ML20005A931

Safety Evaluation Report Related to the Operation of Lasalle County Station,Units 1 and 2.Docket Nos. 50-373 and 374. (Commonwealth Edison Company)

ML20005A931
Person / Time
Site:	LaSalle
Issue date:	06/30/1981
From:	Office of Nuclear Reactor Regulation
To:
References
NUREG-0519, NUREG-0519-S01, NUREG-519, NUREG-519-S1, NUDOCS 8107060050
Download: ML20005A931 (80)

Text

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f NUREGE19 Supplement No.1 Safety Evaksation Report related to the operation of LaSalle County Station,

-Units 1 and 2 Docket Nos. 50-373 and 374 Commonwealth Edison Company U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation June 1981

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TABLE OF CONTENTS Pa!Le

~1 INTRODUCTION AND GENERAL DISCUSSION..................................

1-1 1.1 Introduction..................

1-1

1. 9 Outstanding Issues...........

1-2 1.10 License Conditions............................................

1-2 3 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS AND COMPONENTS..............

3-1 3.c Mechanical Systems.and Components.............................

3-1 3.9.2. Dynamic Testing and Analysis of_ Systems, Components and Equipment........................................

3-1 3.9.2.1 Piping Vibration, Thermal Expansion, and Dynamic Effects Test Program..................

3-1 3.10. Dynamic Qualification of Seismic Category I Mechanical and Instrumentation and Electrical Equipment...................

3-1 3.11. Environmental Qualification of Electrical Equipment..........

3-2 0

4 REACT0R..............................................................

4-1 4.2.-Fuel System Design.........................................

4-1 4.2.3 Design Evaluation......................................

4-1 4.2.3.3 Ballooning and Rupture..........................

4-1

.2.3.4 Seismic and Loss-of-Coolant Accident 4

Loadings................

4-2 4.2.5 Conclusions............................................

4-3.

4.3 Nuclear Design................................................

4-3 4.3.2 Design' Description.....................................

4-3 4.3.2.8 Continuous Rod Withdrawal During Reactor Startup.......................................

4-3 4.3.2.9 -Rod Withdrawal Error at Power...................

4-4 5 REACTOR'C0OLANT SYSTEM AND CONNECTED SYSTEMS.........................

5-1

5. 2 - Integr_ity.of Reactor Coolant Pressure Boundary................

5-1 i

TABLE OF CONTENTS (continued)

.Page 5.2.2 overoressurization Protection..........................

5-1 5.2.2.1 Safety / Relieve Valve Surveillance Program.......

5-1 5.3 Reactor'Vesse1................................................

5-1 5.3.1 Compliance with Code Requirements......................

5-1 5.3.1.1 Compliance'to Appendix G and H to 10 CFR Part 50........................................

5-1 5.3.1.2 Conclusions..................

5-3 5.3.2 Pressure-Temperature Limits............................

5-3 5.3.3 : Reactor Vessel Integrity...............................

5-4 6 ENGINEERED SAFETY FEATURES...........................................

6-1 6.2 Containment _ Systems...........................................

6-1 6.2.7 Containment Pressure Bo'indary Fracture Toughness.......

6-1 6.3 Emergency Core Cooling System.................................

6-2 6.3.2 Evalution..............................................

6-2 6.3.2.3 Functional Design...............................

6-2 6.3.4 Performance Evaluation.................................

6-2 7 INSTRUMENTATION AND CONTR0L.........................................

7-1 7.7 Control Systems not Required for Safety....................

7-1 7.7.3 Specific Findings......................................

7-1 7.7.3.4 Additional Concerns.......................

7-1 8 ELECTRIC P0WER.............

8-1 8.4 Other Electrical Features and Requirements for Safety.........

8-1 8.4.6 Compliance with Recommendations of Regulatory Guide 1.75.......

8-1 8.4.6.1 Physical Idantification of Electrical Cables....

8-1 9 AUXILIARY SYSTEMS.........................................

9-1 9.1 Fuel Storage and Handling...........

9-1 ii

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TABLE OF CONTENTS (continued)

P_ age 13 CONDUCT OF'0PERATIONS...............................................

13-l' 13.3 Emergency Planning...........................................

13-1 13.6 Industrial Security..........................................

13 l16~

TECHNICAL SPECIFICATIONS...........................................

16-1 17-1 L

17 QUALITY ASSURANCE..................................................

17.3 Conclusions..................................................

17-1 f

18 REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS.............

18-1 22-1 22 TMI-2 REQUIREMENTS................................................

22.2 TMI Action Plan Requirements for Applicants for Operating Licenses...................................................

22-1 I

Operational Safety......................................

22-1 I.C Operational Procedures...........................

22-1 t

I.C.8 Pilot Monitoring of Selected Emergency Procedures for NTOL Applicants..........

22-1 I.D.1. Control Room Design Review.......................

22-2 r

II Siting and Derign.......................................

22-2 II.B.7 Analysis of Hydrogen Contro1....................

22-2 II.B.8 Rulemaking Proceedings on Degraded-Core Accidents.....................................

22-2 i.

II.E.2 System Design...................................

22-3 II.E.4.2 Containment Isolation Dependability...

22-3 II.F.2 Instrumentation for Detection of Inadequate 22-3 Core Cooling..................................

II.K.3 Final Recommendations of Bulletins and Orders 22-9 Task Force....................................

Item 18 Modification of Automatic Depressurization System Logic -- Feasibility for Increased 22-9 Diversity for Some Event Sequences.........

Item 44 Evaluation of Anticipated Transients with 22-9 Single Failure to Verify no Fuel Failure...

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TABLE OF CONTENTS (continued)

Page III Emergency Preparations and Radiation Protection.........

22-10 III.A

'NRC and Licensee Preparedness...................

22-10 III.A.1.2 ~ Upgrade Emergency Support Facilities.........................

22-10

)

III.A.2 Improving Licensee Emergency Preparedness -

Long-Term.....................................

22-11 APPENDICES A.

CONTINUATION OF CHRON0 LOGY FOR THE LA SALLE COUNTY STATION........

A-1 B.

LETTER FROM THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS..........

B-1 C.

HUMAN FACTORS ENGINEERING BRANCH CONTROL ROOM DESIGN DEFICIENCIES....................................................

C-1 D.

ERRATA TO THE SAFETY EVALUATION REP 0RT............................

D E.

SUPPLEMENTAL EMERGENCY PREPAREDNESS EVAll'ATION REPORT.............

E-1 F.

LETTER FROM THE FEDERAL EMERGENCY MANAGEMENT AGENCY................

F-1 iv

I' LIST OF FIGURES

~ FIGURE II.F.2.1 BWR SIMULATION-TRUMP CODE CENTER R0D TEMPERATURE........

22-6 FIGURE II.F.2.2 INCORE THERM 0 COUPLE TEMPERATURE RESPONSE................

22-8 FIGURE E.1 LIMITED RESPONSE OFFSITE GSEP ORGANIZATION..............

E-3 FIGURE E.2 FULL RESPONSE OFFSITE GSEP ORGANIZATION.................

E-4 V

c-l 1 INTRODUCTION AND GENERAL DISCUSSION 1.1 Introduction On March 5,1981, the Nuclear Regulatory Commission staff (hereinafter referred to as the_ Commission er NRC staff) issued its Safety Evaluation Report (NUREG 0519) regarding the application by Commonwealth Edison Company (hereinafter referred to as the applicant) for licenses to operate the La Salle County Station, Unit Nos. 1 and 2 (hereinafter referred to as La Salle or facility), Docket Nos. 50-373 and 50-374.

This report is Supplement No.1 to our Safety Evalua-tion Report.

At the time that the report was issued, there were 16 outstanding issues listed in Section 1.9 which had not been resolved with the applicant.

We stated that these matters would be resolved prior to issuing the operating licenses for La Salle.

Since the issuance of NUREG 0519, the Adv.'sory Committee on Reactor Safeguards considered the La Salle operating license application at its 252nd meeting and subsequently 1ssued a favorable letter, dated April 16, 1981, to the Commission (see Appendix B to the report).

The purpose of this supplement to is to update the Safety Evaluation Report by providing:

(1) our evaluation of additional information submitted by the applicant in response to some of the outstanding issues since the Safety Evaluation Report was issued, (2) our evaluation of additional information for

'those sections of the Safety Evaluation Report where further discussion or changes are in order and (3) our responses to the comments made by the Advisory Committee on Reactor Safeguards in its report.

Except for the appendic ~es, each of the following sections of this supplement is numbered the same as the corresponding section of the Safety Evaluation Report that is being updated.

Each section except Section 4.2.3.4 (which is a

. complete revision) and Section 13.3 (which is superseded) is supplementary to and not in lieu o' the discussion in our Safety Evaluation Report.

Appendix A to this supplement is a continuation of the chronology of the staff's actions related to processing of the La Salle application, Appendix B is the letter of the Advisory Committee on Reacter Safeguards on La Salle, Appendix C lists the deficiencies of the control design as a result of our review, Appendix D lists errata to the Safety Evaluation Report, Appendix E is the Supplemental Emergency Preparedness Evaluation Report, and Appendix F is the letter trom the Federal Emergency Management Agency.

As indicated in our Safety Evaluation Report, part of our review of the appli-cation is to appraise the design against the Commission's regulations.

We requested the applicant to verify that La Salle meets the applicable require-ments in 10 CFR Parts 20, 50 and 100.

The applicant responded to this request by a submittal dated March 6, 1981, which contained an indepth comparision of the application with the regulations.

Accordingly, the applicant stated that La Salle complies with the applicable regulations with the exception of those 1-1

instances where specific exemptions have been justified by the applicant and approved by the NRC, i.e., Appendices G, H, and J t.o 10 CFR Part 50.

Based on our review of the applicant's response, our review of its application for la Salle operating licenses, and subject to the resolution of the remaining outstanding issues, we have determined that La Salle will operate in conformity with the provisions of.the Act and the rules and regulations of the Commission, and that there is reasonable assurance that the activities that would be authorized by the operating licenses for these plants can be conducted without endangering the health and safety of the public.

1.9 Outstanding Issues Th; resolution of the outstanding issues listed in Section 1.9 of NUREG 0519 is discussed in the appropriate section of this supplement.

The status of the following issues, which are still outstanding, is also discussed herein:

(1) Dynamic qualification (3.10).

(2) Environmental qualification (3.11).

(3) Adequate barriers or justification for separation between Class 1E and non-Class 1E raceways (8.4.6)

(4) TMI issue (22.2)

(a)

III.A.2 Improving Licensee Emergency Preparedness - Long Term 1.10 License Conditions Since the issuance of our Safety Evaluation Report, additional issues were identified in our review where the operating licenses will be conditioned.

The additional items listed below are discussed further in the sections of this supplement as indicated.

(1) Recalculation of the pressure-temperature limit curves after removal of each surveillance capsule from the reactor vessel (5.3.2)

(2) Scram discharge system pipe break (6.3.2.3, 18)

(3) Additional concerns (7.7.3.4)

(4) Industrial security (13.6)

(5) TMI issues (22.2)

(a)

I.C.8 Pilot Monitoring of Selected Emergency Procedures for NT0L Applicants (b)

II.F.2 Instrumentation for Detection of Inadequate Core Cooling (c)

III.A.1.2 Upgrade Emergency Support Facilities (d) -III.A.2 Improving Licensee Emergency Preparedness - Long Term l

1-2 m

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3 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS AND COMPONENTS 3.9 Mechanical Systems and Components 3.9.2 Dynamic Testing and Analysis of Systems, Components and Equipment 3.9.2.1 Piping Vibration, Thermal Expansion and Dynamic Effects Test Program In our Safety Evaluation Report (NUREG 0519), we expressed concern with respect to the applicant's vibration program which excluded visual inspection of branch piping less than 2 inches in diameter. Our position was that, as a minimum, we require that either a visual or instrumented inspection (as appropriate) be conducted to identify any excessive vibrations that will lead to fatigue failure.

In the applicant's letters from i.. O. DelGeorge to B. J. Youngblood dated March 27, 1981 and April 7, 1981, the applicant has committed to the following vibration monitoring program:

(1) The reactor pressure vessel level instrumentation lines and the control rod drive lines inside containment will be visually inspected to identify any excessive vibration in conjunction with Startup Test Procedure 34,

" Vibration Measurements."

(2) The reactor core isolation cooling instrumentation lines on the reactor core isoletion cooling steam line will be visually inspected to identify any excessive vibration as part of the Drywell Piping Vibration Test.

(3) The main steam line flow instrumentation lines will be visually examined and a line walkdown will be performed prior to fiel loading.

The inspec-tion will be coordinated with the Region III Office of Inspection and Enforcement and will include the participation o an applicant's engineer qualified in the dynamic design of piping and sepports.

If resupporting of any of these lines is judged to be required, the additional supports will be installed prior to pressurization of the lines with steam.

We have reviewed the applicant's response and find the resolution to be acceptable.

3.10 Dynamic Qualification of Seismic Category I Mechanical and Instrumentation and Electrical Equipment In the Safety Evaluation Report, we concluded that an appropriate qualification program has been defined for the seismic Category I mechanical and eledtrical equipment which will provide adequate assurance that such equipment will function properly during and after the excitation from vibratory forces imposed by the safe shutdown earthquake or hydrodynamic loads associated with discharges into the suppression pool, or by the combined earthquake and hydrodynamic loads.

In c w r to complete our review we requested the applicant to provide the following information:

3-1

(1) Provide completed SQRT " Qualification Summary of Equipment" forms for nuclear steam supply system and balance-of plant equipment identified as "0 PEN" in the applicant's lecter of December 12, 1980.

(2) Provide the results and conclusions of the fatigue evaluations.

-(3) Provide results and conclusions of the impedance testing program.

(4) Provide the results and conclusions of the reassess. ment of valve qualifica-tions.

(5) Provide clarifying details concerning the qualification of some pieces of equipment as discussed in our trip report.

The applicant has provided additional information in its letters of December 12, 1980 and February 2 and March 19, 1981. We have reviewed these submittals and have found them acceptable with regard tc the resolution of the qualification of most of the equipment identified under item (5) above.

However, the applicant has not completed the qualification program and has not provided all of the necessary information to resolve the issues relating to items (1) through (4) above.

In addition, the resolution of the qualification of some of the equipment included under item (5) relies on the results of the impedance testing program, additional qualification testir., and the results of a valve flexibility study which is to be conducted by the applicant.

The applicant has committed to provide all of the above information necessary to resolve the outstanding issues by June, 1981.

We will complete our review after the applicant has provided the above infor-mation and has documented the completion of its dynamic qualification program.

We will report on the results of our final evaluation of the applicant's program in a supplement to this report.

3.11 Environmental Qualification of Electrical Equipment In our Safety Evaluation Report, we indicated that the applicant's environmen-tal qualification submittal dated November 1 and 18, 1980, had insufficient information.

In addition, we pointed out that a discussion was held with the applicant on February 13, 1981, at which time the applicant committed to provide a new and complete environmental qualification submittal by March 2.

1981, considering the criteria as outlined in NUREG-0588, " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment."

The applicant indicated that this additional information would be transmitted in July 1981.

However, some information was received in a letter dated Ju.ie 4, 1981, and as a result, our site audit review is underway. We will report our findings of this audit 11 another supplement to NUREG-0519.

3-2 l

4 REACTOR 4.2 Fuel System Design 4.2.3 Design Evaluation 4.2.3.3 Ballooning and Rupture In our Safety Evaluation Report, we indicated that we are generically evaluating three fuel material models that are used in emergency core cooling system eval-uations.

Because we.had not completed our generic review and implemented new acceptance criteria for cladding models, we required that the emergency core cooling system analyses in the Final Safety Analysis Report be accompanied by supplemer.tal calculations to be performed with the materials models of our NUREG-0630, " Cladding Swelling and Rupture Models for LOCA Analysis."

In a letter from L. O. DelGeorge to A. Schwencer dated May 21, 1981, the applicant has responded to our request by endorsing the results of a generic sensitivity study performed by General Electric submitted to the NRC by letter dated May 15, 1981.

As reported in this generic study, Generas Electric has assessed the boiling water reactor emergency core cooling system sensitivity to rupture temperature by using three rupture temperature models:

(1) the General Electric CHASTE model, (2) the NUREG-0630 model, and (3) a proposed General Electric model termed the adjusted model.

For the La Salle type of 8 x 8 with 2 water rod fuel design (designated the " improved 8 x 8 design"),

General Electric found that the use of the NUREG-0630 model resulted in an increased peak cladding temperature of up to 50 degrees Fahrenheit over that which was obtained with the CHASTE model.

However, sensitivity studies per-formed on the adjusted model, which is a combination of the CHASTE and NUREG-0630 models and may be the better of the three models, found the maximum impact on peak cladding temperature to be 5 10 degrees Fahrenheit.

As stattd in the General Electric submittal, a complete loss-of-coolant accident model improvement package will be submitted that will request generic approval of the adjusted rupture temperature model.

Pending the completion of our review o' + hat submittal, we find that our concerns related to the rupture temperature nod.1 are resolved for the La Salle application on the basis of the small ie'sitivity (i.e., 510 degrees Fahrenheit) to rupture temperature.

ragard to the boiling water reactor emergency core cooling system sensi-A

. L'y to burst strain, the General Electric submittal assessed the impact of using a burst strain model that bounds the burst strain model given in NUREG-0630.

However, we have questioned General Electric and found that prior to performing this comparison, the bounding burst strain model was appreciably reduced by axially averaging the clmding strain. Two reduction factors were used to effect this averaging promess; for the region of plant application, the reduction factor was 4.1.

4-1

We are uncertain as to the appropriateness of comparing the reduced strain model to the CHASTE strain model; however, we have estimated from the impact (i.e.,

< 5 degrees Fahrenheit) of the reduced versus the CHASTE model comparison that if the comparison had been made against the unaltered NUREG-0630 strain model, the impact would have been < 115 degrees Fahrenheit.

In light of the calcu-lated 2009 degrees Fahrenheit loss-of-coolant accident peak cladding tempera-ture for La Salle, we thus conclude that sufficient margin exists between the 2200 degrees Fahrenheit peak cladding temperature limit as required by 10 CFR 50.46 and the calculated 2009 degrees Fahrenheit La Salle peak cladding temperature to accommodate an uncertainty of 115 degrees Fahrenheit in the peak cladding temperature.

Consequently, for la Salle we conclude that our concerns related to the burst strain model are resolved, and even in the boiling water reactor loss-of-coolant accident analyses not using flow blockage models, we conclude that all concerns related to the issue of cladding ballooning and rupture are resolved for la Salle.

4.2.3.4 Seismic and Loss-of-Coolant Accident Loadings Analytical results for the fuel assembly response to the effects of an earth-quake and a loss-of-coolant accident are described in a General Electric Report, NEDE-21175-P, "BWR/6 Fuel Assembly Evaluation of Combined Safe Shutdown Earth-quake (SSE) and Loss-of-Coolant Accident (LOCA) Leadings," November 1976.

These results are said to apply also to BWR/4 and BWR/5 fuel assemblies.

We found the analytical methods in NEDE-21175-P acceptable, but the generic fuel assembly design limits in that report were not accepted.

. Essentially, the following two design limits were in question:

(1) Criteria for allowable stress values, and (2) Margin for fuel assembly liftoff.

The original NEDE-21175-P evaluations were based on 0.3g horizontal accelera-tions.

The La Salle site seismic ground motion requirement is 0.2g horizontal (for a safe shutdown earthquake).

Based on additional information provided by the applicant in a letter from L. O. De1 George to B. J. Youngblood dated February 12, 1981, this ground motion at La Salle would produce fuel assembly loads equal to approximately 50 percent of those used in the NEDE-21175-P evaluation.

Based on these smaller horizontal loads, we believe that adequate stress margins exist for La Salle, thus permitting acceptance of the fuel assembly stress evaluations.

Preliminary analysis for fuel assembly liftoff, performed by our consultant, indicated that a liftoff condition may be possible as shown by the Idaho National Engineering Laboratory Report, EGG-EA-5151, " General Electric Fuel Assembly Liftoff Reassessment," May 1980.

However, the conservatisms inherent in the analytical approach suggested that significant liftoff is unlikely.

In addition, a more detailed evaluation of liftoff has been performed by General Electric in response to a generic concern, and submitted in a letter dated June 8, 1981.

A review of the results from this evaluation oy us has indicated that the amount of liftoff which will occur for La Salle's safe shutdown earthquake input (0.2g) is insignificant.

The amount of vertical liftoff is not sufficient to disengage 4-2 e

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the lower tie plate from the fuel support casting such that the resulting loss of lateral fuel bundle positioning could interfere with control blade insertion.

Based on the above information, we consider this issue resolved for La Salle.

4.2.5 Conclusions With the resolution of the ballooning and rupture concern, we conclude that the fuel systems of La Salle have been designed such that:

(1) the fuel system will not be damaged as a result of normal operation and anticipated operational occurrences; (2) fuel damage during postulated accidents will not be so severe as to prevent control rod insertion when it is required; (3) the number of fuel rod failures will not be underestimated for postulated accidents; and (4) core coolability will always be maintained, even after severe postulated accidents.

The applicant has provided sufficient evidence that these design objectives will be met based on operating experience, prototype testing, and analytical predictions.

The applicant has also provided for testing and inspection of new fuel to ensure that it is within design tolerance.

In addition, the applicant has met all the requirements of the applicable regulations, current regulatory positions, and good engineering practice.

4.3 Nuclear Design 4.3.2 Design Descripti_on In our Safety Evaluation Report, we did not discuss design aspects in two areas:

(1) continuous withdrawal during reactor startup, and (2) rod withdrawal error at power.

Our evaluation of these design aspects are presented in Sections 4.3.2.8 and 4.3.2.9 below.

4.3.2.8 Continuous Rod Withdrawal During Reactor Startup Discussion The rod sequence control system and rod worth minimizer are each designed to enforce a particular rod withdrawal sequence.

Following that sequence, the amount of reactivity that could be inserted in one withdrawal action would be limited to an amount that would preclude any violation of fuel thermal limits (the programmed withdrawal sequence constitutes normal operation during startup).

The probability of a failure in these systems that would permit the continuous withdrawal of a high worth rod is low.

Nevertheless, the consequences of such an event have been calculated since it cannot be assured that a single failure in these systems will not cause the event.

The calculation was performed generically by the vendor, General Electric Company, and is reported in the Final Safety Analysis Report.

The calculation was per-formed in two steps - first a detailed analysis, including three-dimensional effects, was performed for a rod worth (1.6 percent reactivity change)-in the upper range of anticipated worth and then a point kinetics calculation was used to extrapolate the results of rod worths to the expected values for out of sequence rods.

The calculation was performed at one percent of full power because calculations have shown that this power level produces the maximum con-sequences.

Transient termination is assumed to occur by means of the average 4-3

power range monitor at the 15 percent power level scram or the degraded (worst bypass condition) intermediate range monitor scram.

The withdrawal speed is assumed to be the maximum attainable and rod worths up to 2.5 percent reactivity change were analyzed.

In no case was a peak enthalpy greater than 60 calories per gram encountered. We conclude that the analysis of this event is acceptable because a conservative analysis has been performed, conservative rod worth values are analyzed and the consequences show a large margin to the acceptance criterion of 170 calories per gram.

Evaluation Findings The possibilities for single failure of the reactor control system which could result in uncontrolled withdrawal of control rods under low power startup condi-tions have been reviewed.

The scope of the review has included investigations of initial conditions and control rod reactivity worths and the course of the resulting transient.

The methods used to determine the peak fuel rod response and the inputs to that analysis have been examined.

The requirements of Criteria 10, 20, and 25 of the General Design Criteria con-cerning specified acceptable fuel design limits are assumed to be met for this event when the peak fuel enthalpy gemrated during the transient is less than 170 calories per gram.

The power transient resulting from this event is very narrow (+200 millisecond full width at half maximum) so that fuel enthalpy is an appropriate measure of fuel duty.

We, therefore, conclude that the require-ments'of Criteria 10, 20, and 25 of the General Design Criteria have been met.

The basis for this conclusion is discussed in the following paragraph.

The applicant has met the requirement of Criterion 10 that specified acceptable fuel design limits are not to be exceeded, Criterion 20 that specifies reactivity control systems are to be automatically initiatea so that specified acceptable fuel design limits are not exceeded, and Criterion 25 that specifies single malfunctions of the reactivity control system shall not cause the acceptable fuel design limits to be exceeded.

These requirements have been met by compar-ing the resulting extreme operating conditions for the fuel (i.e., fuel duty) with the acceptance criterion (peak fuel enthalpy)-to assure that fuel rod failure will be precluded for this event.

The basis for acceptance in our l

review is that the applicant's analysis of the maximum low power condition have been confirmed, that the analytical methods and input data are reasonably j_

conservative and that specified acceptable fuel design limits will not be exceeded.

4.3.2.3 Rod Withdrawal Error at Power l

l Discussion l

l Above a preset power level (approximately 25 percent of full power) the rod withdrawal sequence is no longer enforced by the rod sequence control system or the rod worth minimizer.

Instead the core is protected against exceeding fuel damage limits by the rod block monitor. When a rod is selected for with-drawal, the nearest four strings of local power range monitors are also selected.

The outputs from these monitors serve as inputs to the rod block monitor.

The output from the rod block monitor is the average of the detector signals.

This output is then input to a trip circuit which is adjusted to block the rod with-drawal before fuel damage limits are exceeded.

An analysis was performed to establish the trip setpoint required to accomplish the proper rod block.

l 4-4

The analysis was performed in a conservative manner by assuming the highest worth rod in a pattern to be continuously withdrawn at its maximum speed.

A rod pattern was selected which tends to maximize the consequences though such a pattern would be prohibited during normal operation.

The core was assumed to be operating at rated conditions.

The two local detector strings having the highest readings were assumed to be inoperable.

The rod to be withdrawn was assumed to be fully inserted prior to its withdrawal.

The calculation was performed with the BWR Simulator Code which has been reviewed and approved (our acceptance letter is dated September 22,1976).

This three-dimensional code is suitable since the power rise is slow enough to permit the assumption that the neutron and thermal powers can be calculated by time-

-independent methods.

The core was assumed to be xenon-free for this calculation in order to maximize the reactivity controlled by control rods.

The calculation consists of a number of " snapshots" of the core power distribution as the control rod is withdrawn.

The results of the calculations were examined and curves of linear heat gener-ation rates and critical power ratio were plotted as a function of withdrawal distance for the rod.

Such curves were drawn for assemblies containing the highest linear heat generation rate and lowest critical power ratio during the transient. The curves were used to obtain the maximum travel for the rod which will not violate established heat generation rate or critical power ratio limits.

The more limiting of the two maximum travel distances was then chosen.

For La Salle, the analysis yields a minimum critical power ratio change of 0.208 for a control rod travel of 5.5 feet and an increase of 1.5 kilowatts per foot in the linear heat generation rate.

Final values for these quantities are 1.09 and 15.9 kilowatts per foot, respectively.

Acceptable design limits are 1.06 minimum critical power ratio and 19 to 21 kilowatts per foot in the linear heat generation rate.

The calculations are also used to obtain the responses of the local detectors which provide inputs to the rod block monitor.

These are combined in the appropriate manner to obtain the output of the rod block monitor as a function of rod withdrawal.

Appropriate assumptions regarding inoperable local detectors are made.

The results of the calculations are plotted and the rod block monitor trip is set to the value obtained for the maximum permitted travel.

On the basis that the calculational method used is an approved one and that conserva-tive input assumptions are made, we conclude that the analysis of this event is acceptable.

Evaluation Findings The possibilities for single failures of the reactor control system which could result in uncontrolled withdrawal of control rods beyond normal limits under power operation conditions have been reviewed. The scope of the review has included investigations of possible initial conditions and the range of reactivity insertions, the course of the resulting transient and the instrumen-tation response to the transient.

The methods used to determine the peak fuel rod response, and the initial conditions for that analysis have been examined.

We conclude that th,e requirements of Criteria 10, 20 and 25 of the General Design Criteria have been met.

The applicant has met the requirement of Criterion 10 4-5

that specified acceptable fuel design limits are not exceeded for the antici-pated transient, Criterion 20 that specifies reactivity control systems are automatically actuated to prevent exceeding the specified acceptable design limits, and Criterion 25 that specifies single malfunctions in the reactiv').y control system will not cause the specified acceptable fuel design limits to be exceeded.

These requirements have been met by comparing the resulting extreme operating conditions and response of the fuel (i.e., fuel duty) with the acceptance criteria for fuel damage (boiling transition and one percent plastic strain the cladding) to assure that fuel rod failure will be precluded for this event.

The basis for acceptance in our review is that the applicant's choice of maximum transients for single error control rod malfunctions has been con-firmed, that the analytical methods and input data are reasonably conservative and that specified acceptable fuel design limits will not be exceeded.

~.o 4-6

5 REACTOR COOLANT SYSTEP_ ann CONNECTED SYSTEMS 5.2 Integrity of Reactor Coosant Pressure Boundary 5.2.2 Overpressurization Protection 5.2.2.1 Safety / Relief Valve Surveillance Program We have discussed wit,. the General Electric Company the implementation of a program to monitor the performance of safety / relief valves in boiling water reactors. We noted the purpose of the program in-Section 5.2.2 of NUREG-0152

" Safety Evaluation Report related to the preliminary design'of the GESSAR-238 Nuclear Steam Supply System Standard Design" issued in March 1977. We require that the applicant participate in program to monitor the performance of safety / relief valves in boiling wa w r reactors.

The applicant, in letters dated May 19, 1981 and May 21, 1981 from L. O.

DelGeorge (Commonwealth Edison) to A. Schwencer (NRC), has committed to estab-lishing a formal record for each safety / relief valve that will include signif-icant maintenance ope qting information.

This record will be maintained at the plant site and will oe provided to us on request.

In addition, the applicant will work in conjunction with other boiling water reactor owners whose plants

' employ Crosby valves, s'imila'r to those at La Salle, to establish and implement a joint program for collecting and disseminating valve pe-formance data. We will review the status of this program as more plants using Crosby valves are licensed.

The applicant's commitments in this regard are acceptable for licensing of La Salle.

5.3 Reactor Vessel 5.3.1 Compliance with Code Requirements 5.3.1.1 Compliance to Appendix G and H to 10 CFR Part 50 In our Safety Evaluation Report, we specified that except for Paragraphs IV. A.1 and IV.B of Appendix G for Unit No. 1 we have found that the alternate eethods proposed by the applicant to demonstrate compliance with Appendices G and H had sufficient safety margins.

In Amendment 56 of the Final Safety Analysis Report, the epplicant stated that where dropweight tests were not conducted for beltline weld seam *, the applicant assumed a value 01 no higher than

-50 degrees Fahrenheit for the nil-ductility temperature.

The applicant based this assumption on dropweight tests conducted for several of the beltline weld seams.

The weldment specimens tested were fabricated using 8018 weld wire with shielded metal arc welding for the root pass and MILB-4 weld wire, 1092 flux, with submerged arc welding for the remainder of the weld. All specimens were post weld heat treated at 1150 degrees Fahrenheit for 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />.

Dropweight test values obtained ranged from -80 degrees Fahrenheit to -50 degrees Fahrenheit.

Our review of these data indicate that a nil-ductility temperature value of no higher than -50 degrees Fahrenheit for the beltline weld seams is conservative 5-1

_- ~

for the following reasons:

(1) the applicant has supplied sufficient test data, and (2) all weld seams in the beltline have been fabricated using the same weld wire type, flux type, welding process and post veld heat treatment as used in the test specimens.

Although each weld seam was not individually dropweight tested, we conclude, based on the above, that a nil-ductility temperature value of -50 degrees Fahrenheit along with each weld seams Charpy impact data results in values of a reference temperature for nil-ductility t mperature, RTNDT, that are equivalent to those that would be determined if the tests were conducted in strict compliance to Appendix G.

In conclusion, the RT determined for the vessel plate and weld material NDT will be conservative and equivalent to the requirements of Paragraph IV.A.1 of Appendix G.

Based on the above evaluations, an exemption to the exact require-5 mants of Paragraph IV.A.1 is recommended for Unit No. 1.

For Paragraph IV.B of Appendix G, the applicant had to demonstrate that the weld materials possessed 75 foot pounds of Charpy V notch impact. energy at some test temperature, or by using Regulatory Guide 1.99, Revision 1, " Effects of Residual Elements oa Predicted Radiation Damage to Reactor Vessel Materials,"

show that a lower value of initial upper shelf energy will still provide an end-of-life upper shelf energy of 50 foot pounds.

To comply with Paragraph IV.B by the latter method, the noncempliance welds would have to have at least the followin 1.4x10g8 levels of Charpy energy, based on an end-of-life fluence of neutrons per square centimeter at the quarter thickness wall location:

Lowest reported Minimum USE value Percent USE value @ +10 F req'd by RG 1.99 Weld seam copper foot pounds foot pounds 4-308-A, B, C 0.33 66 70 3-308-A, B, C 0.37 64 74 To demonstrate compliance of weld seams 4-308-A, B and C (4-300) and 3-308-A, B and C (3-308) to the upper shelf requiremnts of Paragraph IV.B, the applicant has supplied additional data and analyses to demonstrate that at a higher temperature, weld seams 3-308 ar.d 4-308 do indeed exhibit greater than 75 foot-pounds of Charpy 'mpact energy.

Additional weldment test specimens for seams 3-308 have been tested at 200 degrees Fahrenheit, and values of 111, 110 and 109 foot pounds were obtained.

The applicant has stated that these results represent weld seams 4-308 for the following reasons:

(1) limiting results for 4-308 weld materials at +10 degrees Fahrenheit (82, 66, 80, 92, 91, 92 foot-pounds) are even higher than those for 3-308; (2) both seams 4-308 and 3-308 have been made by the same fabricator, using MILB-4 weld wire, 1092 flux, and the tandem wire, submerged arc process; and (3) each seaa has been post weld heat treated at 1150 degrees Fahrenheit for 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />.

Based on these data and analyses we find that weld set.ms 4-308 and 3-308 are similar welds, and there-fore, upper shelf energy data for seam 3-308 can be used to demonstrate that seam 4-308 meets the minimum upper shelf requirments of 75 foot pounds.

According to our evaluation, we conclude that an exemption to Paragraph IV.B, requiring that each ferritic beltline material be individually tested to determine a minimum upper shelf energy of 75 foot pounds, is justified based 5-2 r

on the following:

(1) weld seams 3-308 and 4-308 are similar welds, having been made by the same fabricator, using the same weld wire, flux, welding process and post weld heat treatment, and (2) data for seams 3-308 and thus, seams 4-308, indeed show that at higher temperatures values greater than 75 foot pounds are obtained.

Based on these conclusions, an exemption to Paragraph IV.B is recommended for Unit No. 1.

5.3.1.2 Conclu'sions In our Safety Evalution Report, we indicated that except to Paragraphs IV.A.1 and IV.B of Appendix G, the alternate methods proposed by the applicant to demonstrate that compliance with Appendix G and H for Unit No. I have been achieved and exemptions were justified.

As a result of the additional informa-tion provided by the applicant, the exemptions requested for Paragraphs IV.A.1 and IV.B for Unit No. I are granted.

5.3.2 Pressure-Temperature Limits In our Safety Evaluation Report, we indicated'that we could not complete our evaluation of the adequacy of the proposed pressure-temperature limits for La Salle. We stated that the RT requirement of Paragraph IV.A.1 of Appendix G NDT was not sufficient for Unit No. 1 and we were provided insufficient information for Unit No. 2.

However, for Unit No. 1, the applicant has supplied the additional informatici for us to make an assessment regarding the adequacy of the proposed pressure-temperature limits.

Appendices' G and H to 10'CFR Part 50 require the applicant to predict the'"

s

. shift in reference temperature due to neutron irradiation.

The shift in RTNDT due to neutron irradiation is then added to the initial RT to establish the NDT adjusted reference temperature.

The base plate or weld seam having the highest adjusted reference temperature is considered the most limiting materials on which the pressure-temperature operating limits are based.

In the case of l

Unit No. 1, the most limiting materials at end-of-life are weld seams 3-308-A, l

-B and -C.

Once in service, the pressure-temperature limits must be revised to reflect the actual neutron radiation damage as determined from the results l

of the reactor vessel materials surveillance program.

According to our evaluation, the proposed heatup and cooldown pressure-temperature limits (Figures 3.4.6.1-2 and -3 of the Technical Specifications and Figure 5.2-8 of the Final Safety Analysis Report) are acceptable for Unit i

No. 1 based on the following time limits:

(1) the limit curves representing the stage in which the feedwater nozzles are the controlling factor (solid l

lines in Figures 3.4.6.1-2 and -3, and curves B and C in Figure 5.2-8) are l

acceptable for the first 10 effective full power years, and (2) the limit curves representing the stage in which the beltline weld seams, 3-308A, B and C, are the controlling factor due to neutron irradiation effe-ts (slashed lines in Figures 3.4.6.1-2 and -3, and curves B' and C' in Figure 5.2-8), are acceptable for the first 20 effective full power years.

It is our position that upon removal of every surveillance capsule, the applicant must recalculate the pressure-temperature limits based on the greater of the following:

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(1) the actual measured shifts in RT as determined by the materials sur-NDT veillance program of La Salle No. 1, or (2) the predicted shift in reference temperature for weld i,eams 3-308A, B and C, as determined by Regulatory Guide 1.99.

The data obtained must be compared to those that were used to develop the pressure-temperature limit curves in the Technical Specifications.

If this information indicated anomalies to the then-existing predictions, the curves must be redrawn to reflect the actual or predicted shif t in RT as discussed NDT in items (1) and (2) above. We will condition the license for La Salle Unit No. I to recalculate the pressure-temperature limits after the removal of each surveillance capsule.

Since the proposed initial pressure-temperature limits based on the feedwater nozzles as the controlling factor are acceptable for 10 effective full power years, and the limits based on the beltline weld seams, 3-308A, B and C, as the controlling factor are acceptable for 20 effective full power years, we will limit their use to those intervals in the Technical Specifications.

The pressure-temperature limits to be imposed on the reactor coolant system for all operating and testing conditions, to assure adequate safety margins against nonductile or rapid 1/ propagating failure are in conformance with established criteria, codes and standards acceptable to us.

The use of operat-ing limits based on these criteria, as defined by applicable regulations, codes, and standards, provides reasonable assurance that nonductile or rapidly propagating failure will not occur, and constitutes an acceptable basis for identifying the applicable requirements of Criterion 31 of the General Design Criteria.

5.3.3 Reactor Vessel Integrity In our Safety Evaluation Report, we stated that we reviewed the information in the Final Safety Analysis Report according to Section 5.3.3 of the Standard Review Plan for all areas contributing to the structural integrity of the reactor vessels. We concluded that the applicant complied with Appendices G and H, 10 CFR Part 50 for Unit No. 1 except for the Paragraphs IV.A.1 and IV.B of Appendix G and that insufficient information was provided to determine Unit No. 2.

In Amendment 56 to the Final Safety Analysis Report, the applicant has submitted sufficient additional information to justify exemptions for Paragraphs IV.A.1 and IV.B for Unit No. 1.

We reviewed all factors contributing to the structural integrity of the reactor vessel and conclude that there are no special con-siderations that make it necessary to consider potential reactor vessel failure for Unit No. 1.

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6 ENGINEERED SAFETY FEATURES

6. 2 Containment Systems 6.2.7 Containment Pressure Boundary Fracture Toughness In our Safety Evaluation Report, we indicated that we did not have sufficient information to conclude that the reactor containment pressure boundary meets Criterion 51 of the General Design Criteria because the Final Safety Analysis Report did not adequately characterize the fracture toughness of those materials.

The applicant in a letter dated May 26, 1981 from L. O. DelGeorge to A. Schwencer provided this information.

In our review, we considered the ferritic material components of the containment system which are load bearing and provide a pressure boundary in the performance of the containment function under operating, maintenance, testing and postulated accident conditions as addressed in Criterion 51.

These components are the drywell head, equipment hatch, personnel airlock, penetrations, anti elements of the main steam and main feedwater systems.

In some cases, materials were not fracture toughness tested or were inapprop-riately tested.

Generally, those materials, which were not fracture toughness tested, were not tested because the American Society of Mechanical Engineers Code edition and addenda in effect at the time the components were ordered did not require that they be tested.

Our assessment of the fracture toughness of materials that were not fracture toughness tested or that were inappropriately tested is based on the metallurgical characterization of these materials and fracture toughness data presented in NUREG 0577, " Potential for Low Fracture Toughness and Lamellar Tearing on PWR Steam Generator and Reactor Coolant Pump Supports," October 1979 and the American Society of Mechanical Engineers Code Section III, Summer 1977 Addenda, Subsection NC.

l The metallurgical characterization of these materials, with respect to their fracture toughness, was developed from a review of how these materials were fabricated and what thermal history they experienced during fabrication.

This characterization, when correlated with the data presented in NUREG 0577 and the Summer 1977 Addenda of the American Society of Mechanical Engineers Code Section III, provided the technical basis for our evaluation of compliance with the code requirements for materials that were not fracture toughness tested.

Based on our review of the available fracture toughness data and material fabrication histories, and the use of correlations between metallurgical char-acteristics and material fracture toughness, we conclude that the ferrit'ic components in the La Salle containment pressure boundary meet the fracture toughness requirements that are specified for Class 2 components by the 1977 Addenda of Section III of the American Society of Mechanical Engineers Code.

Compliance with these code requirments provides reasonable assurance that the La Salle reactor containment pressure boundary will behave in a nonbrittle manner, that the probability of rapidly propagating fracture will be minimized, 6-1

and that the requirements of Criterion 51 of the General Design Criteria are satisfied.

6.3 Emergency Core Cooling System 6.3.2 Evaluation 6.3.2.3 Functional Design In our Safety Evaluation Report, we addressed the available net positive suction head for pumps in the emergency core cooling system and indicated that there was adequate margin to prevent cavitation in the La Salle design.

However, it has come to our attention in some recent reviews that the potential may exist for damage to emergency core cooling system pumps from cavitation because of local flashing in the system suction lines.

The flashing potential can result, for example, from local elevatiun changes in the piping runs.

Calculations of net positive suction head available at the pump suction may erroneously assume liquid continuity up to the point of pump suction.

The applicant has indicated in discussions that this concern has been addressed in the La Salle emergency core cooling system design.

By letter dated May 18, 1981 from L. O. DelGeorge (Commonwealth Edision) to A.

Schwencer (NRC), the applicant forwarded a calculation of net positive suction head available at the most limiting point.

These calculations considered the piping elbow with the highest elevation, located in the greatest distance from the suppression pool and in the emergency core cooling sytem line with the longest equivalent length.

Pump flow was assumed to be at its maximum, and the suppression pool water was assumed to be saturated at 212 degrees Fahrenheit and at its minimum level.

The calculated margin was 12.1 feet of head.

This is acceptable to us.

Scram Discharge System Pipe Break Since issuance of our Safety Evaluation Report, concerns have been rai.ed on a generic basis for boiling water reactors with regard to the quality of the scram disf.1arge volume piping, the ability to detect and isolate breaks in this system, the potential to detect and isolate breaks in this system, and steam degradation of available emergency core cooling system equipment as a result of a break in this system. We notified the applicant by letter dated April 16, 1981 from R.L. Tedesco (NRC) to J.S. Abel (Commonwealth Edison) that these concerns must be addressed prior to the issuance of an operating license for La Salle. We will condition the operating license of La Salle regarding this action.

6.3.4 Performance Evaluation In our Safety Evaluation Report, we discussed the low pressure coolant diversion, and we noted the conditional acceptability of the La Salle analysis supporting diversion of low pressure coolant injection flow to the wetwell spray.

The La Salle analysis is based on operator action diverting flow to the wetwell exactly 10 minutes into the event.

In our review of some later boiling water reactor operating license applications, we have found that not all units include provisions for preventing flow diversion prior to 10 minutes by other than 6-2

administrative means.

Our concern is that an operator action to divert flow to wetwell spray prior to 10 minutes could result in a violation of the 2200 degrees Fahrenheit peak cladding temperature required by 10 CFR 50.46.

Accordingly, we require the applicant to provide one of the following:

(1) an analysis demonstrating that the low pressure coolant injection diversion at any time prior to 10 minutes will not result in a peak cladding temperature greater than 2200 degrees Fahrenheit, or (2) hardware or logic modifications which preclude diversion prior to 10 minutes, or (3) other justification as to why the current design should be accepted.

In a letter _ dated May 19, 1981 from L. O. DelGeorge (Commonwealth Edison) to A. Schwencer (NRC), the applicant assumed the position that the La Salle County emergency procedures contain adequate cautions to deter the operator from pre-mature flow diversion.

These procedures, which have been approved by us (see Item I.C.1 of Section 22 of our Safety Evaluation Report), caution the operator against diversion unless " adequate core cooling is assured." Low pressure coolant injection diversion is identified as secondary to core cooling require-ments except in those instances outside the design envelope involving multiple failures, for which maintenance of containment integrity is required to minimize risk to the environment.

In light of the applicant's position, we have reviewed the containment response analyses for the design basis event to determine the need for low pressure coolantinjectiondiversion.

These analyses indicate that there should be no need for wetwell spray actuation in the time frame during which the peak cladding temperature is reached.

The operator's focus should, therefore, be on main-taining core cooling.

Based on these analyses and the emergency procedures discussed above, we find the applicants position on low pressure coolant injection diversion to be acceptable.

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7 INSTRUMENTATION AND CONTROL 7.7 Control Systems Not Required For Safety 7.7.3 Specific Findings 7.7.3.4 Additional Concerns Since the issuance of the Safety Evaluation Report, two additional specific concerns materialized as a result of the ongoing review of operating license applications. The specific concerns are delineated as follows.

(1) Common Electrical Poyer Sources or Sensor Malfunctions May Cause Multiple Control System Faiiures With regard to the effects of control system failures or malfLnctions, the analyses reported in Chapter 15 of the Final Safety Analysis Report are intended to demonstrate the adequacy of safety systems in mitigating anticipated operational occurrences and accidents, including those related to control systems.

Based on the conservative assumptions made in defining these " design bases" events and the detailed review of the analyses by us, it is likely that they adequately bound the consequences of single control system failures. To provide assurance that the Chapter 15 analyses adequately bound events initiated by a single credible failure or malfunc-tion, we require that a review be conducted to identify any power sources or sensors which provide power or signals to two c. more control systems, and to demonstrate that failures or malfunctions of these power sources or sensors will not result in consequences outside the bounds of the Chapter 15 analyses or beyond the capability of operations or safety systems.

(2) High Energy Line Breaks and Consequential Control System Failures We have also requested a review be conducted to determine whether the harsh environments associated with high energy line breaks might cause control system malfunctions and result in consequences more severe than

' hose of Chapter 15 analyses or beyond the capability of operations or safety systems.

We require that the applicant resolve these concerns before startup following the first refueling.

Accordingly, we will condition the operating license of Unit No. 1 to reflect this requirement.

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8 ELECTRIC POWER 8.4 Other Electrical Features and Requirements for Safety 8.4.6, Compliance with Recommendations of Regulatory Guide 1.75 8.4.6.1 Physical Identification of Electrical Cables In our Safety Evaluation Report, we concluded pending our review of an independent inspection cable routing report that there is reasonable assurance that the Class 1E and associated cables at La Salle are installed in conformance with the separation criteria recommended by Regulatory Guide 1.75, " Physical Indepen-dence of Electric Systems." Subsequently, the applicant submitted a letter dated March 27, 1981, for our review of a cable routing report titled, "Separa-tion of Electrical Equipment Plant Wide Field Audit Procedure," Numbered 4266-02, and dated July 1, 1980. The following is our evaluation of this report.

The subject audit report icentified specific separation criteria consistent with the Final Safety Analysis Report in the areas of separation of redundant Class 1E cables and separation of associated circuits.

The report concluded based on a 10 percent audit that the La Salle cables are installed in accord-ance with the specific separation criteria cited in the Final Safety Analysis Report. We agree with the reports conclusion.

Thus, we can conclude that there is reasonable assurance that separation between redundant Class IE and between redundant associated cables meet the recommendations of Regulatory Guide 1.75, and is acceptable.

The report, however, did not " clearly" identify criteria for separation between Class 1E and non-Class 1E cables or separation between associated cables and i

non-Class 1E cables.

The separation criteria specified in the Final Safety l

Analysis Report similarly are not clearly identified.

The Final Safety Analysis Report criteria include:

l (1). Separation between Class 1E and non-Class IE cables Class 1E components are physically separated froa,non-Class 1E components that could cause loss of redundancy as a result of a design-basis event effecting failure.nf these components.

l (2) Separation between associated and non-Class 1E cables Non-Class 1E cables can be installed in Class 1E raceways.

However, once committed to a Class 1E raceway of one division that cable cannot be run in any raceway of the other divisions nor is it permitted to cross from the Class 1E raceway to a non-Class 1E raceway.

Non-Class IE equipment that share power supplies with Class 1E equipment can have their cables l

installed only in Class 1E raceways assigned to the same divisions as the t

power supply.

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The Final Safety Analysis Report also states that "Each nonsafety-related cable which has any part of its length in a Divsion 1, 2, or 3 tray, or which shares an enclosure with a Class IE circuit, or which is not physically separated from safety-related cables by acceptable distance or barriers is defined as a

' division-associated cable.'"

The audit report permitted the auditor's judgment to be used as the basis for identifying deficiencies in the areas where separation criteria were not clearly defined.

Based on an auditor's judgment, we cannot reach a conclusion of acceptability.

Clearly defined separation criteria are included in Section 4.6 of the Institute of Electrical and Electronics Engineers Standard 384-1974, "IEEE Trial-Use Standard Criteria for Separation of Class 1E Equipment and Circuits." This separation is defined to be 1 foot between trays separated horizontally and 3 feet between trays separated vertically for the cable spreading area includ-ing entrances to control room panels and 3 feet between trcys separated horizontally and 5 feet between trays separated vertically for general plant areas.

Appendix B, page 8.194 of Amendment 55 of the Final Safety Analysis Report states that their decign meets the separation criteria included in the Institute of Electrical and Electronics Engineers Standard 384-1974.

Thus, we conclude that the La Salle design should meet the separation guidelines defined in Section 4.6 of the Institute of Electrical and Electronics Engineers Standard 384-1974 and that the audit report be revised to include specific criteria defined in the Institute of Electrical and Electrical Engineers Standard 384-1974 for separation between Class 1E and non-Class 1E cables.

This item was discussed with the applicant and as a result Appendix B to the Final Safety Analysis Report was revised by letter dated April 30, 1981 to indicate an additional exception to Regulatory Guide 1.75 or Institute of Electrical and Electronics Engineers Standard 384-1974 recommendations.

Appendix B was revised to indicate a separation criteria of 3 inches hori-zontal and one foot vertical for separation between Class IE and non-Class 1E cable trays.

In justification of the adequacy of this separation criteria, the applicant documented primarily the following items:

(1) Where non-Class IE trays com.e in close proximity to Class 1E trays, only one division of Class 1E trays are present in the room, (2) all non-Class IE cables have the same high quality fire retardant charac-teristics as the Class 1E cables, and (3) there are only a minimal number of locations in the auxiliary building where non-Class 1E trays are installed in close proximity to Class 1E trays.

Based on the above justification, we are unable to conclude that a non-Class 1E cable failure (such as a cable fire induced by a short circuit under a design basis event) will not affect Class 1E circuits that are located in cable trays

.or in conduits in proximity to non-Class 1E cable trays.

It is, therefore, our requirement that suitable barriers be installed between non-Class 1E cable 8-2

trays and Class 1E trays or conduits that are separated by less than that separation recommended by the Institute of Electrical and Electronics Engineers Standard 384-1974 or provide justification for the present design.

This item will be pursued with the applicant and the results will be reported in a supplement to this report.

1 I

r l

I l

l 8-3 l

9 AUXILIARY SYSTEMS 9.1 Fuel Storage and Handling In Sections 9.1.2 and 9.1.4 of the Safety Evaluation Report, we concluded that the design for the spent fuel storage and fuel handling facilities conform with the appropriate requirements of Criteria 61 and 62 of the General Design Criteria and with the appropriate positions of Regulatory Guide 1.13, " Spent Fuel Storage Facility Design" and NUREG-0612 (formerly Regulatory Guide 1.104),

" Control of Heavy Loads at Nuclear Power Plants."

In Appendix C of the Safety Evaluation Report, we discussed the Commission's unresolved safety issues including Task Number A-36 " Control of Heavy Loads at Nuclear Power Plants." NUREG-0612 presents the resolution of this matter, and provides guidelines for necessary changes to assure the safe handling of heavy loads once a plant becomes operational.

By means of a generic letter dated December 22, 1980, we requested the applicant to review the design of La Salle against the guidelines of NUREG-0612 and to provide the results of the study to the ERC staff. attached to the December 22, 19f.0 generic letter identified a number of measures dealing with safe load paths, procedures, operator training, and crane inspections, testing and maintenance. We will require all applicants for an operating license to implement these interim actions prior to the final implementation of the NUREG-0612 guidelines and prior to tne receipt of their operating license. We believe that these interim actions will provide ressonable assur-ance of safe handling of heavy loads until NUREG-0612 can be fully implemented and is, therefore, acceptable.

In a letter dated May 15, 1981, the applicant has indicated the implementation of the interim actions.

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13 CONDUCT OF OPERATIONS 13.3 Emergency Planning Since the issuance of our Safety Evaluation Report, the applicant has updated its Emergency Plan. We reviewed the applicant's plans for co, ing with emergencies at the La Salle facility.

These plans consist of the generic Generating Station's Emerger.cy Plan and the La Salle Station Site Specific Annex.

The plans were reviewed against the requiremcr.t: of A,npendix E to 10 CFR Part 50; NUREF-0654, Revision 1, " Criteria for Preparation and Evaluation of Radiologi-cal dmergency Response Plans and Preparedness in Support of Nuclear Power Plants; and 10~CFR Part 50.47(b). Our findings are discussed in Section 22 cf our Safety Evaluation Report and this supplement, in Appendix 0 to our Safety Evaluation Report, and in Appendix E in this supplement.

The previous information presented in Section 13.3 of our Safety Evaluation Report has been superseded because of the more stringent requirments and new rule making in the area of emergency preparedness.

Regulatory Guide 1.101,

" Emergency Planning for Nuclear Power Plants," and Section 13.3 of the Standard Review Plan are superceded.

NUREG-0654, NUREG-0696, " Functional Criteria for Emergency Response," and the new rule are the current requirements which we used to evaluan the applicant's emergency plans.

Accordingly, Items III.A.1.2 and III.A.2 of Section 22 of our Safety Evaluation Report and this supplement addresses our current findings relevant to the applicant'sbrogramofemergencypreparedness.

13.6 Industrial Security The applicant has submitted security plans entitled "La Salle County Station Security Plan," "La Salle County Station Guard Training and Qualification Plan," and "La Salle County Station Safeguaras Contingency Plan," for the l

protection of La Salle against radiological sabotage.

As reported in our Safety Evaluation Report, certain portions of the plans were identified as requiring additional information and upgrading to satisfy the requirements of Section 73.55 and Appendices B and C of 10 CFR Part 73.

In response to this evaluation, the applicant filed revisions to these plans which, when modified by an implementing license condition, are considered to meet the requirements of 10 CFR Part 73 and therefore are acceptable.

An ongoing review of the progress of the implementation of these plans will be performed by the NRC staff to assure conformance with the performance require-ments of 10 CFR Part 73.

This matter is considered to be resolved.

As indicated in the Safety Evaluation Report, the applicant's security plan is being withheld from public disclosure in accordance with Section 2.790(d)(I) of 10 CFR Part 2.

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16 TECHNICAL SPECIFICATIONS We have worked with the applicant and have prepared a draft of the Technical Specifications for La Salle. We are reviewing these draft Technical Specifica-

~

tions and they will be part of the operating license.

On the basis of our review to date, we conclude that normal plant operation within the limits of the Technical Specifications will not result in offsite exposures in excess of the 10 CFR Part 20 limits.

Furthermore, the limiting conditions for operation and surveillance requirements will assure that necessary engineered safety features will be available in the event of malfunctions within the plant.

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+

17 QUALITY ASSURANCE 17.3 Conclusions Our review of the quality assurance program description for the operations phase of the La Salle County Station Unit Nos. 1 and 2, has verified that the criteria of Appendix B to 10 CFR Part 50 have been adequately addressed in Chapter 17 of the Final Safety Analysis Report.

This determination of accept-ability included a review of the list of items to which the quality assurance program applies.

The list of items was reviewed by our technical review branches to assure that safety related items within their scope of review fall under the quality assurance program controls.

Differences between the applicant and ourselves regarding the list have been resolved to our tatisfaction.

The list has been expanded to include safety-related items reflected in NUREG-0737, " Clarification

'of TMI Action Plan Requirements," November 1980.

Therefore, we conclude that the quality assurance program for operation is acceptable.

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18 ' REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS A Subcommittee of the Advisory Committee on Reactor Safeguards (Committee) visited the site for La Salle and considered the application for operating licenses for La Salle County Station, Unit Nos. 1 and 2 on April 3 and 4, 1981.

The full Committee completed its review of the applicant at its 252nd meeting on April 9, 1981.

A copy of_the Committee report dated April 16, 1981 is attached as Appendix B.

A discussion of the current status of each item on which the Committee commented or made recommendations in the report is included in the following paragraphs.

(1) The-Committee noted that the applicant is emphasizing the plant staffing and personnel training.

In addition, the Committee stated that it believes that efforts to improve staff capabilities should continue, particularly in the area of: health physics. With respect to the latter area, in our

~ Safety Evaluation Report we specified two items of concern that the applicant agreed to resolve.

The first item of concern we indicated was the work experience of the Radiation Protection Manager at La Salle.

It is our position that the Rad / Chem Supervisor at La Salle does not have adequate " applied radiation protection" experience for a plant Radiation Protection Manager as specified by Regulatory Guide 1.8, " Personnel Selection and Training." As we addressed in the Safety Evaluation Report, in order to resolve this issue the applicant has agreed to the following program modifications.

The Health Physicist at La Salle, who serves as the Radiation Protection Manager backup, and who is the 'ine manager to the health physics group, meets the requirements of Regulatory Guide 1.8, and will assist the Radiation Protection Manager in the daily conduct of radiation protection matters.

The applicant has appointed a full-time "as low as is reasonably achievable" (ALARA) coordinator to monitor ALARA performance and assist the Radiation Protection Manager.

l The related concern is the applicant's organization of a combined health l

physics and chemistry (HP/ Chem) department at La Salle.

Health Physics appraisal findings from the applicant's other plants having similar HP/ Chem l

structures have shown that weaknesses do exist in this type of joint organization.

In order to resolve these weaknesses and assure proper operation of its radiation protection organization, the applicant has cenissioned a consultant to perform a management assessment of its organizational structure. The applicant committed to implement the recommendations of this study in order to assure the proper functioning of the radiation protection program at La Salle and its other plants.

This study is being performed in two phases.

The first phase of this study has been completed, and the recommendations from this first phase i

has been substantially implemented.

The second phase of this study has resulted in a draft report.

Two changes which have occurred at La Salle as a result of this study are the addition of a full-time ALARA coordinator and a Chemical Foreman to oversee all chemical analysis activity performed by the rad-chem technicians. We will review the findings of this study when submitted by the applicant.

18-1

i As discussed above, the applicant has made several improvements to its Health Physics organization and has committed to make further changes based on the recommendations of the consultant's study to improve their health physics organization. We will continue to review these changes as submitted.

(2) The Committee noted that the NRC staff recently issued an Office of Analysis and Evaluation of Operational Data (AEOD) report, " Safety Concerns Associated with Pipe Breaks in the BWR Scram System," March 1981, con-cerning the risk potential for pipe breaks in the boiling water reactor scram system.

The Committee stated that this issue should be treated generically and need not be resolved prior to operation of La Salle.

With respect to this concern, we had previously notified the applicant by letter dated April 16, 1981 from R. L. Tedesco to J. S. Abel to address the following items:

(1) The quality of the scram discharge volume piping, (2) the ability to detect and isolate breaks in this system, (3) the potential to detect and isolate breaks in this system, and (4) steam degradation of available emergency core cooling system equipment resulting from a break in this system..Also, we specified that these concerns must be addressed prior to the issuance of an operating license for La Salle, and that we would condition the operating license of La Salle.

(3) The Committee also noted that the NRC staff proposes to require the installation of core thermocouples in the La Salle reactor vessel as specified by Regulatory Guide 1.97, Revision 2, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Condition During and Following an Accident"; and the applicant did not agree.

The Committee recommended that a study be made to determine the feasibility of the use of core outlet and core subassembly thermocouples before imposing such a requirement on La Salle.

In Section 22 of this report, Item II.F.2, " Instrumentation for Detection of Insdequate Core Cooling," the NRC staff presents its study of the feasibility of using incore thermocouples for boiling water reactors, and based on this study, its conclusion that, prior to fuel load of L: Salle, we require the applicant to commit to:

(a) incorporate incore thermocouples into the inadequate core cooling monitoring system prior to June 1983 in accordance with Regulatory Guide 1.97; and (b) provide documentation required by Item II.F.2 of NUREG-0737, address-ing the inclusion of thermocouples in the final inadequate core cooling monitoring system on a schedule acceptable to the NRC staff.

l i

18-2

=.

22 THI-2 REQUIREMENTS 22.2 TMI Action Plan Requirements for Applicants for Operating Licenses I Operational Safety I.C Operating Procedures I.C.8 Pilot Monitoring of Selected Emergency Procedures for NTOL Applicants Discussion and Conclusions In our Safety Evaluation Report, we reported that the review of the La Salle emergency operating procedures was complete except for making a few final revisions, including some plant-specific numbers and operator action level graphs, and training the operators on the revised procedures.

In a letter dated May 4, 1981, the applicant forwarded revised draft copies of the appli-cable emergency operating procedures. We have reviewed these procedures to ensure that the action levels incorporated are consistent with the BWR Owners Group guidelines.

In the May 4, 1981 letter, four items was reported to be in progress with final resolution to be made during the third week in May.

These items are the suppression pool heat capacity limit, the minimum reactor pressure vessel pressure which assures sufficient alternate shutdown cooling flow, maximum reactor pressure vessel pressure which assures sufficient alternate shutdown cooling flow, and primary containment pressure limit. Our Inspection and Enforcement Resident Inspector verified that these items are now included in the emergency operating procedures.

However, the calculated plant-specific numbers in these procedures have not been verified by General Electric.

This verification will be completed prior to fuel load.

r Two additional items were identified in the May 4, 1981 letter.

These were l

sequential opening of safety relief valves for core cooling without injection and minimum single safety relief valve pressure rise.

A final method of calculating the plant-specific numbers and operator action levels have not been finalized by the BWR Owners Group.

Sequential opening of safety relief valves to provide steam cooling to the core will be used only in the event of no injection systems being available and the water level in the core has l

dropped to the core midplane.

The safety relief valves are opened drawing l

steam past the fuel rods to provide cooling until an injection system can be operated to restore reactor pressure vessel level.

The applicant has committed l

to opening valves more rapidly to provide more steam flow cooling than is assumed in the BWR Owners Group guidelines until tl* BWR Owners Group provides final resolution of this calcuation.

The operator action level graph not provided is the minimum single safety relief valve pressure rise.

This graph is necessary when safety relief valves are opened and the reactor pressure vessel is flooded until water is passed through the safety relief valves.

One safety relief valve is shut and the resulting pressure rise indicates whether 22-1 t

l

steam or water is being passed through the valve.

These actions will be used only if all reactor pressure vessel level indication is lost, making it impossible for the operator to ensure the core is covered and cooled.

The applicant has committed to direct the operator to continue with reactor pressure vessel flooding until some positive indication, other than safety relief valve pressure rise, indicates that the reactor pressure vessel is flooded, therefore ensuring core coverage.

Both of the proposed actions are considered to be more conservative than required by the BWR Owners Group guidelines.

Therefore, we conclude that these proposed actions are acceptable.

In the May 4, 1981 letter, the applicant reported that training on the revised emergency operating procedures is in progress and retraining on the sections yet to be completed will be performed to the satisfaction of our Inspection and Enforcement Resident Inspector prior to fuel load.

Our Resident Inspector will verify that the operators are trained on the revised emergency operating procedures, and we will condition the operating license of La Salle. We find the actions taken to be acceptable to meet this requirement.

Therefore, we conclude that the La Salle emergency operating procedures are acceptable for operation.

I.D.1 Control Room Design Review Discussion and Conclusion In our Safety Evaluation Report, we indicated that there were some deficiencies in the control room design of La Salle that were required to be corrected prior to fuel load, some before full power operation, and that there were some deficiencies needing further consideration that will be required to be addressed in the applicant's detailed control room design review.

These deficiencies are so categorized in Appendix C to this report.

II Siting and Design II.B.7 Analysis of Hydrogen Control, and II.B.8 Rulemaking Proceeding on Degraded-Core Accidents Discussion and Conclusion In its letter of November 17, 1980 from J. S. Abel to B. J. Youngblood, the applicant stated its intention to inert the primary containments of La Salle.

In Ammendment 55 to the Final Safety Analysis Report, the appiicant addressed the item regarding containment inerting. We have reviewed the information presented in Chapter 9.5.9 of the Final Safety Analysis Report and conclude that the statement made by the applicant fulfills the staff's requirements that evolved from Items II.B.7 and II.B.8 of NUREG-0660, "NRC Action Plan Developed as a Result of the THI-2 Accident" and Commission guidance that the matter of hydrogen control for degraded core accidents in plants with small containments should not be deferred to rulemaking.

22-2

II.E.2 System Design II.E.4.2 Containment Isolation Dependability Discussion and Conclusion In our Safety Evaluation, we stated that as a result of the applicant's commit-ment to inert the primary containment structure, and its intention to use the containment bypass purge system during plant operation, six valves (1VQ027, IVQ043, IVQ030, IVQ042, IVQ047, and IVQ048) which would have been required to be closed have now been classified as " active" valves.

Therefore, we require the applicant to demonstrate that these valves satisfy the operability criteria set forth in Branch Technical Position CSB 6-4 " Containment Purging during Normal Plant Operations," and our Guidelines for demonstration of operability of purge and vent valves (see Table II.E.4.2 of the Safety Evaluation Report).

The applicant has stated in the May 13, 1981 letter from L. O. De1 George to A.

Schwencer that at fuel load the applicant shall satisfy the " Interim Position" in Section II.E.4.2 of NUREG-0737.

The applicant has committed to limit the opening of the purge valves to no more than 50 degrees.

This is an acceptable interim resolution until long term operability is demonstrated on the condition that the applicant provide the basis for the limitation on the valve opening position. We will require that this basis be provided before licensing for operation.

II.F.2 Instrumentation for Detection of Inadequate Core Cooling Discussion and Conclusion In our Safety Evaluation Report, we concluded that the applicant was not in full compliance with Item II.F.2 requirements since incore thermocouples were not addressed.

The applicant, in its letter of May 4, 1981 to A. Schwencer, addressed this issue.

Our review of this information is discussed below.

In its transmittal, the applicant repeated the design description of installed water level instrumentation which was provided in an earlier submittal and recognized in our Safety Evaluation Report.

The applicant has also repeated the results of the BWR Owners Group evaluation provided in NED0-24708A "Addi-tional Information Required for NRC Staff Generic Report on Boiling Water Reactors," December 1980, supporting the adequacy of installed level instrumen-tation as the basis for operator recognition and response to all postulated events whereby the approach or existence of inadequate core cooling conditions would be indicated.

This evaluation concludes that incore thermocouples would provide no meaningful indication of fuel or clad temperature while the core spray is operating and actively cooling the core.

We concur and concluded in our Safety Evaluation Report that the BWR Owners Group analyses to support emergency procedure guidelines based on installed level instrumentation are adequate.

To further explain its position, the applicant's submittal presented an evalua-tion which was provided to the Advisory Committee on Reactor Safeguards as comments on core exit thermocouple requirements for boiling water reactors per Draf t 2 of Revision 2 to Regulatory Guide 1.97, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an 22-3

Accident," by the BWR Owners Group, which the applicant is a member.

(Revision 2 of Regulatory Guide 1.97 was issued in final form in December 1980.) The

" Technical Background" Section of this evaluation indicates a misunderstanding of the purpose for boiling water reactor incore thermocouples and their quantity needed as given in Regulatory Guide 1.97.

The purpose is twofold:

(1) Core Cooling - to provide diverse indication of water level, and (2) Fuel Cladding - to monitor core cooling (only the latter is addressed by the applicant).

The applicant's evaluation presented arguments that thermocouples are not needed to indicate that fuel cladding has been breached s hce other indicators such as radiation monitors would indicate the extent of the resultant core damage.

The need for diverse water level indication and for monitoring core cooling effectiveness with or without breach of the cladding is not addressed.

The evaluation talks about using approximately fifty thermocouples whereas only four thermocouples per core quadrant are specified as being required in Regulatory Guide 1.97.

With this evaluation, the applicant contends that inclusion of a requirement for core exit thermocouples is unreasonable since they are useful only when all core spray systems are nonfunctional.

The applicant notes that inadequate core cooling requirement is based on a nonmechanistic assumption of inadequate core cooling conditions which would only exist in a boiling water reactor when the core spray is fully nonfunctional.

Considering the TMI-2 accident during which the emergency core cooling system effectiveness was defeated by operator actions, we believe that the assumption of no core spray is a rational one, which is consistent with the basis for this requirement to require core exit thermocouples.

The applicant contends that even under conditions of an uncovered core (for which superheated steam would be an indication of fuel,heatup and a low water level), the operator would be taking all appropriate actions to restore water level above the core based only on knowledge that water level is low and no injection has occurred.

We do not require that all inadequate core cooling indicators be used as a trigger for operator response.. Me do expect that the status of all available indicators would be noted by the operators in their assessment of the situation.

Finally, the applicant presents the results of the General Electric analyses l

which indicate that thermocouples located adjacent to a blocked and overheated l

fuel bundle would not measure superheat in the blocked area after core level l

recovery and that the fuel damage would not nropagate due to admittance of bypass flow after melting of the fuel can.

de reviewed these calculations I

during our generic evaluation of the need for boiling water reactor incore l

thermocouples and concluded that the assumptions and models employed were not i

demonstrably valid.

Additional calculat M s in this generic review were l

requested but were never provided.

With respect to the feasibility for thermocouple placement, we also considered this aspect generically during the development of Regulatory Guide 1.97 require-l ment.

Based on meetings with General Electric, we concluded, as does the 1

22-4 l

applicant, that the location of thermocopules within or on the fuel assembly or on the shroud head with leads projecting downward to near the fuel assembly discharge is not practical.

However, it was concluded by the NRC staff and General Electric in those meetings that it would be feasible to locate one thermocouple in each power range monitor assembly using the existing vessel penetrations.

This conflicts with the unsupported statement in the May 4, 1981 applicant's submittal that the addition of thermocouples to power range monitor assemblies is impossible.

Until the applicant can provide information to support this claim, we have no basis to accept this statement.

Based on the available information, including a generic evaluation by General Electric while we were establishing the feasibility of incore thermocouples, we believe that the incorporation of one thermocouple into each of four power range monitor assemblies at various elevations in each core quadrant is feasible.

We had independent calculations performed by two different organizations to evaluate the feasibility of using thermocouples located in the power range monitor assemblies to detect inadequate core cooling, based on no core spray.

These two organizations were the Pacific Northwest Laboratories and our Office of Nuclear Regulatory Research.

The Pacific Northwest Laboratories (our consultant) performed an analysis on the response of thermocouples located in the incore monitoring tubes to a temperature excursion inside a boiling water reactor bundle.

They used their ESSOR computer program (which moueis an instantaneous boil-off and 36 rods) and modified the exterior model of the program to depict the boiling water reactor bundle.

Our consultdnt assumed the power to be two percent of normal operation per rod.

This analysis indicates that:

(1) the lag time in registering an excursion is on the order of 1 to 1 minutes; (2) the fuel rod temperature is sensitive to the normalized water level (ratio of mixture level to core height).

The maximum central rod temper-ature would change from approximately 1600 degrees Kelvin for a normalized water level of 0.48 (1.0 represents 'op of the core) to a temperature of greater than 3000 degrees Kelvin fo.

normalized water level of 0.46 (see Figure II.F.2-1); and (3) the temperature difference between the central rod and the measuring thermocouples depends on the temperature rise rate, but temperature difference values of 400 degrees Kelvin would not be unreasonable for a clad temperature rise rate of 2 to 3 degrees Kelvin per second.

Our office of Nuclear Regulatory Research has performed an analysis calculating the time dependence of the temperature rise of typical fuel rods during boildown i

of water in a boiling water reactor core, and the corresponding indication of l

a thermocouple located at several elevations in the instrument guide tube outside the fuel channel box.

At first, the calculation was made to obtain l

the mixture level within the bundles, as a function of time, assuming:

(1) uniform power generation (average), and (2) no inflow occurs at core inlet.

l Time is set equal to zero when the dropping mixture level just starts to l

unce.:r the top of the core.

Secondly, the calculation was made to obtain radiation heat transfer from the fuel rod to the cannister wall and from the l

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cannister wall to the tube (outside the cannister box) containing the thermo-couple, for a location just above the mixture level.

In order to make the problem amenable to hand calculations, assumptions were made as follows:

(1) the time interval between local uncover and the time when the adjacent, or higher-up thermocouples see temperature change is small.

Due to this assumption, the local heat generation can be regarded as a constant, whose magnitude depends on the local (spatial) power factor and on the instantaneous heat generation rate; i.e., 2 percent, 1 percent, etc., of full power; (2) heat transfer by convection to steam is neglected; (3) the fuel rod heats up adiabatically, i.e., the radiation heat loss to the cannister wall is neglected in calculating the fuel clad temperature.

Similarly, in calculating the cannister box temperature, heat radiated to the thimble containing the thermocouple is also neglected.

The results of the calculation are provided in Figure II.F.2.2, and confirm the conclusions of the Pacific Northwest Laboratories' analysis.

Figure II.F.2.2 shows that:

(1) for the thermocouple at the 80 percent elevation and the core at 2 percent power, the foam level will clear that location in about 90 seconds after it clears the core top.

At that time the fuel tempera-ture will start to rise adiabatically above saturation temperature.

By the time the thimble thermocopule can see a 25 degrees Centigrade rise (above saturation temperature), the thermocouple temperature measurement lag, the fuel temperature would have risen to about 790 degrees Kelvin, and (2) for the thermocouple at the 60 percent elevation and the core at 2 percent power, the foam level will clear that location in about 210 seconds after it clears the core top.

By the time the thermo-couple at the 60 percent elevation sees a 25 degree Centigrade temperature rise (above saturation temperature), the fuel temper-ature at the 60 percent elevation would have reached 830 degrees Kelvin and at the 80 percent elevation the fuel would be at 950 degrees Kelvin.

These calculations show that the core exit thermocouples will function indepen-dent of location in the incore instrument assemblies, and can provide opera-tionally meaningful data (corresponding to fuel conditions) with respect to inadequate core cooling when there are no emergency core cooling system core sprays.

Therefore, we have reviewed the applicant's positions taken with respect to our boiling water reactor core thermocouple requirements. We find that the response, with one exception, does not present any information which was not previously considered by us in establishing our boiling water reactor core thermocouple requirements.

The exception, concerning the nonfeasibility of placing taermocouples in the power range monitor assemblies, is unsupported.

Unless we receive new information which would modify our conclusions, our 22-7

i FUEL ROD i

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TIME WHEN LEVEL POWER JUST CLEARS TOP OF CORE Figure II.F.2-2 Incore Thermocouple Temperature Response 22-8

position is that prior to issuance of an operating license, we require the applicant to commit to:

(1) incorporate incore thermocouples into the inadequate core cooling monitor-ing system prior to June 1983 in accordance with Regulatory Guide 1.97; and (2) provide documentation required by Item II.F.2 of NUREG-0737, " Clarification of TMI Action Plan Requirements," addressing the inclusion of thermocouples in the final inadequate core cooling monitoring system, on a schedule acceptable to us.

We will ccndition ',he operating license of La Salle to reflect (1) and (2) above.

I I. K..; Final Recommendations of Bulletins and Orders Task Force Item 18 Modification of Automatic Depressurization System Logic--Feasibility for Increased Diversity for Some Event Sequences Discussion and Conclusions In our Safety Evaluation Report (NUREG-0519), we indicated that the applicant is a participant in the BWR Owners Group Program to study potential automatic depressurization system logic modifications.

The BWR Owners Group study was submitted on April 1, 1981. The applicant, by letters dated May 4, 1981 and May 6, 1981 from L. De1 George (Commonwealth Edison) to A. Schwencer (NRC), has endorsed the study results as applicable to La Salle and has committed to implement modifications recommended in the study.

The logic will be modified by either:

(1) eliminating the high drywell pressure trip or (2) bypassing the high drywell pressure trip after runout of a timer started at the low pressure emergency core cooling system initiation level.

In the event that the second alternative is chosen, the applicant has committed to demonstrate by analysis the appropriate time delay setting for La Salle and submit the results of the analysis with the logic modification package.

Either of these modification; will provide for automatic depressurization for the events of interest and either is acceptable to us.

The applicant has committed to l

completing logic modifications prior to the start of the second fuel cycle for l

Unit 1 and before fuel loading for Unit 7 In the interim, should rapid l

Vessel depressurization be required due to a break outside containment or a stuck open relief valve, manual actuation of the automatic depressurization l

system can be accomplished.

The logic modification implementation schedule is l

therefore acceptable.

Item 44 Evaluation of Anticipated Transients with single Failure to Verify No Fuel Failure Discussion and Conclusions l

In our Safety Evaluation Report (NUREG-0519), we noted that the applicant had l

committed to providing an updated response based upon a study performed for the BWR Owners Group. We also required that the applicant provide a discussion of the operator actions assumed in his analyses.

l 22-9

By letters dated May 4, 1981 and May 6, 1981 from L. DelGeorge (Commonwealth Edison) to A. Schwencer (NRC), the applicant endorsed as applicable to La Salle the results of the BWR Owners Group study in this area.

The BWR Owners Group report was submitted to us by letter dated December 29, 1980 from D. B. Waters (BWR Owners Group) to D. G. Eisenhut (NRC).

The evaluation states that the worst case transient-with-single-failure combination for BWR/5 plants is the loss of feedwater event with failure of the high pressure core spray system.

A stuck open relief valve was also considered in addition to +he high pressure core spray failure.

The results of these studies indicate that the core remains covered during the whole course of the transient either due to reactor core isolation cooling system operation, or automatic or manual depressurization permitting low pressure inventory makeup.

The operator action assumed in the analysis is to assure that the automatic depressurization system functions, if needed, or to manually depressurize the vessel to permit low pressure injection.

Based on the results of the BWR Owners Group study and its applicability to La Salle, we find the applicant's response acceptable for this item.

III.

Emergency Preparations and Radiation Protection III.A NRC and Licensee Preparedness III.A.1.2 Upgrade Emergency Support Facilities Discussion and Conclusion The applicant submitted a response to our position relevent to Emergency Response Facilities (NUREG-Co96 " Functional Criteria for Emergency Response")

by letter dated April 10, 1081.

The applicant indicated that the conceptual design description and documentation of the Emergency Response Facilities will be forthcoming in June 1981.

In a letter dated June 1,1981, the applicant submitted its conceptual design description of the Emergency Response Facilities.

The applicant provided the following schedule for completion of the Emergency Response Facilities.

Safety Parameter Technical Emergency Display Support Operations System Center Facility Parameters La Salle Unit 1 October 1, 1982 October 1, 1981 October 1, 1982 1982 La Salle Unit 2 Will be addressed in our continued review.

The applicant intends to locate the Emergency Operations Facility in accordance with Option 2 of NUREG-0696. The Emergency Operating Facility, which will function as part of the emergency facility for the applicant's Braidwood, Dresden and La Salle plants, is being considered to be located in Mazon, Illinois. This is approximately 11 to 15 miles from any of the above facilities.

22-10

.The applicant believes the permanent Technical Support Center will be functional by fuel load except for full operation of the safety parameter display system and other parameters, such as meteorology and real-time radiological information.

~These will be functional by October 1982 or sooner. With respect to Regulatory Guide 1.97 (" Instrumentation for Light Water Cooled Nuclear Power Plants to

' Assess Plant and Environs Conditions During and Following an Accident") para-meters, a schedule will be provided when preliminary scoping is completed.

The implementation of these parameters as called for by Regulatory Guide 1.97 is June 1983, except for.those items as required by NUREG 0787 which were further discussed in Section III.A.1.2 of our Safety Evaluation Report.

This satisfies our current requirement for this Item.

Our Office of Inspection and Enforcement will inspect the construction and operational testing of these facilities. We will condition the operating license of La Salle consistent with the above schedules.

'III.A.2 Improving Licensee Emergency Preparedness - Long Term Discussion'and Conciision Since the' issuance of our Safety Evaluation Report, the applicant has submitted an upgraded Emergency Plan by letter dated March 27, 1981.

This Plan dated April 1981, Revision 2, was raviewed by us. 'Our evaluation of this Plan is provided in Appendix E to this' supplement.

The staff received a letter Jated May 13, 1981 from the Federal Emergency Management Agency setting fort'1 the status of the plans and preparedness for the La Salle site and includes post exercise evaluation of the full scale joint exercises with the Commonwealth Edison Company, the State of Illinois and various counties.

These exercises were conducted at both the Dresden and La Salle sites on October 28, 1980, and December 4, 1980, respectively.

Grundy County was tested in both exercises because it is included in the principal 10 mile emergency planning zone for both the La Salle and Dresden facilities.

The findings.of the Federal Emergency Management Agency state that Grundy County, having performed poorly in both the Dresden and La Salle exercises must be judged as not having a capability to provide protection for the public, and substantial improvements are needed to meet a level of adequacy.

Based on the above,'we conclude that the Grundy County deficiencies identified by the Federal Emergency Management Agency must be corrected prior to liceasing of La Salle, and we will report on the resolution of this issue in a sup'plement

~

to this report. The May 13, 1981 letter from the Federal Emergency Management

-Agency is provided in Appendix F to this supplement.

We have reviewed the applicant's plans for upgrading the meteorological program in accordance with the provisions and schedule set forth in Appendix 2 of NUREG-0654, " Criteria for Preparation and Evaluation of Radiological Emergency Response Plans'and Preparedness in Support of Nuclear Pner Plants." This information was submitted to us in July 1980 and April 1981.

Additional areas of clarification were identified as a respit of that review and the applicant was requested to provide us with additional information. These areas concern

long term improvements in instrument location, siting and height criteria as 22-11 m

1 described in Appendix 2 of NUREG-0654 and proposed Revision 1 of Regulatory Guide 1.23, "Onsite Meteorological Programs." further, analog recorders of meteorological information in the control room need to be installed, as well as a backup meteorological capability.

These issues are long-term improvements and are not required to be operational prior to fuel load.

The applicant currently has adequate compensatory measures for meteorological measurements.

The. applicant's June 1981 submittal will be evaluated by us and a supplement to this report will be provided.

We conducted an extensive two week emergency Preparedness Implementation Appraisal of the La Salle facility during the month of April 1981. This appraisal is in accordance with 10 CFR 50.47(a)(2) which, in part, states that the staff must find that the applicant onsite emergency plans are adequate and capable of being implemented.

As a result of that review, several significant findings relevent to the applicant's capability to adequately implement its plan were identified.

These findings, rela ng mostly to training and proce-i dures and documented in IE Inspection Reporc hos. 50-373/81-14; 50-374/81-09, must be corrected prior to the licensing of La Salle. We will condition the operating license for La Salle to address these items.

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APPENDIX A CONTINUATION OF CHRONOLOGY FOR THE LA SALLE COUNTY STATION February 19, 1981 Letter from applicant transmitting the 1980 Annual Report for Commonwealth Edison Company.

r February 20, 1981 Letter to applicant (Generic Letter 6.-11) concerning NUREG-0619.

February 20, 1981 Letter from applicant concerning resolution of safety evaluation report issues.

February 23, 1981 Letter from applicant cor.cerning Resolution of Reactor Systems Branch Questions.

February 25, 1981 Letter to applicant (Generic Letter 81-04) concerning emergency procedures and training for station blackout events.

February 25, 1981 Letter from applicant concerning Resolution of Open Issues.

February 26, 1981 Letter to applicant transmitting NUREG-0313, Revision 1,

" Technical Report on Material Selection and Processing Guidelines from BWR Coolant Pressure Boundary Piping."

(Generic Letter 81-03).

February 26, 1981 Letter to applicant concerning Periodic Updating of FSARs (Generic Letter 81-06).

March 2, 1981 Letter to applicant concerning Seismic Category I Masonry Walls.

March 2, 1981 Letter from applicant transmitting Revision 9 of the La Salle County Station Security Plan.

March 5, 1981 Letter to applicant concerning Functional Criteria for Emergency Response Facilities (NUREG-0696) (Generic Letter No. 81-17).

March 9, 1981 Letter from applicant concerning Environmental Qualification of Safety-Related Equipment in Harsh Environments.

March 10, 1981 Letter to applicant concerning Environmental Qualification of IE Equipment (Generic Letter No. 81-15).

March 11, 1981 Letter from applicant concerning Preoperational Fog and Ice Observation Program.

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+

March'13, 1981 Letter to applicant concerning accelerated schedules for NTOL plants.

March'13, 1981 Letter to applicant transmitting 20 copies of the Safety Evaluation for OL application.

March 19, 1981 Letter from applicant concerning Supplemental Response Regarding Seismic Qualification Review.

March 19, 1981

' Letter from applicant transmitting Amendment No. 55 to the FSAR.

March 26, 1981 Letter to applicant concerning Issuance of NUREG-0487, Supplement 2, " Mark II Containment Lead Plant Program Load Evaluation and Acceptance Criteria."

March 27, 1981 Letter to applicant concerning Request for Further Revisions to La Salle's Security Plan.

March 30, 1981 Letter to applicant concerning BWR Scram Discharge System; Clarification of Diverse Instrumentation Requirement (Generic Letter 81-18).

April 7, 1981 Letter from applicant concerning Resolution of SER Open Item 2 Small Pipe Vibration Monitoring.

April 13, 1981 Letter to applicant concerning a visit of Los Alamos

. National' Laboratory Engineers to La Salle Station Site.

April 16,-1981 Letter from applicant concerning Initial Test Program -

Special Test Simulated Loss of AC Power.

April-16, 1981 Letter to applicant concerning Safety Concerns Associated with Pipe Breaks in the BWR Scram System.

April 22, 1981' Letter to applicant transmitting a copy of a letter to Chairman Hendrie from the Chairman, ACRS concerning the

" Report on La Salle County Station, Units 1 and 2."

-April 24, 1981 Letter from applicant concerning Thermal Recombiner tifectiveness.

April-24, 1981 Letter to applicant requesting additional information -

Instrumentation and Control Systems Branch.

April ~24, 1981 Letter from applicant concerning Seismic Category I Masonry Walls.

April 30, 1981 Letter from applicant concerning Separation Audit of Electrical Equipment - NUREG-0519 Section 1.8 Item 8.

May 1, 1981 Letter from applicant concerning Supplemental Information in Response to NUREG-0519 Open Items.

A-2

May 4, 1981 Letter to applicant concerning Qur.lification of Inspection, Examination, and Testing and Audit Personnel (Generic Letter 81-01).

May 5, 1981 Letter from applicant concerning the fuel load schedule.

May 5, 1981 Letter to applicant concerning Engineering Evaluation of the H. B. Robinson Reactor Coolant System Leak on January 29, 1981 (Generic Letter 81-22).

May 6, 1981 Letter from applicant concerning additional information in response to NUREG-0519.

May 11, 1981 Letter to applicant concerning postponement of preoperational vibration testing.

May 11, 1981 Letter to applicant concerning special test for simulated loss of alternating current power initial test program.

May 15, 1981 Letter from applicant concerning control of heavy loads.

May 18, 1981 Letter to applicant concerning schedule delay.

May 18, 1981 Letter from applicant concerning NRC question regarding flashing in ECCS suction lines.

May 19, 1981 Letter from applicant concerning NRC questions concerning S/RV performance review program and LPCI flow diversion.

May 21, 1981 Letter from applicant concerning NRC questions on S/RV 9

performance review program.

May 21, 1981 Letter from applicant concerning NRC questions related to fire protection.

May 21, 1981 Letter from applicant concerning Supplemental Information responding to Open Item 5 (NUREG-0519, Section 1.9).

May 29, 1981 Letter to applicant concerning Security Plan Review request for Management Appeal.

June 4, 1981 Letter from applicant concerning Environmental Qualification of Safety-Related Equipment in Harsh Environment - Supple-mental Report.

June 4, 1981 Letter from applicant concerning Seismic Qualification Review (Volume 7) - Impedance Test Report.

June 8, 1981 Letter from applicant concerning Fuel Lift Analysis.

June 10, 1981 Letter from applicant concerning Technical Specifications.

June 10, 1981 Letter from applicant concerning Q-List.

A-3

i' APPENDIX'B LETTER FROM THE ADVISORY COPMITTEE ON REACTOR SAFEGUARDS 4

B-1

I SA AfG

/

o UNITED STATES

~,

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t NUCLEAR REGULATORY COMMISSION

.E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS o

wassmoron, o. c. 2 osse April 16,_1981 The Honorable Joseph M. Hendrie Chairman U. S. Nuclear Regulatory Commission Washington, DC 20555

SUBJECT:

REPORT ON LA SALLE COUNTY STATION UNITS 1 AND 2

Dear Dr. Hendrie:

During its 252nd meeting, the ACRS completed its review of the application of the Commonwealth Edison Company (Applicant) for a license to operate the La Salle County Station Units 1 and 2.

A subcommittee meeting was held in Morris, Illinois on April 3-4, 1981 to consider this project. A tour of the facility was made by members of the Subcommittee on April 3, 1981. During its review, the Connittee had the benefit of discussions with representa-tives of the Applic.nt and the NRC Staff. The Cornittee also had the bene-fit of the documents listed. The Committee reported on the construction permit application for this plant in a letter to AEC Chairman James R.

Schlesinger dated December 17, 1971.

The La Salle County plant is located in La Salle County, Illinois about 70 infles southwest of downtown Chicago. The nearest population center is Ottawa, Illinois about 11 miles northwest of the site.

The La Salle plant uses GE BWR-5 nuclear steam supply systems with a rated power level of 3323 MW(t) each. The La Salle plant has a Mark II pressure suppression containment with a design pressure of 45 psig. The La Salle plant is one of three plants included in the Mark II Owners Group lead plant program. The NRC Staff has concluded review of the lead plant program and has issued Supplements 1 and 2 to NUREG-0487, " Mark II Containment Lead Plant Program Load Evaluation and Acceptance Criteria," which specify generic acceptance criteria. The Staff has concluded that the La Salle facility satisfies the criteria. We concur in this finding.

l The Applicant described the organization of the plant staff, including main-i tenance, engineering, operations, and health physics personnel. The safety review functions and training programs were also discussed. The Applicant is emphasizing plant staffing and personnel training. The Committee believes that efforts to improve staff capabilities should continue, particularly in j

the area of health physics.

I The Applicant and the Staff have under consideration a recently issued AE00 i

report (Reference 4) concerning the risk potential for pipe breaks in the l

BWR scram system. This report is being reviewed by, the NRC Staff and by a i

BWR Owners Group. We believe that this issue should be treated generically

[-

and need not be resolved prior to operation of the La Salle plant.

l B-2

Honorable Joseph M. Hendrie April 16, 1981 The NRC Staff proposes to require the installation of core thermocuuples in the La Salle plant as specified by Regulatory Guide 1.97, Revision 2, "In-strumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident." The Applicant has not yet agreed to this requirement.

In a letter to Comissioner Gilinsky dated July 16, 1980, the Comittee recomended that careful examination of the feasibility of the use of core outlet or core subassembly thermocouples, and the pros and cons of such use, be undertaken. We recommend that such a study be completed for the La Salle plant before a decision is reached on this requirement.

The Committee wishes to be kept informed.

The NRC has identified a number of additional outstanding issues. We be-lieve that these can be resolved in a manner acceptable to the NRC Staff.

The ACRS believes that if due consideration is given to the recommenda-tions above, and subject to satisfactory completion of construction, staff-ing, and preoperational testing, there is reasonable assurance that La Salle County Station Units 1 and 2 can be operated at power levels up to 3323 MW(t) each without undue risk to the health and safety of the public.

Sincerely, J. Carson Mark Chai rman

References:

1.

Commonwealth Edison Company "La Salle County Station Final Safety Analysis Report," Volumes 1-12 and Amendments 1-55.

2.

U.S. Nuclear Regulatory Commission " Safety Evaluation Report Related to the

. Operation of La Salle County Station Units 1 and 2," USNRC Report NUREG-0519, dated March 1981.

3.

U.S. Nuclear Regulatory Commission, Supplements 1 and 2 to NUREG-0487,

" Mark II Containment Lead Plant Program Load Evaluation and Acceptance Criteria," dated September 1980 and February 1981.

4.

U.S. Nuclear Regulatory Commission " Safety Concerns Associated With Pipe Breaks in the BWR Scram System," Office for Analysis and Evaluation of Operational Data, March 1981.

B-3

APPENDIX C HUMAN FACTORS ENGINEERING BRANCH CONTROL ROOM DESIGN DEFICIENCIES A.

Deficiencies to be corrected prior to fuel loading 1.

Annunciators and Alarms a.

Approximately 92 individual annunciator tiles will be rearranged to improve organization of annunciator panels.

b.

Annunciator tiles with legends using small character size and weight will be replaced.

c.

Audible alarm signal intensiti'tes will be adjusted to provide equal detection levels for all audible alarms.

d.

Reactor control panel aduio alarm frequency will be lowered to a level well below 5000 Hz.

e.

Annunciator tile reflash capability will be demonstrated.

2.

Controls a.

Demarcation will be applied to control panels to promote visual recognition of closely-spaced J-handle controls.

b.

Guardrails will be installed on all benchboards and consoles.

c.

Demonstrate clockwise direction of Feedwater Turbine control to increase turbine speed.

d.

J-handles for pumps, open-close valves and throttleable valves will be differentiated by color, shape and texture coding.

e.

Institute administrative procedures to control removal and replacement of interchangeable control indicator light covers.

f.

Demonstrate availability of tools for replacement of indicator lamps.

g.

Demonstrate acailability and use of administrative procedures to assess indicator light status, failure or degradation.

h.

Provide positive visual indication of armed / disarmed statu's of pushbutton trip switches.

C-1

3.

Displays a.

Provide ' normal range' banding on vertical indicators located just below annunciator panels.

Banding shall be permanent (i.e., on meter face) whereever possible.

b.

Relabel meter 'stringings' of 5 or more meters to aid in meter identification.

c.

Identify those meters which have a ' normal' operating range, and provide temporary range bands.

Range bands shall be made permanent where possible, and all applicable meters shall have permanent ' normal' range bands 18 months after full power operation, d.

Provide' demarcation to organize and identify related meters on electrical panel 1PM01J.

e.

Replace all hand-made meter scales with permanent scales.

f.

Provide demarcation lines to improve the visual association of legend / indicator lights with their associated controls.

g.

Replace blue indicator lamp covers for those indicators where luminance contrast is too low.

4.-

Control / Display Relationships Demonstrate that valve position indications are based on posi-2.

tive indication of valve status.

b.

Identify isolation valve pressure displays as inboard or

outboard, c.

I'slable HPCS pump displays.

d.

Provide demarcation lines and possible color coding to associate the HPCS current meter with the HPCS pump controls.

e.

Provide identical Penetration Pressure B/C Duct and B/C Room temperature meters.

f.

Relocate labels that are partially obscured by indicator light covers at base of panel 1H13P601.

g.

Provide RHR system white indicator lights with labels / legends that indicate system status, h.

Provide label for LPCS pump cooler valve.

i.

Demonstrate human factors enhancement and appropriate control /

display relationships for Containment Isolation and Leak Detec-tion panel / console IPM13J and 1PM16J.

C-2

.~ - -

1 s-r j.

Provide the following Standby _ Gas Treatment panel corrections:

i.

Provide Unit 1/ Unit 2 demarcation.

ii.

Label strip chart recorders and install correct chart paper.

iii.

Provide missing indicator light covers.

iv.

Complete panel mimics.

.v.

Correct inconsistent annunciator legend.

k.

Provide Fee'dwater and Condensate Panel 1PM03J demarcation and color coding to enhance system / subsystem discrimination, visually segregate unrelated control / display components, improve control /

display relationships, and improve the visual relationship between valve controls and valve position indicators.

1.

Relabel motor-driven feedwater pump.

m.

Provide demarcation and color coding to enhance control / display relations and discrimination for Auxiliary Systems Panel / Console 1PM09J and 1PM10J.

n.

Provide Electrical Control Panel 1PM01J demarcation and color coding to improve control / display relationships.

o.

Correct Turbine Control Panel labeling errors and revise turbine bearing lift pump status indicators to read off-autotrip-on.

p.

Revise Reactor Water Cleanup and Recirculation Panel 1H13P602 filter demineralizer status light sequence to correspond to vertical panel indicator sequence, and' provide missing labels for this panel.

l q.

Provide Reactor Control Panel range swit6h A label.

r.

Provide HVAC Console and Panel 1PM05J and IPM06J with color coding had shading to associate related controls and displays, and provide demarcation to discriminate specific HVAC subsystems.

5.

Labels a.

Increase size of system or major function labels for all panels, b.

Replace all temporary labels with permanent labels.

Pnovide and demonstrate an administrative procedure that will control the use of temporary labels.

c.

Provide demarcation lines to improve visual discrimination of auxiliary system indicator lights (10x11 matrix).

C-3

d.

Replace labels having white legends on red backgrounds with labels having black legends on white backgrounds.

e.

Demonstrate accomplishment of the labeling revisions specified in the Preliminary Assessment Report, items Id 2-5, and 3i 1-6.

l 6.

Recorders-a.

Demonstrate record operability / maintenance.

b.

Establish consistent systems for dual-recorder upper / lower pen colors, and provide appropriate labels for each recorder that will be consistent with pen location and color coding.

c.

Demonstrate that recorder scales and chart paper scales correspond.

d.

Modify SRV temperature recorder (Panel 1H13P614) to eliminate ambiguity as to which channel is being displayed.

7.

Workspace, Layout and Environment a.

Relocate general procedures, annunciator procedures and emergency procedures at the Unit 1 operating desk.

Provide color coding to identify procedures.

Demonstrate rapid and uninhibited selection of appropriate procedures by the operators.

b.

Demonstrate computer capability to printout historical meteor-ological data on demand.

c.

Replace white indicator lights with white backlight legend lights.

d.

Replace black emergency trip pushbuttons with red pushbuttons.

e.

Demonstrate installation and accuracy of new mimics and demar-cations as described in the Preliminary Assessment Report.

f.

Install additional control room lighting to increase lighting levels on panels and consoles.

g.

Install new covers for blue status lights where luminance contrast is too low.

h.

Install modified ceiling grids to reduce glare on panels and consoles.

Note:

time period for completion of this item may be extended to accommodate vendor delivery schedules.

i.

Reduce sound levels of audible alarms to a maximum 65dbA levels.

j.

Improve visual recognition of phone jacks on control room back panels.

C-4

, ~. _ _ _. _.... - -. _

- = -

k.

Provide index and location labels for tne sound powered phone patch panel.

1.

Organize communications equipment located at the center desk.

m.

Provide accessible, designated storage for operator protective equipment, and demonstrate that equipment is installed and available for use by operators.

Demonstrate availability of adequate communication systems for use with operator protective equipment. Demonstrate availability and adequate storage of individual ~ operator corrective lenses for use with protective equipment.

8.

Remote shutdown panel a.

Provide panel mimics and demarcation.

b.

Demonstrate emergency lighting provisions in the remote shutdown panel area.

9.

Computers a.

Demonstrate trending capability of computer printouts and CRT displays.

Install labeling system for stripchart trend recorders.

b.

Demonstrate resolution of the problem of glare on CRT displays, c.

Modify CRT displays t7 improve CRT readability.

d.

Demonstrate availability of operational procedures covering operator actions in the event of total loss of the process computer system.

e.

Relabel point ID display to be consistent with thumbwheel control.

f.

Reorganize ' point ID index by component number and by system.

B.

Deficiencies to be corrected before full power operation 1.

Annunciators and alarms a.

Improve visual access to annunciator controls by color and shape coding.

C.

Deficiencies to be addressed in the Detailed Control Room Design Review The following items do not represent the entire scope of the detailed control room design report called for in Task I.D.1 of NUREG-0660, but are those deficiencies which have been identified at this time as requiring additional analysis to arrive at an appropriate solution, or which are scheduled for correction after initial full power operation.

l l

C-5

c 1.

Annunciators and Alarms a.

Provide a plan for annunciator tile relocation that locates annunciator tiles above their related systems or controls /

displays, and complete the analysis of annunciator tile organi-zation within individual annunciator panels.

b.

Provide a detailed analysis of annunciator panel arrangement that provides unambiguous distinction between alarms with direct plant safety irrplication and alarms not having a direct effect on plant safety.

c.

Provide a detailed analysis of the use of additional audible alarms (and associated tone / frequency coding) to improve directional cueing and operator location of alarmed panels, and to avoid the possibility of failure to recognize individual alarms during multiple alarm events.

2.

Controls a.

Provide a detailed analysis of the potential operators error that could be caused by the close spacing of adjacent J-handle controls (e.g., panel 1PM03J).

Improve operator visual access to controls by relocating control indicator lights.

Indicator light relocation to be accomplished prior to second startup cycle.

b.

Analyze all J-handle controls to ensure that visual access to legend / identification material obscured by the J-handle is not material that would be required during emergency situations.

Identify and evaluate positive means of determining indicator c.

light failure or degradation.

3.

Displays a.

Review all indicators with either temporary or permanent ' normal' range bands, and ensure that all applicable meters have been identified and that permanent ' normal' range bands have been installed prior to the second startup cycle.

Identify all indicators where the provision of additional range bands (e.g.,

abnormal operation) would improve crerator performance, and provide schedules for installing these range bands.

b.

Provide plans and schedules for reorganization of electrical panel 1PM01J meters.

c.

Provide detailed analyses of means to improve the positive association of legend / indicator lights with their associated controls.

i C-6

4.

Control / display relationships a.

Develop improved Div. I Outboard Isolation valve labeling to minimize the potential confusion resulting from the A-B-C-D/

A-E-J-N relationship as it now exists.

b.

Provide a detailed analysis of the RHR system Loop B Div. II mirror-imaging between Pump A and Pump B.

Justify the arrange-ment as it now exists, or provide plans and schedules for meter relocation.

c.

Provide detailed aralyses for the relocation of Feedwater and Condensate panel 1PM03J controls and displays to enhance subsystem discrimination ar.d improve control / display relat.icn-ships.

Provide plans and schedules for accomplishing relocations.

d.

Analyze Auxiliary Systems panel / console 1PM09J and IPM10J and identify display relocations to improve control / display relation-ships.

Complete relocations prior to second startup.

e.

Analyze Electrical Control panel IPM01J and relocate selected meters to improve control / display relationships.

Complete relocations prior to second startup.

f.

Provide an analysis of the Reactor Centrol Panel 1H13P603 status indicator mirror imaging.

Justify this arrangement or provide plans and schedules for revised sequencing of these indicators.

5.

Labels No requirements.

6.

Recorders a.

Provide an analysis of the potential problems associated with use of dual-scale chart paper.

7.

Workspace, layout and environment a.

Provide a detailed analysis of the duties and functions of the center desk operator, acting as a communications command center controller, during the course of an emergency requiring communi-cations to-and-from the control room.

b.

Provide an analysis and operational plans for Unit 1 or IJnit 2 operator interfaces with the mirror-imaged Unit 1/ Unit 2 consoles and panels, c.

u conjunction with the detailed analysis of annunciator panel rearrangement to distinguish between safety and non-safety alarms, provide an analysis investigating the use of color C-7

(over and above 'first-out' red) to categorize the degree of j

severity or potential safety consequences of individual alarms.

8.-

Remote shutdown par.el No requirements.

1 9.

Computer-a.

Provide an analysis of the apparent violation of design convention with respect to location of the number thumbwheel switch and the function thumbwheel switch.

4 i

L 1

C-8 1

APPENDIX D ERRATA TO THE SAFETY EVALUATION REPORT Page 1-9 Line 29

- Delete "16 Physical Separation Beti:een Division / Associated Cables (8.4.6.4)"

Page 1-9 Line 30

- Change 17 to 16 Page.1-9 Line 31

- Change 18 to 17 Page 2-18 Line 5

- Add A Period After the Word Acceptable and Delete "Except For The Cooling Lake Spillway As Noted Above."

Page 2-18 Line 29

- Change the word Supplies to Suppliers Page 2-23 Line 32

- Change 20 to 10 Page 3-17 Line 10

- Change Loos to Loss Page 3-17 Line 11

- Change Releif to Relief Page 3-22 Line 27

- Change To to In Page 3-26 Line 33

- Chai.ge Applio to Appli-Page 4 Line 38

- Change An to And Page 4-13 Line 4

- Change Blades to Blade Page 5-12 Line 39

- Change III.C.1 to III.C.2 Change IV.A.2.a to IV.A.1 Page 5-13 Line 4

- Change IV.A.2.a to IV.A.1 Page 5-14

'Line 1

- Change IV.A.2.a to IV.A.1

~

Page 5-14 Line 48

- Add the following after Line 48 "(5) When dropweight tests were not conducted for a beltline weld seam, a nil ductility temper-was assumed. greater than -5' 'egrees Fahrenheit ature of no Page 5-15 Line 30

- Change IV.A.2.a to IV.A.1 and 36 Page 5-16 Line 2

- Change IV.A.2.a to IV.A.1 D-1

Page 5-18 Line 18

- Change IV.A.2.a to IV.A.1 Page 5-19 Line 13

- Change IV.A.2.a to IV.A.1 Page 5-19 Line 42

- Change IV.A.2.a to IV.A.1 Page 5-24 Line 35

- Change Two to Three Page 5-24 Line 36

- Change Two to Three Page 5-24 Line 40

- Start paragraph with the following line "The reactor water cleanup system has an inboard and outbaard valves at the containment interface." Change the following sentences to "The outboard isolation valve closes auto-matically on a signal... to the reactor vessel.

This valve also operates if the

... preset level. The design of the out-board and other remote isolation valves is such... manually."

Page 5-25 Line 31

- Delete "ph and chloride concentration" Page 5-25 Line 33

- Change Their to Its Page 5-25 Line 34

- After " control room" add the sentence "The ph and chloride concentration will be sampled daily."

Page 6-31 Line 40

- Delete "a small reactor building fan room" Page 6-55 Line 17

- Delete "a high efficiency particulate air filter" and add "prefilter."

Page 7-2 Line 41

- Change Fu. tional to Function Page 7-4 Line 35

- Delete the word Of Page 7-16 Line 26

- Change 7.7.3 to 7.7.3.2 Page 7-16 Line 43

- Change An to A Page 8-S Line 21

- Change Or to Of Page 8-8

- Change 8-8 to 8-9 Page 8-9

- Change 8-9 to 8-10 Page 8-10

- Change 8-10 to 8-8 Page 8-15 Line 17-

- Delete repeated Sentence "Each battery bank for each unit is located in a separate seismic Category I room in accordance with its electrical divisional assignment."

D-2

Page 8-16 Line 35

- Change "... systems monitoring derive...."

to "... systems monitoring are derived..."

Page 9-1 Line 37

- Change Even to Event Page 9-13 Line 37

- Change 175 to 160 Page 9-14 Line 1

- Change the first line to " power to start the fire pump is supplied by individual pump battery systems."

Page 9-14 Line 20

- Change "...to the sprinklers are electrically

.." to "...to the automatic sprinkler

.system are electrically..."

Page 9-14 Line 31

- After " Turbine building" add "(various areas)"

Page 9-14 Lines 45 & 46 - Delete both lines Page 9-15 Line 6

- Delete " Auxiliary building lower ventilation equipment floor, columns 12 to 8 and L to Rt" Page 9-15 2nd para.,

- Change to "Since the containment is inerted 1st line during operation, standpipe hose stations with adequate hose will be provided outside the containment access openings.

Also an additional hose station has been added in Zone 4A at column 21.1N" Page 9-15 Line 23

- Delete " Zone 4A, column 20L" Page 9-15 Line 34

- Change 50 to 34 Page 9-15 Line 35

- Cnange fourteen to seventeen Page 9-15 Line 43

- Delete the word and and replace by a comma Pagg 9-15 Line 44

- After "(heat sensors)" add "and ultra-violent" Page 9-16 Line 11

- After the word Those add ionization Page 9-17 Line 31

- Delete "and justified" Page 9-17 Lines 33

- Delete "and found acceptable in the fire and 34 protection safety evaluation report." And replace by "by us.

As a result of our evalua-tion, the applicant has agreed to make snodifi-cations identified in Section 9.5.4 of this report."

Page 9-19 Line 20

- Change It's to Its l

D-3

Page 9-20 Line 24

- Delete " Areas that come under these considera-tions include the following:

(1) Fire Zone SCll in the hallway adjacent to the diesel generator rooms.

(2) Fire Zone 48, auxiliary building, lower ventilation equipment floor.

(3) Fire Zone 2E, elevation 761 feet 0 inches of the reactor building.

(4) Fire Zone 4F1 bus duct will be enclosed in a 2-hour fire rated enclosure.

(5) Fire Zone SC11, which has cables from Unit 2 engineering safety features and Division 1 battery room, will have cables re-routed to avoid the turbine building."

and add "In addition to acceptable alterna-tive modifications discussed elsewhere in this Section of Fire Protection, the hall-way adjacent to the diesel generator rooms in Fire Zone SC11 is included in this consideration.

Also, cables from Unit 2 engineering safety features, Division 1 battery room in Fire Zone SC11 will have cables re-routed to avoid the turbine building."

Page 9-21 Line 41

- Change 2 to 3 Page 9-21 Line 43

- After the word single add personnel Page 9-22 Line 1

- Change period to comma after the word aisle add "and a certified 3-hour roll-up fire door will be provided to access the back of panels adjacent to the wall.

In addition, the a'pplicant will provide a liquid-tight sill at this door opening and install a remote pull chain for the overhead door.

The remote pull chain should be located at the single fire door in the center eisle."

Page 9-22 Line 1

- Delete sentence "In addition, an automatic water... the auxiliary electric equipment room."

Page 9-22 Line 17

- Delete "2-hour fire-rated enclosure will be provided to enclose the bus duct in this area."

and replace by "a 2-hour rated fire vapor seal where the bus duct penetrates the floor / ceiling D-4

assembly to the Division 1 essential switch gear room."

Page 9-23 Line 6

- ' Change intakes to intake and add "of the Division 1 diesel generator" Page 9-23 Line 8

- After the word expose add both Page 9-23 Line 9

- Change Mofification to Modification Page 9-23 Line 33

- Change inside to outside Change coolant to recirculation and delete the Page 9-23 Line 42 word and and add by oil and Page 9-23 Line 43

- Change but to and Page 22-5 Line 13 Change 1980 to 1981 Page 22-49 Line 10

- Change Areas to Area Page 22-53 Line'2 After Report add "... as transmitted by 1-letter,"

Page 22-64 Line 23 Change 14 to 24 Page 22-67 Line 44

. Change sentense to read "If valve operability cannot be demonstrated, prior to plant startup, the operator..."

Page 22-73 Line 23

- Delete one Reactor Page 22-83 Line 2

- Change Lease to Least Page 22-84 Line 3

- Change 17Kevto81 Kev Page C-12 Line 8

' - Change Candicates to Candidates Page D-14 Line 25

- Change Interivew to Interview Page D-22 Line 35

- Change Testing to Tested D-5

APPENDIX E SUPPLEMENTAL EMERGENCY PREPAREDNESS EVALUATION REPORT

Introduction The purpose of this appendix is to update Appendix D of our Safety Evaluation Report by providing our evaluation of additional information for those sections of Appendix D where further discussion or changes are in order for the onsite emergency preparedness review.

Each section is supplementary to and not in lieu of the discussion in Appendix 0 to our Safety Evaluation Report.

Each section of this appendix is designated the same as the corresponding section of Appendix D that is being updated.

Since the' issuance of the Safety Evaluation Report, the applicant has upgraded their Emergency plan.

In addition, we conducted an Emergency Preparedness Implementation Appraisal (EPIA) of the La Salle facility during the weeks of April 20 through May 1, 1981.

The EPIA is part of the Office of Inspection

'and Enforcement preoperational program for licensing of this facility.

EPIA findings are documented in IE Report No. 50-373/81-14; 50-374/81-09.

E-1

EVALUATION A.

Assignment of Responsibility (Organization Contro!)

Emergency Plan Since the issuance of our Safety Evaluation Report the applicant has reorganized their offsite Generating Station Emergency Plan (GSEP) organization.

The size of the offsite organization will vary depending upon the nature and extent of the emergency.

Two separate organizational arrangments are defined.

The offsite GSEP organization for emergencies of limited extent is shown in Figure E.1.

The limited reponse' consists primarily of Corporate Command Center personnel and may be activated for transportation accidents and unusual events or alert emergencies when the nearsite Emergency Operations Facility

_(EOF) is not activated.

In the event of an emergency at La Salle that is classified as a Site Area or General Emergency, the GSEP full response organi-zation will be activated at the nearsite E0F.

This organization is shown in Figure E.2.

We find that these changes to the GSEP do not downgrade the effectiveness of the emergency response and continue to meet the above standard of 10 CFR 50.47(b).

B.

Onsite Emergency Organization Emergency Plan We reviewed the applicant's program for onshift augn.'ntation within the first hour of a significant emergency.

Clarification relevant to this issue is provided.

The applicant submitted by letter dated October 2, 1980, their current capabilities for a prioritized shift augmentation. -That submittal indicated that certain personnel living within 10 miles of the plant could be made available, weather permitting, within 30 minutes and 60 minutes depending on their distance from the plant.

Although this description is not provided in the Plan, we reviewed this response as part of the applicant's overall emergency response capability.

Because the applicant's augmentation capability may change from time to time due to new employees, moves, etc., the applicant agreed to address augmentation priority in procedures and maintain its commit-ment to always provide at least t;he augmentation capability of 60 minutes as stated in the Plan. We find this adequate and will review any future changes in these procedures as required by the regulations.

C.

Emergency Response Support and Resources Emergency Plan No significant changes were made to the applicant's program in this area.

E-2

FIGURE E.1 LIMITED RESPONSE OFFSITE GSEP ORGANIZATION 1

CORPORATE COMMAND CiNTER DIRECTOR SYSTEM POWER DISPATCHERS LEGAL CONSULTANT MEDICAL

)

DIRECTOR i

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1 m

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ENGINEERING INTELLIGENCE ENVIRONMENTAL MANPOWER AND I

DIRECTOR DIRECTOR DIRECTOR LOGISTICS DIRECTOR l

I I

i I

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I I

i l

STAFF HEALTH PHYSICS INFORMATION STAFF ACCOUNTING COMMUNICATIONS DIRECTOR DIRECTOR DIRECTOR DIRECTOR I

I STAFF STAFF ERP DIRECTOR l

I I

I ENVIRONS STAFF DIVISION DIRECTOR DIRECTOR l

• STATION GROUP (S) l STAFF STAFF (ONSITE)

{

m

FIGURE E.2 FULL RESPONSE OFFSITE GSEP ORGANIZATION REC 0VERY MANAGER I

l I

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EMERGENCY ADMINISTRATION /

ADVISORY TECHNICAL i

PLANT i WASTE SYSTEMS ENVIRONMENTAL /

SCHEDULING DESIGN AND NEWS CENTER LOGISTICS SUPPORT SUPPORT

0PERATIONS:

RADIATION EMERGENCY PLANNING CONSTRUCTION DIRECTOR MANAGER DIRECTOR MANAGER i MANAGER i CONTROL C0ORDINATOR MANAGER MANAGER i

MANAGER g

g J

I i STATION i e

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GROUP :

m i

TRAINING DIRECTOR

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STAFF l (CNSITE) !

l STAFF l ENVIRONS l STAFF l l STAFF l

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I s.............t DIRECTOR j

STAFF l STAFF '

COMMUNICATIONS DIRECTOR STAFF NE ARSH E REWOTE J

INFORMATION MANPOWER INTELLIGENCE HEALTH

' ENVIRONMENTAL FROW ENGINEERING m

i DIRECTOR AND LOGISTICS DIRECTOR PHYSICS DIRECTOR SITE DIRECTOR i*

OfRECTOR g

g DIRECTOR g

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STAFF l STAFF l l STAFF l p

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I ACCOUNilNG '

ERP DIRECTOR DIRECTOR E

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pivlSION DIRECTOR CORPORATE COMMAND CENTER DIRECTOR l

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l STAFF l MEDICAL DIRECTOR LEGAL CONSULTANT SYSTEM POWER DISPATCHER I

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D.

Emergency Classification System Emergency Plan As stated in our Safety Evaluation Report, the applicant committed to incorporate and classify analyzed accidents in the Plan by fuel load.

This has.been completed and will be incorporated inte the applicant's next revision to the Plan.

4 Regarding predetermined Emergency Action Levels for the high range effluent monitors, the applicant stated this will be completed after final installation i

tnd calibration of the subject monitors.

The applicant confidently indicated t

j that this will be accomplished by January 1982.

However, the applicant expects this to be completed prior to fuel load.

We find this acceptable.

4 E.. Notification Methods and Proced'ure

' Emergency Plan t

The applicant submitted by letter dated April 27, 1981, updated information regarding a prompt public notification system which will be implemented at the La Salle Country Station.

The implemer.tation date of July 1981, has changed to October 19, 1981, due to equipment delays.

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The prompt notification system consists of three parts.

First, a permanently

-installed outdoor notification sytem within 0 to 5 miles radius around the station.

This system will essentially cover all inhabited areas with a minimum I

noise-level of 60 db.

For those areas where a dwelling is not exposed to 60 db, a local coverage siren or an inhouse warning receiver will be utilized.

Secondly, a permanently installed outdcor notification system' covering the Th:avily populated areas within the 5 to 10 mile radius will be installed.

- Areas such as Grand Ridge, Kinsman, Marseilles, Ranson, Seneca and Verona are included..These six communities will be covered by installed notification i

systems. -The same siren coverage (60 db minimum or 10'db above daytime back-ground) will be utilized.

Thirdly, a mobile notification system for the remainder of the area (5 to 10 mile radius) will be utilized.

This portion of the population will be alerted by a mobile system including sirens and public address. This includes the use of law enforcement vehicles and will be accom-plished by the county sheriff and state police.

This is further described in the State and County Plan.

We find this program adequate for fuel load and will await the FEMA findings on-this system.

F.

Emergency Communications Emergency Plan

- The microwave / radio communications system described in our Safety Evaluation R: port'is operational.

The microwave / voice channel telephone sytem, as described

-in our Safety Evaluation Report, will be operable by fuel load.

The applicant's

. system now includes primary and backup means of communications.

The system includes normal business and dedicated telephone lines backed up by a microwave 8

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radio channel which includes consoles, mobile units and handie-talkies. When the microwave / voice channel is operable, it will serve as a backup also.

G.

Public Information Emergency Plan No changes have been made to the Plan in this area.

The Office of Inspection and Enforcement verified it.at the public information program has been implemented during the April visit.

H.

Emergency Facilities and Equipment Emergency Plan The long-term requirements for a permanent E0F are described in the applicant's submittal of April 10, 1981.

The applicant intends to provide facility concept-ual design and documentation by June 1981.

See Item III.A.1.2 in Section 22 of this supplement for more details.

The applicant has committed to complete a permanent EOF by October 1982 in accordance with the current schedule of NUREG-0696.

Full parameter display will be accomplished by 1982.

We find the applicant nas provided a description of and a completion schedule for a permanent EOF.

This is adequate for full power operations. We will evaluate the conceptual design and other documentation of the EOF when the June 1981 submiti.C is received.

Since issuance of our Safety Evaluation Report, the applicant has upgraded the meteorological program.

Backup power to the meteorological tower has been accomplished, 6nd the Class 8 dose assessment model has been implemented at the Corporate Command Center.

The arplicant currently is in compliance with the three milestones of Appendix 2 of NUREG-0654 and has demonstrated that l

adequate compensating actions exist for those ar 7 which require long-term i

upgrading.

A revised meteorological program will be submitted to use in June 1981.

The Class A model with display in the TSC will be accomplished prior to fuel load.

I.

Accident Assessment l

Emergency Plan l

No changes were made in the Plan in this area.

J.

Protective Response Emergency Plan l

l The applicant submitted by letter dated March 27, 1981, upgraded evacuation time estimates including those for adverse weather conditions. We reviewed i

the applicant's procedures, which included a description of these time estimates, and their use for recommending protective actions.

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This portion of the applicants program is adequate.

K.

Radior:.gical Exposure Control Emergency Plan No changes were made in the Plan in this area.

L. Medical and Public Health Emergency Plan The revised applicant's plan provides additional medical backup support at the Northwestern Meeorial Hospital in chicago.

This hospital would be used to treat significant radiological emergencies requiring extensive hospital care.

St. Mary's Hospital is still the primary hosp' ital.

M.

Recovery and Reentry Planning and Post-Accident Operations Emergency Plan Since issuance of our Safety Evaluation Report, the applicant has adopted the

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INP0 criteria for Recovery Operations.

This organization (Figure E.2) is essentially the same as the full response offsite GSEP organization and will be activated for a Site Area or General Emergency.

A Recovery Manager is designated from the Commonwaalth Edison Company who has requisite authority, management ability, and technical knowledge to manage recovery operations.

The primary Recovery Manager is the Division Vice-President, Nuclear Stations.

Task assignments for this position are described in the Plan.

We find this portion of the applicant's Plan adequate.

N. Exercises and Orills Emergency Plan Clarification of applicant's conducted full-scale and small-scale exercises was provided.

Full-scale exercises which test as much of the GSEP and State and local emergency plans as is reasonably achievable without mandatory public participation shall be scheduled in order to permit agencies to fulfill their full-scale exercise frequency requirements as listed below:

(1) States with jurisdiction in the plume exposure emergency plaining zone (EPZ) of a nuclear station at least once very five years at each nuclear

_f station and at least one full-scale exercise per year (somewhere).

(2) States with jurisdiction in the ingestion expo.sure EPZ of a nuclear station--at least once every three years (somewhere).

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(3) Federal emergency response agencies--at least once every five years at each nuclear statio-l (4)

Small-scale exercises which test the adequacy of communication links, establish t. hat emergency respcnse agencies understand the concept of emergency action levels, and test at least one other component of the offsite emergency response plan shall be conducted at each stuclear station each year that a full-scale exercise is not conducted.

The above mentioned exercise frequency is in accordance with the requirements of Appendix E of 10 CFR Part 50 and is acceptable.

For Illinois, this means that the State plan will be tested at least once every year with one of the applicant's plant (Zion, Dresden, Quad-Cities or la Salle), and each plant will have this test at a frequency not to exceed five years.

Other states (Wisconsion and Iowa) within the 10 mile EPZ will be jointly exercised with Illinois at least once every five years.

No other states, other than Wisconsin and Iowa, are within the EPZs; therefore, the ingestion EPZ will be exercised with the inhalation 10 mile EPZ.

The revised plan indicates that Health Physics drills will be conducted semi-anually which involve response to and analysis of simulated elevated airborne and liquid samples and direct radiation.

At least one of these drills will include an actual test of the post-accident sampling system with actual radiation levels from normal plant operation.

We find this portion of the applicant's program adequate.

O.

Radiological Emergency Response Training and P.

Responsibility for the Planning Ef fort:

Development Periodic Review, and Distributien of Emergency Plans Emergency Plan i

Both of these sections of the applicant's plan have not changed except for title changes to some managers.

These sections of the applicant's plan are considered adequate.

1 E-8

Conclusion Based on our review of the applicant's emergency preparedness plan, we conclude that the Commonwealth Edison GSEP and the La Salle Station Site Specific Annex meet the planning standards of 10 CFR 50.47(b) and the requirements of 10 CFR Part 50, Appendix E.

Findings made in accordance with 10 CFR 50.42(a)(2) concerning emergency preparedness implementation will be examined by our Oftice of Inspection and Enforcement as part of their routine prelicensing program.

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APPENDIX F j

LETTER FROM THE FEDERAL EMERGENCY MANAGEMENT AGENCY M

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MAY 13 19 81 Mr. Brian Grimes

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Dear Brian:

This responds to your memoranda of March 10 and March 17, 1981, requesting findings and detrmination on the status of State and local off-site preparedness relative to the LaSalle nuclear station in Illinois.

We have reviewed the State and local plans in draft as completed in December 1980. The State of Illinois has since, by letter dated March 31, 1981, formally submitted its radiological emergency plans to the Federal Emergency Management Agency (FEMA) Region V office.

It included the " State Plan for Radiological Accidents for the State of Illinois" and the local plans for LaSalle and Grundy Counties which are impacted by the plume zone.

The F"MA Region V Regional Assistance Committee (RAC) is currently reviewing the plans.

Two joint exercises have been held on the State plan; the first for the site specific Dresden plan on October 28, 1980, and the second on December 4,1980, for the LaSalle site. The LaSalle exercise evaluation confirmed the conclusion of the earlier exercise, that the State had effectively demonstrated a capability to protect the public in the event of a plant emergency.

The two counties impacted by the plume exposure zone, LaSalle and Grundy, showed widely differing degrees of capability in the exercise. This was LaSalle County's first experience with a radiological exercise and it generally revealed lack of knowledge ard familiarity of the procedures.

This is minor and correctable through further training, the use of checklists and SOPS. Grundy County, however, has given a very poor performance in both the LaSalle and Dresden exercises by casual indifference and lack of knowledge of its plans and pro-cedures.

There is no reference in any of the plans to meeting the requirements for alerting and notification as described in Appendix 3 of Criteria E-6 of NUREG-0654/ FEMA-REP-1, Revision 1.

A public meeting was held on December 5, 1980, at Ottawa, Illinois', in accordance with 44 CFR 350.10.

The FEMA Region V has submitted its recommendations resulting from the LaSalle exercise to the State, but as of this date, the State has not responded.

From the foregoing, we find that the State of Illinois per se appears adequataly prepared to protect the public in the event of a nuclear accident at the LaSalle site. We find that while deficiencies exist in the capabilities of LaSalle County, it ooes have an adequate capability to respond, but conditioned upon further training of ite officials and developing checklists and SOPS. We find that Grundy County, having performed poorly in both the Dresden and LaSalle exercises, must be judged as not having a capabili.ty to protect the public and at substantial improvements are needed to meet a leve of adequacy. Finally, we find that th alerting and notification system does not meet the current criteria.

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The foregoing describes the status of the plans and preparedness for the La5911e site at this time. At such time as we complete the formal process outined in 44 CFR 350 (proposed), we will promptly p;' ovide final findings and determination.

Sincerely yours, Q oE61.~Dihke'y

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BIBLIOGRAPHIC DATA SHEET Supplement 1 TITLE AND SUBTITLE LAM Volume No. of apprmnavl

2. (Leave blankt Supplement No. 1 to the Safety Evaluation Report related to the operation of LaSalle County Station, 3 RECIPIENT'S ACCESSION NO.

Units 1 and 2 AU T HO R (S)

5. DATE REPORT COMPLE TED M ON TH l YE AR June 1981 PERF ORMING ORGANIZATION N AME AND MAILING ADDRESS (/nclude Zip Codel DATE REPORT ISSUED Office of Nuclear Reactor Regulation I"1981

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

20555 6 " **" ' #

8. (Leave Nankl

. SPONSORING ORGANIZ ATION N AME AND M AILING ADDRESS (Include lep Codel p

11. CONTR ACT NO.

Same as 9 above

1. TYPE OF REPORT PE RIOD COV E HE D (inclus.ve d.#res)

Safety Evaluation Report SUPPLEMENTARY NOTES 14 (Leeve n/ankt 1.

Pertains to Docket Nos. SC-173 and 50-374

h. ABSTR ACT 200 words or less/

jSupplement No. 1 to the Safety Evaluation Report of Commonwealth Edison Company's application for licenses to operate its LaSalle County Station, Units 1 and 2, located in La Salle County, Illinois, has been prepared by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission.

This supplement provides further information on outstanding items from the Safety Evaluation Report.

7. KE Y WOR DS AND DOCUME N T AN ALYSIS I la OL SC HIP T O HS 7b IDEN TIFIE RS OPE N ENDE D TE HMS S AV AILABILITY ST ATEMENT 19 SE CURITY CL AZ /Tn s r porr/

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