the handling of soils-related matters have continued to occur, subsequent to issuance of the SALP report. We believe that the NRC Staff is handling the corrective actions for specifically identified quality assurance deficiencies in an appropriate manner. .

In view of the overall concern about Midland quality assurance the NRC should arrange for a broader assessment of Midland's design adequacy and construction quality with emphasis on installed electrical, control, and mechanical equipment as well as piping and foundations. We wish to receive a report which discusses design and construction problems, their disposi-tion, and the overall effectiveness of the effort to assure appropriate quality.

Our reservation concerning seismic design relates to the lack of adequate assurance that the Midland Plant will be capable of accomplishing shutdown heat removal for low probability earthquakes more severe than the safe shutdown earthquake (SSE). The Midland seismic design basis at the con-struction permit stage corresponded to a MMI VI, peak ground acceleration of 0.129, employing a modified Housner spectrum. For the operating license review, the NRC Staff has reevaluated the original seismic design basis and the Applicant and the NRC Staff have agreed on the use of site-specific analyses which have led to increases in the design response spectra for frequencies above about 2 cycles /sec.

Historically, no earthquakes stronger than the newly proposed SSE have occurred within 200 miles of the Plant. However, expert opinion differs widely on the exceedance frequency of the proposed SSE and on thg severity at 5the site of earthquakes whose likelihood is less than 1 in 10 or 1 in 10 per year.

Midland SSER 1 G-2 1

Honorable Nunzio J. Palladino June 8, 1982 The Applicant is currently reevaluating by selective audit the seismic capability of the plant, as originally designed, to withstand the revised SSE. Measures taken to assure safe shutdown in the event of an earthquake include the use of dewatering to reduce the potential for soil liquefaction.

We recommend that all systems and components important to decay heat removal be carefully evaluated for their ability to accomplish necessary functions in the unlikely event of lower-probability, more severe earthquakes in order to provide the necessary degree of assurance. This matter should be re-solved in a manner satisfactory to the NRC Staff. We wish to be kept informed about the resolution of this matter. We believe that any recom-mendations for changes in the plant resulting from this evaluation should be implemented by the end of the second refueling outage.

The Applicant has agreed to provide core exit thermocouples, a hot-leg-level measurement system, and subcooled margin monitors as instrumentation l to detect inadequate core cooling. Consumers Power Company also plans to l include a remotely operable vent on top of both inlet loops to the steam '

generators; however, Consumers has not committed to supply a high point vent on the reactor vessel head. This matter should be resolved in a manner satis factory to the NRC Staff. The ACRS recommends that the Applicant review further the potential for providing indications of water content or level within the reactor vessel.

The staff of the Applicant includes many personnel who have had nuclear power plant experience. However, operating experience with this B&W type power reactor is limited, and the NRC Staff is requiring that at least one person having experience on a large commercial PWR be included on .each shift for one year. We support the NRC Staff position.

The Applicant's experience with the operation of nuclear power plants should, in principle, place Consumers in a favorable position to provide continuing, careful oversight of the operations at the Midland Plant. In view of some prior adverse operating experience at the Palisades Plant however, we recommend that the NRC Staff institute an augmented audit of operations at Midland, at least during the early years of operation at power.

Me have reviewed the evaluation made of the tertiary process steam system for use by Dow Chemical Company. This system appears not to impose any unacceptable impacts either on the safe operation of the Midland Plant or on the people working at the Dow Chemical Company.

The Applicant has undertaken an effort to have a probabilistic risk assess-ment (PRA) performed for the Midland Plant and stated that the results will be available in the fall of 1982. We believe it desirable to have plant-specific PRAs performed for each commercial nuclear power plant and that Midland SSER 1 G-3

l-Honorable Nunzio J. Palladino June 8, 1982 i it is particularly appropriate for the Midland Plant because of its rela-tively high, close-in population density. We wish to have the opportunity to review the Midland PRA with assistance from the NRC Staff, and to offer h comments or recommendations as appropriate. We do not believe that this review need delay licensing of the Midland Plant for operation. I Recently, questions have come to light in connection with B&W plants con-cerning the availability of natural circulation in the presence of an interrupted or continuing small break loss-of-coolant accident. We wish to see a proposed NRC Staff resolution of this issue.

The Applicant described an extensive systems interactions study being undertaken for the Midland Plant. We wish to be informed of the results of this study.

We believe that, in view of the population density near this plant, addi-tional prudence is appropriate for the Midland Plant in the resolution of the ATWS issue and other Unresolved Safety Issues.

We endorse the participation of Dow Chemical Company plant personnel in emergency procedures developed on the basis of an assumed failure at the Midland Plant. Similarly, there should be active participation by Midland Plant personnel in emergency procedures developed on the basis of an assumed failure at the Dow Chemical plant. The Applicant and the NRC Staff should promote continued coordination of these types of relationships, as well as those involving appropriate state and local groups to assure that the capability for an effective emergency respons.e is developed and main-tained.

With regard to the eleven items identified in the ACRS Supplemental Report '

on Midland Plant, Units 1 and 2 dated November 18, 1976, we have the follow-ing comments. The issues related to vibration and loose-parts monitoring, potential for axial xenon oscillations, behavior of core-barrel check valves during normal operation, fuel handling accidents, effects of blowdown forces on core internals, LOCA-related fuel rod failures, and improved quality assurance and in-service inspection for the primary system have all been resolved or are in a confirmatory stage of being resolved. Separation of protection and control equipment has been accomplished in an appropriate manner; however, the safety implications of control systems remains an Unresolved Safety Issue directly applicable to Midland. Resolution awaits completion of the NRC Staff Task Action plan A-47. The effect of ECCS induced thermal shock on pressure vessel integrity has been resolved in part; however, the Unresolved Safety Issue on pressurized thermal shock will apply. Environmental qualification of equipment remains a generic Midland SSER 1 G-4 i

Honorable Nunzio J. Palladino June 8, 1982 il t' issue which is under review by the NRC Staff and whose resolution will apply to the Midland Plant. Instrumentation to follow the course of an l accident has been resolved in part by the development of revised Regulatory Guide 1.97. We do not believe that licensing of the Midland Plant for ,

operation need await further resolution of any of the eleven issues dis- l cussed above.  !

The various other matters identified by the NRC Staff as open or confirma-tory in the Safety Evaluation Report should be resolved in a manner satis-factory to the NRC Staff. We wish to be kept advised concerning resolution of the turbine missile issue.

The ACRS believes that, subject to satisfactory completion of construction and staffing and if due regard is given to the comments above, the Midland Plant, Units 1 and 2 can be operated at power levels up to 5 percent of full power with reasonable assurance that there is no undue risk to the health l and safety of the public.

We defer our recommendation regarding operation at full power until we have had the opportunity to review the plan for an audit of plant quality and the proposed resolution of the question regarding natural circulation in the presence of a small break LOCA.

Dr. Kerr did not participate in the Committee's review of this matter.

Sincerely,

\.

P. Shewmon Chairman

References:

1. Consumers Power Company, " Midland Plant Units 1 and 2 - Final Safety Analysis Report" including Amendments 1-43

2. U.S. Nuclear Regulatory Commission, " Safety Evaluation Report Related to the Operation of Midland Plant, Units 1 and 2," NUREG-0793, dated May 1982

3. U.S. Nuclear Regulatory Commission, "NRC Licensee Assessments,"

NUREG-0834, dated August 1981

4. Letter from J. Cook, Consumers Power Company, to J. Keppler, NRC, Subj ect : Midland Project Response to Draft SALP Report, dated May 17,1982

5. Letter from J. Cook, Consumers Power Company, to J. Keppler, NRC, Subject : Midland Project Quality Assurance Program Update, dated April 30,1981 Midland SSER 1 G-5

Honorable tbnzio J. Palladino June 8, 1982 i

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6. Letter from J. Hind, NRC, to J. Cook, Consumers Power Company, Subject : Systematic Assessment of Licensee Performance (SALP),

dated April 20, 1982

7. Letter from J. Cook, Consumers Power Company, to H. Denton, NRC, Subject : Summary of Soils-Related Issues at the Midland Nuclear l Plant, dated April 19, 1982

8. Letter from K. Drehobl, Consumers Power Company, to D. Fischer, ACRS, Subject : Midland Project Soils Information, dated April 12, 1982

9. Statement of Ms. M. Sinclair to ACRS, dated June 4,1982

10. Letter from B. Stamiris to Dr. D. Okrent and ACRS Members,

Subject:

Midland OL Review, dated May 29, 1982

11. Letter from M. Sinclair to Dr. P. Shewmon, ACRS,

Subject:

Midland OL Review, dated May 28, 1982

12. Statement by Dr. C. Anderson to ACRS Midland Plant Subcommittee dated May 20-21, 1982

13. Statement by Ms. M. Sinclair to ACRS Midland Plant Subcommittee dated May 20-21, 1982

14. Letter from B. Stamiris to D. Fischer and ACRS Members,

Subject:

Soil Settlement and QA Issues, dated May 20, 1982

15. Letter from M. Sinclair to Dr. C. Siess, ACRS,

Subject:

Midl and Soil Settlement, dated April 26, 1982 l 1

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Midland SSER 1 G-6 l

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ll APPENDIX H ERRATA TO MIDLAND PLANT SAFETY EVALUATION REPORT l Section Pg Change 1.1 1-1 In last sentence of first paragraph, change

" Appendix F "to" Appendix E".

1.1 1-3 In third full paragraph, change last two sen-tences to read, " Appendix E is a list of NRC staff contributors and consultants for the radiological safety review of Midland Plant, Units 1 and 2. Appendix F lists the principal consultants to Consumers Power Company."

1.7 1-14 In Item (7), change "(5.9.4.2)" to "(5.4.4.2)".

1. 8 1-15 In Item (1), change "(2.4.2)" to "(2.4.4)".

1.8 1-16 In Item (25), change "(10.2.3)" to "(10.2.1)".

1.8 1-16 Delete "(28) Applicability of Power Train Code (15.1.2)" since SER Section 15.1.2 accepts POWERTRAIN results on the basis of staff audit calculations.

1. 9 1-17 In Item (8), change "(10.3.5)" to "(10.3.3)".

1.12.4 1-22 In last sentence of first partial paragraph, change " established for" to " established from".

2.1.3 2-5 In fifth sentence of second paragraph, change "27,000" to "37,000".

2.4.6.2 2-27 In seventh sentence of second full paragraph, change " valves" to " values".

2.4.6.3 2.28 In second paragraph, change "15 pgm" to "15 gpm".

3.9.3.3 3-31 In first sentence, delete " surface" 4.4.4.2 4-34 In first and second paragraphs, change "Regula-tory Guide 1.33" to " Regulatory Guide 1.133" 5.2.4.3 5-12 In first paragraph, change " March 5, 1981" to

" March 5, 1982" Midland SSER 1 H-1

Section Page Change ,

5.2.5 5-15 In first paragraph, change Item (3) to read,

" Reactor vessel head closure leakage is moni- (

tored for pressure increase, which is alarmed in I the control room. Reactor coolant pump shaft seal overflow leakage is piped to the reactor building collection header, and this leakage flow is alarmed and monitored in the radwaste control room."

5.3.1.2 5-19 In first paragraph, change "RT NDT valves" to "RT NDT values".

5.3.1.2 5-19 In seventh sentence of first paragraph, change

" valves" to " values" 5.3.2 5-23 In fourth sentence of first full paragraph, change " forthcoming in an to the FSAR" to

" forthcoming in a revision to the FSAR" 5.4.1.1 5-26 In second paragraph, change " stress integrity factor" to " stress intensity factor" 6.1(1) 6-1 In first sentence, change "and" to "an". In last sentence of first paragraph, change "requir-ments" to " requirements".

6.1.1.1 6-2 In third paragraph, exchange the positions of

" borated water" and " disodium phosphate" 6.2.4.4 6-15 In second paragraph, change third sentence to read, "The staff, therefore, concludes that the containment isolation system, including isolation provisions for the branch lines, meets the requirements of GDC 55, 56, or 57."

6.2.7 6-21 In first full paragraph, change " reactor coolant pressure boundary (RCPB)" to " containment pressure boundary" In next paragraph, change "RCPB" to " containment pressure boundary" 6.4 6-35 In first paragraph, second sentence, change

" topic gases" to " toxic gases" 7.2.3 7-9 In last sentence of partial paragraph, change "but not on reactor trip" to "but not on turbine trip".

7.4.1.1 7-23 In first partial paragraph, change "safedown" to

" shutdown".

7.4.1.3 7-24 In fifth sentence of first full paragraph, change " form" to "from" Midland SSER 1 H-2

Section M Change 7.7.1 7-37 In second paragraph, change " starter" to " stator".

7.7.2(2) 7-38 In title, change " Low Signal" to " Low Flow Signal".

9.3.4.3 9-20 and Delete all text of this section and substitute 9-21 " Refer to SER Section 6.1.1.3."

9.5.1.7(2) 9-37 Delete second paragraph (it is repeated by fourth paragraph).

9.5.3 9-24 In first partial paragraph, change " safety-related equipment" to " shutdown equipment". ,

l In last sentence, change "duriing" to "during".

10.4.9.1(5)(c) 10-20 11.4.2 11-9 and In all cases except first, replace "SWS" by 11-10 " solid waste system". Delete "(SWS)".

18.1 18-1 In last sentence of last paragraph, change "is" to "in".

19 19-1 In second paragraph (1) Change " core internols" to " core internals" (2) Add at end, "Effect of pressure vessel in-tegrity on ECCS induced thermal shock (5.3.5)".

19 19-1 In last paragraph, change "fo" to "of".

19 19-1 In last paragraph, last sentence, change "th" to l "the" Appendix C C-8 In last full paragraph, change " Electric Power (USI A-5) Reactor Institute" to " Electric Power Research Institute".

Appendix C C-12 In first full paragraph, fifth sentence, change (USI A-17) "(NUREG-75/087)" to "(NUREG-0800)".

Appendix E E-4 Under Organization for H. Singh, delete "and Waterways Equipment Station".

Appendix E E-4 Under Organization for P. Hadala, delete " Detroit and", and change " Equipment" to " Experimental".

Appendix E E-4 Under Consulting Area for P. Abramson, delete

"; Instrumentation and Control Systems".

Appendix E E-4 Under Consulting Area for S. Halverson, delete

" Computer Analyses of Postulated Accidents and Transients;"

Midland SSER 1 H-3

Section jaf Change Appendix E i-4 Under Consulting Area for W. Apley, delete

, Preservice Inspection".

Appendix E 2.- 5 Under Consulting Area for T. Taylor, delete

" Initial Test Program,".

Appendix E E-5 Under Consulting Area for P. Nagata, delete

, Heavy Loads (NUREG-0612)".

Appendix E E-5 Under Consulting Area for 1. Stickley, delete l

" Materials Engineering,".

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Midland SSER 1 H-4

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) NUREG-0793 BIBLIOGRAPHIC DATA SHEET Supplement No. 1 4 TIT LE AND SUBTITLE (Aad vo&me No. J aparer' eel 2 (Leeve ese** /

Safety Evaluation Report related to the operation of l Midland Plant, Units 1 and 2 3 RECIPIENT S ACCESSION NO 1 AU THOHISI l 5. D ATE REPORT COYPLE TED M ON TH

{ lvtAR June 1982 l 9 PE RF OHVING ORGAN 12ATION N AVE AND VAILING ADoRESS (laciwar 20 Codel DATE REPORT ISSUEO l U. S. Nuclear Regulatory Com:sission 90sta i jvEma Office of Nuclear Reactor Regulation June 1982 Washington, D. C. 20555 6 / tee e bea*l 8 (Leeve kwki 12 $PON50HING ORG ANIZ ATION N AVE AND M AILING ADDRESS /taciwde l<a Coael Same as 9 above in CONTRACT NO 13 T Y PE OF RE POR T PE Rico Cove RE D Itac/ss ve dams) 1 *> suPPLEVE N TARY NOTES 14 (Leave o/** A A Docket Nos. 50-329 and 50-330 16 ABST H ACT /200 voras or less)

This report supplements the Safety Evaluation Report, NUREC-0793, issued May 1982 by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Cot:aission with respect to the application filed by Consumers Power Company, as applicant and owner, for licenses to operate the Midland Plant, Units 1 and 2 (Docket Nos. 50-329 and 50-330). The f acility is located in the city of Midland in Midland County, Michigan. This supplement provides recent information regarding resolution of some of the open items identified in the Safety Evaluation Report and discusses recoc=endations of the Advisory Committee on Reactor Safeguards in its interim report dated June 8, 1982.

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