ML18018B716

Safety Evaluation Report Related to the Operation of Shearon Harris Nuclear Power Plant,Unit 1.Docket No. Stn 50-400. (Carolina Power and Light Company,North Carolina Eastern Municipal Power Agency)

ML18018B716
Person / Time
Site:	Harris
Issue date:	06/30/1984
From:	Office of Nuclear Reactor Regulation
To:
References
NUREG-1038, NUREG-1038-S01, NUREG-1038-S1, NUDOCS 8407180053
Download: ML18018B716 (53)

Text

NUREG-1038 Supplement No.

1 SKIQ 2'VK/MK'iZDotE IRpoEPR rellatedl tc tlute ope~atio~ of Slhearon Harris Mucllear Power Plant, Unit Mc.

'll Docket No. STN 50-400 Carolina Power and Light Company North Carolina Eastern Municipal Power Agency U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation June 1984

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

1.

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

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NUREG-1038 Supplement No.

1 Safety Evaluation Report related to the operation of Shearon Harris Nuclear Power Plant, Unit No.

1 Docket No. STN 50-400 Carolina Power and Light Company North Carolina Eastern Municipal Power Agency U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulatiori June 1984

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ABSTRACT This report, Supplement No.

1 to the Safety Evaluation Report for the applica-tion filed by the Carolina Power and Light Company and North Carolina Eastern Municipal Power Agency (the applicant) for license to operate the Shearon Harris Nuclear Power Plant Unit 1 (Docket No. 50-400),

has been prepared by the Office of Nuclear Reactor Regulation of the U.S.

Nuclear Regulatory Commission.

This supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report.

'hearon Harris SSER 1

ABSTRACT TABLE OF CONTENTS

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INTRODUCTION AND GENERAL DESCRIPTION OF PLANT..

l. 1 Introducti on.

1.7 Summary of Outstanding Issues.........

1.8 Confirmatory Issues 1.9 License Conditions..

3 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS...........

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1-1 1-3 1-5 3-1 3.5 Missile Protection..

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3-1 3.5.1 Missile Selection and Description......................

4 REACTOR.

4.4 Thermal-Hydraulic Design.

4.4.6 NUREG-0737 Item II.F.2.......~..

3-1 4-1 4-1 10 STEAM AND POWER CONVERSION SYSTEM...

10.4 Other Features 10.4.9 Auxiliary Feedwater System.

12 RADIATION PROTECTION..............

12.5 Operational Radiation Protection Program.

12. 5. 1 Organization.

13 CONDUCT OF OPERATIONS

13. 3 Emergency Pl arming.

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13. 3. 1

13. 3. 2

13. 3. 3 Introduction.

Evaluation of the Emergency Plan..

Conclusions 13-1 13-1 13-6 13.6 Physical Security.

13.6.3 Proposed License Condition.

13-6 13-6 Shearon Harris SSER 1

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TABLE OF CONTENTS (Continued) 18 HUMAN FACTORS ENGINEERING.

18. 1 Gener al.

18.2 Planning Phase 18.3 gualifications and Structure of the DCRDR Team.

18.4 Review Phase.

18.4. 1 System Function and Task Analysis 18.4.2 Control Room Inventory..

18.4.3 Control Room Survey 18.5 Assessment and Implementation Phase 18.5.1 Assessment of HEDs 18.5.2 Selection of Design Improvements 18.5.3 Coordination of Control Room Improvements with Other Programs 18.5.4 Proposed Schedule for Implementing Design Changes.

18.5.5 Justification for HEDs with Significance that Are Left Uncorrected or Partially Corrected.............

18.6 Conclusions Pacae 18-1 18-1 18-2 18-2 18-3 18-3 18-3 18-4 18-4 18-4 18-5 18-5 18-6 18-6 18-6 19 REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS.............

19-1 APPENDICES APPENDIX A CONTINUATION OF CHRONOLOGY OF NRC STAFF RADIOLOGICAL REVIEW OF SHEARON HARRIS APPENDIX B

BIBLIOGRAPHY APPENDIX D

ACRONYMS AND INITIALISMS APPENDIX E

PRINCIPAL STAFF CONTRIBUTORS APPENDIX G

ACRS REPORT Shearon Harris SSER 1

vi

1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT

1. 1 Introduction In November 1983, the Nuclear Regulatory Commission staff (NRC staff or staff) issued a Safety Evaluation Report, NUREG-1038, regarding the application by Carolina Power and Light Company and North Carolina Eastern Municipal Power Agency (the applicant) for a license to operate the Shearon Harris Nuclear Power Plant Unit 1.

This report is the first supplement to that Safety Evaluation Report (SER).

This supplement provides more recent information regarding resolution of some of the open items identified in the SER.

This supplement also provides and discusses the recommendations of the Advisory Committee on Reactor Safeguards (ACRS) in its report on the Shearon Harris Plant, dated January 16, 1984.

Each of the following sections or appendices of this supplement is numbered the same as the section or appendix of the SER that is being updated, and the discussions are supplementary to and not in lieu of the discussion in the SER unless otherwise noted.

Accordingly, Appendix A is an updated bibliography,"

Appendix D is a list of abbreviations used in this supplement.

Appendix E is a list of principal contributors to this supplement.

Appendix G is a copy of the ACRS report.

The Project Manager is Bart C. Buckley; he may be reached on (301) 492-8379.

1.7 Summar of Outstandin Issues Section 1.7 of the SER noted that certain information had not yet been provided by the applicant for several identified items.

This supplement updates those items for which additional information has subsequently been provided by the applicant.

These

items, and the sections of this supplement discussing the review conclusions, are (2)

Missiles Outside Containment (3. 5. 1. 1)

(13) Emergency Preparedness (13. 3)

(14) I.D.1 Control Room Design Review (18)

II.F.2 ICC Instrumentation (4.4.6)

The remaining outstanding items and the references to applicable sections of the SER are given in Table 1.2 below.

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

Shearon Harris SSER 1

Table 1.2 Outstanding items Item SER Section(s)

(1)

Design of retaining wall (2)

Missiles outside containment (3)

Functional capability of Class 1 auxiliary piping systems

2. 5.4 3.5.1.1 3.9.3 (4)

Control of minimum wall thickness in ASME Class 1,

2, and 3 piping systems 3.9.3 (5)

Equipment qualification (6)

Preservice/inservice inspection'rogram 3.10, 3.11 5.2.4, 6.6 (7)

Periodic testing of instrument air quality (8)

Fire protection'9)

Unmonitored release of condenser discharge during hogging operations (10) Method of estimating noble gas activity from atmospheric steam dump valves 9.3.1

9. 5.1

10. 4. 2, 11. 5

10. 4. 2, ll.5 (ll) Monitoring of all inputs to the service water system 11.5 (12) Emergency preparedness II (13) Steam generator tube rupture isolation time

13. 3

15. 6. 3 (14) TMI Action Plan Items (NUREG-0737 and Supplement No.

1 to NUREG-0737)

I.A.1.2 I.C.2 I.C.3 I. C.4 I.C.5 I. C.6 I.D.1 II.E.1.1 III. A. 1. 2 III. D. 1. 1 Shift supervisor administrative duties Shift and relief turnover procedures Shift supervisor responsibilities Control room access Feedback of operating experience Verification of correct performance of operator activities Control room design review Auxiliary feedwater system reliability evaluation Emergency support facilities Leak reduction program

13. 5. 1

13. 5. 1

13. 5. 1

13. 5. 1

13. 5. 1

13. 5. 1 18

10. 4. 9

13. 3. 4 9.3.5 Shearon Harris SSER 1

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1. 8 Confirmator Issues Section 1.8 of the SER stated that certain confirmatory information will be provided by the applicant.

One of these items, identified as item (30), Radia-tion Protection Manager (Section 12.5) in the SER, is resolved in Section 12.5. 1 of this supplement.

The remaining confi rmatory issues, with reference to the applicable sections of the SER,, are listed in Table 1.3.

Table 1.3 Confirmatory issues Issue SER Section(s)

(1)

Emergency plan meteorological program (2)

Revision of FSAR Table 3.2. 1-1 (3)

Turbine missiles (4)

Design documentation of ASME components (5)

Piping supports (6)

Plant-specific submittal concerning testing of safety and relief valves 2.3.3 3.2.2 3.5.1.3 3.9.3.1 3.9.2 3.9.3.2 (7)

Leak rate test program for pressure isolation valves 3.9.6 (8)

Calculation of ultimate strength capacity of contain-ment building under uniform internal pressure 3.8 (9)

Additional information on excore detectors (10)

PORV setpoint values (ll)

Revised pressure-temperature curves (12)

Examination of steam generators and NUREG-1014 revisions 4.3 5.2.2 5.3.2

5. 4. 2. 2 (13)

Revision of FSAR on containment penetrations (14)

Additional information on adequacy of the ECCS during shutdown and startup

6. 2.4 6.3.5.1 (15)

Design modifications for automatic reactor trip using shunt coil trip attachment 7.2.2.4 (16)

Solid-state logic protection system test circuit (17)

Testing for remote shutdown operation (18)

RCS overpressure protection during low temperature operation

7. 3. 3. 11 7.4.2.2 7.6.2.2 Shearon Harris SSER 1

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Table 1. 3 (continued)

Issue SER Section(s)

(19)

Adequacy of station electrical distribution (20)

Use of load sequencer with offsite power (21)

Compliance with Phase I and Phase II of NUREG-0612 (22)

Pressure differential alarms (23)

Emergency lighting (24)

Radiation monitors for turbine building vent stack (25).

Ability to continuously sample radioiodine and particulates (condenser vacuum pump effluent)

(26)

Location of high range noble gas monitors (turbine building vent)

(27)

Drawings for the filters handling sludge (28)

Process control program (29)

Polymer binder system (30)

Corporate management and technical support organization (31)

Initial test program

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Additional testing to verify the capacity of the steam generator safety and relief valves

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Amend FSAR to incorporate additional information on AMP endurance tests

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Expansion of natural circulation tests to fully comply with NUREG-0737, Item I.G. 1 (32)

TMI Action Plan (NUREG-0737)

I. C. 7 NSSS vendor review process II.K.3.5 Automatic trip of RCPs during LOCA 8.4.2.3 8.4. 7 9.1.5 9.4. 5. 2 9.5.3

10. 4

10. 4. 2

10. 4. 2,

10. 4. 3

11. 5

11. 4. 1 11.4. 1 ll.4. 1

13. l. 1. 6 14 13.5.2. 3

15. 9. 9 Shearon Harris SSER 1

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1.9 License Conditions The proposed license condition related to security plan adherence to the regu-lations was inadvertently omitted from the Shearon Harris SER dated November 1983.

(6)

Security plan adherence to the regulations (13.7)

Shearon Harris SSER 1

3 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS 3.5 Missile Protection 3.5. 1 Missile Selection and Description t

3.5.1.1 Internally Generated Missiles (Outside Containment)

J SER Section 3.5. 1. 1 contained three open issues.

"These issues are discussed below as Items 1, 2, and 3.

During the review process, a fourth issue was raised concerning the possibility of missiles from pumps affecting safety-related

systems, structures, or components.

This issue, designated as Item 4, is also discussed below.

(1)

Fan Blades as Missiles The staff was concerned that fan blades could become missiles as illus-trated by the event at Palo Verde in 1982.

In that event, a fan blade became an internally generated missile and damaged the containment liner.

In response to the staff's concern, the applicant stated that specifica-tions for inline, centrifugal, and axial fans require that fan housing design prevent the expulsion of any missiles generated by fans operating at maximum conditions, as adjusted in the field, into areas outside the housing.

To accomplish this, vendors perform analyses and furnish calcu-lations to show that expulsion of postulated fan-generated missiles is precluded by the fan housing design.

The staff finds this acceptable.

(2)

Dama e from Missile Resultin from Failure of the AFW Turbine-Driven Pum In response to this staff concern, the applicant had stated that the tur-bine is designed with redundant overspeed protection to prevent failure as a result of overspeed; in addition, in case of such failure, the motor-driven auxiliary feedwater (AFW) pumps are prote'cted by barrier walls.

Finally, the applicant provided the results of an analysis of AFW turbine-missile trajectories and an examination of physical plant arrangement that conclude that there will be no adverse safety consequences in the event of a turbine missile:

The staff finds this acceptable.

(3)

Omission of Essential Services Chilled Water S stem (ESCWS and Waste Process Buildin Coolin Water S stem WPBCWS from List of S stems Re uirin Protection from Internall Generated Missiles The SER noted that neither the ESCWS or the WPBCWS appears on the list of structures,

systems, and components requiring protection against internal-ly generated missiles (outside containment)..

In,response, the applicant stated that the following had been considered as, potential sources of!mis-siles'hat could damage the ESCWS:

high pressure

systems, rotating ma-chinery, gravitational mi'ssi les, and secondary missiles (resulting from Shearon Harris SSER 1

3-1

the impact of primary missiles).

However, the applicant concluded that these missiles were either not credible or would not affect safety-related equipment in the ESCWS area.

The staff finds this conclusion to be unacceptable.

Missiles from these potential sources are considered credible unless some deliberate element in the design or extra precaution is provided to prevent their generation.

Design of equipment to appropriate codes is not a satisfactory means for preventing missile generation.

Therefore, to justify the conclusion that such missiles are not credible, the applicant must show, in detail, that the design specifically considered the problem of missile generation from these potential

sources, or the applicant must show that the ESCWS is pro-tected against such missiles.

Either of these approaches will satisfy the staff's concern regarding protection of the ESCWS against internally gen-erated missiles outside containment.

As for the WPBCSW, the applicant noted that no adverse safety or radiolog-ical impact results from failure of the nonnuclear WPBCWS and, therefore, the WPBCWS need not be protected against missiles.

The staff finds this acceptable.

(4)

Internall Generated Missiles from Pum s

During the staff review, a concern arose regarding the possibility of in-ternally generated missiles resulting from pump failure.

The applicant was made aware of this concern, and, in response, the applicant noted that missiles from pumps within the nuclear steam supply system (NSSS) scope.

that are outside of containment have been designed so that their maximum no-load speed is equivalent to their operating speed.

Thus, a sudden loss of load (resulting, for example, from a line break) will not result in the generation 'of missiles.

Further, the.FSAR states that the balance-of-plant (BOP) is designed so that missiles from internal sources will not damage engineered safety features in a way that would jeopardize the minimum re-quired safety functions.

This staff finds this unacceptable.

Although the pumps in the NSSS scope are prevented from overspeeding, the staff concern relates to the possi-bilityy of missile generation from well-designed pumps operating normally (see Item 3, above).

In addition, the applicant must provide detailed information to explain how safety-related structures,

systems, and compo-nents are protected against internally generated missiles generated from pumps within the BOP scope.

(5)

Conclusion The staff'finds the applicant's resolution of Items 1 and 2 acceptable.

However, the applicant must provide further information regarding protec-tion of the ESCWS against internally generated missiles outside of con-tainment (Item 3) and the protection of safety-related structures,

systems, and components against the effects of missiles generated from all pumps outside of containment (Item 4).

f.

Shearon Harris SSER 1

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4 REACTOR 4.4 Thermal-H draulic Desi n

4.4.6 NUREG-0737 Item II.F.2, Instrumentation for Inadequate Core Cooling Detection NUREG-0737 Item II.F. 2 clarifies the requirements for inadequate core cooling instrumentation (ICCI) that is to be installed and operational before fuel load.

On November 4, 1982, the Commission determined that an instrumentation system for detection of inadequate core cooling (ICC) consisting of an upgraded sub-cooling margin monitor, core exit thermocouples, and a reactor coolant invento-ry tracking system is required for the operation of pressurized water reactor (PWR) facilities.

In response, the applicant described the ICC system proposed for Shearon Harris in the following letters from M. A. McDuffie (CP8 L) to H.

R.

Denton (NRC):

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"Draft Safety Evaluation Report Responses,"

August ll, 1983

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"Responses to Requests for Additional Information," November 4, 1983

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"Draft Safety Evaluation Report Core Performance Branch,"

December 6,

1983 The staff's review of the information in those letters follows.

The applicant has selected an ICCI system that consists of three subsystems:

(1) subcooling margin monitor (SMM), (2) incore thermocouple

system, and (3) reactor vessel level instrumentation system (RVLIS).

The design uses a com-puter-based processing system (the emergency response facility information sys-

tem, ERFIS) for primary display of incore exit thermocouple and margin of subcooling data.

Although the ERFIS is not Class 1E, it receives qualified pressure and tempera-ture signals through an accessible

isolator, and is powered from a highly reli-able power source that is backed up by a battery.

The ERFIS computer margin of subcooling data and the incore exit thermocouple temperature are displayed on the safety parameter display system (SPDS) cathode ray tube (CRT) that is on the main control board (MCB).

The ERFIS has two redundant channels and is sin-gle failure proof.

The Shearon Harris emergency operating procedures (EOPs) and functional restoration procedures (FRPs) will be based on the Westinghouse owners group emergency response guidelines and functional restoration guide-lines.

These procedures employ ICCI (RVLIS, the core exit thermocouples, and the subcooling data) and other post-accident monitoring capabilities (reactor c'oolant system pressure, reactor coolant pump status, and safety injection flow).

Shearon Harris SSER 1

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Thus, the Shearon Harris ICCI will be used in accordance with the emergency response guidelines developed by the Mestinghouse owners group.

(2)

Subcoolin Mar in Monitor An SMM will be installed and operational before fuel load.

The ERFIS processes

)e wide-range reactor coolant system (RCS) pressure indicators and the incore

it thermocouples, both of which span the necessary range to preclude the need r overlapping instrumentation.

The ERFIS computes the margin of subcooling ta which is displayed, as well as. the incore exit thermocouple temperature on

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SPDS CRT on the MCB.

case both primary channels fail to determine the margin of subcooling, the licant proposes to use the fully qualified wide-range RCS pressure indica-s, in conjunction with the fully qualified incore exit thermocouple tempera-as and the American Society of Mechanical Engineers steam tables.

Core Exit Thermocou le S stem core exit thermocouple system was designed to meet the criteria in Attach-1 to Item II. F. 2.

socouples used for the core exit for each core quadrant (in conjunction core inlet temperature data) are sufficient to provide indication of radi-stribution of the coolant enthalpy (temperature) rise across representa-regions of the core.

rimary display includes the following:

h spatially oriented core map indicating the temperature or temperature lifference across the core (at each thermocouple location) is displayed on he CRT.

selective reading of core exit temperature, which is consistent with parameters pertinent to operator actions in connection with plant-specific ICC procedures, will be continuous on demand.

Direct readout and hard copy capabilities are available for all thermo-couple temperatures.

The range extends from 200'F to 2300'F.

Hard copy will be provided by computer printout.

Trend capability showing the temperature-time history of representative core exit temperatures is available on demand.

Ala: ms are provided in the control room.

The alarm setpoints, will be con-sistent with the decision points i~ the emergency operating procedures.

The CRT interface will be located in accordance with h'uman factors de$ ign principles to facilitate rapid access to requested displays.

CPLL provid-ed its human factors methodology for the MCB in a letter to the staff dat-ed June 1, 1983.

This submittal identified the methodologies and human engineering specifications that apply to items such as ICCI that were not defined when the detailed control room design review was performed.

>ron Harris SSER 1

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The backup display, which is in the control room, that has the capability for selective reading of each of the operable thermocouples.

The range extends from 200 F to 2300 F.

The types and locations of displays and alarms will consider the use of this information by an operator during both normal and abnormal plant conditions integration into emergency procedures integration into operator training other alarms that are activated during an emergency and the need for pri-oritization of alarms This system will be fully operational by fuel load.

(4)

. Reactor Vessel Level Measurement Information utilized to give the operator an advance warning of the approach to ICC and to monitor the recovery from ICC, if it occurs, is obtained through a

qualified instrumentation package.

The information is obtained by the use of the RVLIS and incore exit thermocouples.

The Westinghouse RVLIS being installed at Shearon Harris represents the most recent Westinghouse design.

It is a fully qualified, redundant system for mon-itoring water inventory in the reactor vessel.

Each of the two channels in-cludes differential pressure cells and transmitters for narrow-and wide-range monitoring over the full length of the vessel, with the reactor coolant pumps off (natural circulation) and on, respectively.

Additionally, narrow-range monitoring is provided for each channel of the upper plenum during natural cir-culation.

The microprocessor in each chanel utilizes these differential pres-sure signals in conjunction with other inputs (such as RCS pressure and temperature, loop resistance temperature detectors or incore thermocouples, and RVLIS reference leg temperature sensors) to compensate for density changes in the system legs and to provide direct water level readers for the operator.

gualified incore thermocouples are used to determine core exit temperature.

These 51 thermocouples (26 channel A, 25 channel

8) are inputs to and are pro-cessed by the RVLIS microprocessors.

Both RVLIS water level readings and in-core exit thermocouple data wi 11 be data-linked to the ERFIS computer for primary display on the SPDS CRT on the MCB.

The data link is supplied by an isolated non-Class lE output from the qualified RVLIS microprocessors.

Al-though ERFIS is not Class lE, it is powered by a highly reliable source.

The isolation device cabinets and ERFIS are readily accessible and adjacent to the main control room.

Additionally, qualified microprocessor outputs (RVLIS water level and thermo-couple data) will be transmitted to dedicated redundant backup displays.

These backup displays are alphanumeric and qualified (Class 1E),

and are located in the control room.

The primary and backup displays have a selective capability for providing RVLIS water level, thermocouple

data, and temperature mapping Shearon Harris SSER 1

4-3

functions.

A technical description of the system is in the Westinghouse manual ehtitled "RVLIS-Summary Report, December 1980."

This system will be fully operational by fuel load.

(5)

Staff Evaluation On the basis of its review of the information submitted by the applicant, the staff finds that the ICC detection system is in conformance with these require-ments of NUREG-0737, Item II.F.2 and II.F.2, Attachment 1.

However, the staff will ensure that Technical Specifications relating to the final ICCI system are submitted and approved before fuel load.

'.n addition the Unit 1 license wi11 be conditioned as follows:

1)

ICCI will be installed and preoperational tests will be completed before fuel load.

Startup tests and calibrations for which the core must be in place will be completed before operation above 5X of full power.

!)

Before the plant exceeds 5X power, an implementation letter report must be provided for staff review.

)

Before criticality, the modified emergency procedures that incorporate the generic Westinghouse RVLIS system for Shearon Harris must conform to ge-neric EOP guidelines relating to the use of the RVLIS, or deviations must be identified and explained.

hearon Harris SSER 1

10 STEAM AND POWER CONVERSION SYSTEM 10.4 Other Features 10.4.9 -Auxiliary Feedwater System The staff and its contractor, Brookhaven National Laboratory (BNL), have re-viewed the reliability analysis provided by the applicant relative to the un-availability of the AFW system to be able to provide the required amount of water to the steam generators in the event of loss of main feedwater (LMFW),,

loss of offsite power, and loss of all ac power.

The event consisting of loss of main feedwater with offsite power available imposes a greater demand on the AFW system than does the event consisting of loss of main feedwater without offsite power.

Therefore, this review considers only the case with offsite power available.

Table 10. 1 compares the applicant's results with the BNL results.

Both the applicant and BNL show satisfactory results when comparing the reliability of the AFW system when only one of the three AFW pumps (two motor-driven, one turbine-driven with the turbine-driven pump having twice the capacity of each motor-driven pump) is needed with the criterion of no more than 1 x 10-4 to 1 x 10-s failures per demand for loss of main feedwater (LMFW) and loss of offsite power (LOOP).

However, when two pumps are, needed for an LMFW, the AFW system unreliability (4. 6 x 10-4) exceeds the criterion of 1 x 10-4.

For loss of all alternating current (LOAC), there is no criterion for numerical unavailability.

The results of the study are utilized to ensure AFW availability independent of ac power.

Table 10. 1 Unavailability of the Shearon Harris AFW system Applicant's Results BNL Results Transient Mission Success A*

Mission Success A"

Mission Success 8"~

LMFW LOOP LOAC 6.6E-6

6. 1E-5

1. 9E-2

9. 2E-6 4.9E-5 2.5E-2

4. 6E-4 "Mission Success A refers to the cases in which only one pump is necessary for success.

""Mission Success B refers to the case (LMFW only) in which flow from both motor-driven pumps (or the turbine-driven pump) is required.

Shearon Harris SSER 1

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The applicant stated that mitigation of LMFW with offsite power available re-quired the AFW system to supply water to the steam generators at a greater rate than was required for the other two events (LOOP and LOAC).

The applicant pro-vided a certified AFW pump test curve to show that one motor-driven pump could supply more than the amount required to the steam generators for an

LMFW, LOOP, and LOAC event.

However, this did not take into account the recirculation line, which diverted part of the water back to the condensate storage tank.

This left insufficient water for the steam generators in the event of an LMFW.

Thus, LOOP and LOAC can be accommodated by one motor-driven pump, but LMFW (without offsite power) cannot if the recirculation 1'ine remains open.

This accounts for the need for two pumps and the unavailability value of 4.6 x 10-4 for the LMFW event.

When asked about the ability of a motor-driven pump to del.iver the flow rate required to mitigate the LMFW event, the applicant noted that if one of the two motor-driven pumps is unavailable because of loss of voltage, the recirculation line on the other pump (the operating pump) would close to allow sufficient water to the steam generators.

The valve on each recirculation line may also be closed by an operator, either from the control room or from the auxiliary control panel, thus permitting a sufficient flow of water to the steam genera-tor from one operating motor-driven AFW pump.

This information was relayed to BNL when it was used to perform a sensitivity study based on the results of the'eliability analysis that assumed that only one motor-driven AFW pump is required to mitigate the LMFW event with offsite power available (resulting in the unavailability value of 9. 2 x 10- ).

The original analysis was modified by multiplying the cutsets containing random failure unavailabi lities of either or both motor-driven pumps by conservative factors.

This conservatively calculated result showed an unavailability of approximately 2 x 10-s.

Therefore, the staff finds the unavailability of the Shearon Harris AFW system for the LMFW event also acceptable because the un-availability value is less than 1 x 10-4.

Shearon Harris SSER 1

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12 RADIATION PROTECTION 12.5 0 erational Radiation Protection Pro ram

12. 5. 1 Organization In Section 12.5. 1 of the SER, the staff stated that the person to fill the po-sition of Environmental and Radiation Control Manager had not been selected.

The person now selected has qualifications that meet the provisions of RG 1.8.

The staff considers this matter resolved.

Shearon Harris SSER 1

8 12-1

13 CONDUCT OF OPERATIONS 13.3 Emer enc Plannin 13.3. 1 Introduction The staff's evaluation of the applicant's emergency preparedness plan is provided in SER Section 13.3.

The deficiencies identified in that evaluation have been addressed by the applicant in (1) Revision 2 to the Shearon Harris Emergency Plan, February

1984, and (2) revisions to the Corporate Emergency Plan and Implementing Procedures, February 1984.

The revised sections of the plans and procedures were'reviewed against (1) the appropriate planning standards in 10 CFR 50.47, (2) the requirements of Appen-dix E to 10 CFR 50, and (3) the specific guidance criteria of NUREG-0654/

FEMA-REP-l, Revision 2, entitled "Criteria of Preparation and Evaluation of Radiological Emergency

Response

Plans and Preparedness in Support of Nuclear Power Plants,"

November 1980, which has been endorsed by RG 1. 101 (Rev. 2).

Section 13.3.3 contains the conclusions of the staff, based on the previous review and this review.

13.3.2 Evaluation of the Emergency Plan I

13. 3. 2. 1 Assignment of Responsibility (Organization Control)

The SER identified four open issues.

The applicant's response to these 'issues is as follows:

(1)

The applicant has provided specific information listing the response organizations, the contact (and alternate),

the location for the response, the response

time, and the agent for initial notification in Table G. 1-1 of Appendix G of the plan.

(2)

The applicant has shown the interrelationships of the emergency response organizations (a) before the activation of the technical support center (TSC) and the emergency operations facility (EOF), (b) after the activa-tion of the TSC, and (c) after the activation of both the TSC and EOF in Figures G-l, G-2, and G-3 of Appendix G of the plan.

(3)

The applicant has described the duties and responsibilities of the Emer-gency

Response

Manager and members of the manager's staff in the revised plan.

The EOF staff organization is shown in Figure 2.4-1 of the plan and described in detail in the Corporate Emergency Plan.

(4)

The applicant has provided signed agreements with 17 offsite response and support organizations, including local fire, rescue, and medical organiza-tions; county and state emergency management organizations; the Department of Energy; and the Institute of Nuclear Power Operations in Revision 2 of the plan.

Shearon Harris SSER 1

13-1

The staff has reviewed and evaluated the additional information provided by the applicant and finds that the resolution of staff concerns on these four issues is acceptable.

13.3.2.2 Onsite Emergency Organization The SER identified three open issues.

The applicant's response to these issues is as follows:

(1)

The applicant has clearly identified the conditions for transfer of authority from Shift Foreman (Site Emergency Coordinator-Control Room) to the Site Emergency Coordinator (TSC),

and from the Site Emergency Coordi-nator (TSC), to the Emergency

Response

Manager (EOF) in. Revision 2 of the plan.

These conditions are (a)

The facility (TSC or,EOF) is ready to be activated and assume emer-gency functions.

(b), The Site Emergency Coordinator (TSC) or the Emergency

Response

Manager (EOF) has received a,briefing on the status of the emergency.

(2)

(3)

II The applicant has revised Tables 2.2-1 and 2.2-2 to correspond with Table B-1 of NUREG-0654 so that. they specify the minimum onshift staffing available for emergencies and the capability for augmentation in 30 to 45 minutes with additional augmentation in 60 to 75 minutes.

The variations in times of arrival are determined by weather conditions.

The minimum staffing and the augmentation of the emergency staff follow the guidelines of NUREG-0654 and Supplement 1 of NUREG-0737 and are adequate.

The applicant has shown the relationship and interfaces between the various onsite and offsite response and support agencies in the revised plan.

These relationships are given in Table 4.0-1 and illustrated in Figures 2.2-1 and 2.4-1 and Figures G-l, G-2, and G-3 of Appendix G.

The

, table shows the organizations with the primary responsibility and the secondary'r support responsibility for 'each of 17 major emergency functions.

The staff finds that the applicant's response to the above three issues is satis-factory and that the staff's concerns have been fully resolved'.

/

13.3.2.4 Emergency Classification System The SER identified two open issues.

The applicant's response to these issues is as follows:

(1), The applicant plans to submit the Plant Emergency Procedures by September 28, 1984.

The staff will evaluate the Emergency Action Level (EAL) sets at that time.

(2)

The revised plan clearly states the role of the judgment of the Shift Foreman (or Site Emergency Coordinator) in assessing the status of the'plant, which is that he will declare any, one of the four emergency classes where EALs have been

exceeded, or in his judgment, the status of the plant warrants such a declaration.

The staff considers this matter resolved.

Shearon Harris SSER 1

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13. 3. 2. 7 Public Information The SER identified one open issue.

In response to this concern, the applicant plans to submit the public information brochure in June 1984 for staff review.

13.3.2.8 Emergency Facilities and Equipment The SER identified three open issues.

The applicant's response to these issues is as follows:

The applicant has described three seismic monitoring systems for the Shearon Harris site in Section 3.9.2 of the revised plan.

These consist of one system on site, a second system off site to provide remote reading of seismic activity at the Corporate Headquarters in Raleigh, and a third system consisting of two independent detectors with tape recorders located at points on site.

The recorded signals can be played back either in the control room or at Corporate Headquarters.

(2)

The applicant has stated in Section 3.4. 1 of the revised plan that approx-imately 2650 ft~ in the lunch room are available for use by approximately 200 people.

Table 3. 1-1 has been revised to include battery-powered lanterns and a Polaroid camera in the operational support center emergency supplies.

(3)

The applicant has described the onsite laboratory facilities and specified the onsite laboratory as the central point for receipt and analysis of all onsite samples in Section 3.9.7 of the revised plan.

The applicant also has described the mobile environmental monitoring laboratory, based at Shearon Harris Energy and Environmental Center (SHEEC),

and designated the mobile laboratory as the central point for receipt and analysis of all offsite radiological samples.

The staff fin'ds the applicant's response to these three issues acceptable

and, therefore, considers them closed.

13.3.2.9 Accident Assessment The SER identified four open issues.

The applicant's response to these issues is as follows:

(1)

The applicant has discussed the emergency response facilities information system (ERFIS) and the safety parameter display system (SPDS) in Section 3.9 of the revised plan.

The ERFIS is a computer system (with backup units for critical hardware) that samples flows, pressures, temperatures, fluid levels, radiation levels, radiological monitoring system (RMS) equipment, and valve status.

The SPDS is a software subsystem of the ERFIS.

The applicant also has described the RMS with process radiological monitors and effluent radiological monitors, the postaccident sampling system (PASS),

the meteorological monitoring system, and field monitoring capability.

A detailed description of this equipment and resources is contained in the FSAR.

(2)

The applicant has identified the meteorological data input to the ERFIS and the access to the ERFIS computer by the corporate meteorological Shearon Harris SSER 1

13-3

center in the revised plan.

The corporate center meteorologists analyze national and local weather data and provide localized weather forecasts for the CP8L system or Shearon Harris as required.

NRC and State agencies contact the corporate center for appropriately formatted and verified

data, both current and forecast.

(3)

The applicant has provided a discussion of the techniques used in dose projection by the ERFIS if any indicators are off scale or otherwise tagged as suspect or false.

The computer data bank contains the radio-nuclide-mix assumptions based on the accident source terms of FSAR Table 15.0.9-1.

These data are used as default values until actual sample data can be substituted.

(4)

The applicant has provided a discussion of the radiation monitoring system in Section 3.9.3 of the revised plan.

This system, described in detail in FSAR Sections 11 and 12, can provide the means for relating various measured parameters and activity measurements to the key isotopes listed in Table 3 of NUREG-0654.

Based on its evaluation of additional information provided by the applicant, the staff concludes that these issues are fully resolved.

13.3'. 10 Protective

Response

The SER identified six open issues.

The applicant's response to these issues is as follows:

The applicant has provided a description of the plant public address (PA) system and the areas that are covered by this system, including the controlled area outside the protected

area, in the revised plan.

Security personnel with portable loudspeakers may be used to augment the PA system.

(2)

The applicant has described the method by which continuous accountability will be maintained by the designated team leaders and managers in the revised plan.

These methods are discussed in detai 1 in Plant Emergency Procedure O-PEP-382.

(3)

The applicant has provided a summary of evacuation time estimates, con-densed from the Evacuation Time Study received in January 1984, in Table G.8-1 of the plan.

The staff is reviewing the study and will provide its evaluation in a future supplement to the SER.

(4)

The applicant has shown the radiological monitoring sampling locations in Figure 3.9-1 of the plan.

These locations are discussed in detail in the Plant Emergency Procedures.

(5)

(6)

The applicant has shown the locations of public shelter areas in Figure G-4 of the plan.

The exact location of the areas and other details are discussed in Parts 2.E, 3.E, 4.F, and 5.E of the state plan.

h The applicant has discussed the shelter protection provided by local housing in Section 4.5.2 of the plan.

Shearon Harris SSER 1

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On the basis of its review of the applicant's responses, the staff concludes that the applicant's protective response capability is acceptable subject to the staff's confirmation of the evacuation time study (item (3) above),

which will be addressed in a future supplement to the SER.

13.3.2. 11 Radiological Exposure Control The SER identified three open issues.

The applicant's response to these issues is as follows:

(1)

The applicant has provided for the decontamination of supplies, instru-

ments, and equipment according to plant procedures or their disposal as radwaste.

(2)

The applicant has described the measures to control access to drinking water and food supplies on site in Section 4.6.3.8 of the plan.

The potable water clear well will be manually isolated until samples can be analyzed.

If the well water is contaminated, potable water will be shipped into the plant.

Most of the food on site is packaged food in vending machines that can be disabled or isolated until samples are analyzed.

(3)

The applicant has provided for the storage of decontamination supplies and clean protective clothing at personnel decontamination facilities located in the plant, in the administration building, and the Shearon Harris Energy and Environmental Center.

On the basis of its review of the additional information provided by the appli-cant on these three issues, the staff concludes that the radiological exposure control program is acceptable and, therefore, considers these issues closed.

13.3.2. 13 Recovery and Reentry Planning and Postaccident Operations The SER identified one open issue.

In response to the staff's concern, the applicant has revised Section 6.4 of the plan to specify that notification of onsite and offsite organizations concerning the activation of the recovery organization will be initiated by the Emergency

Response

Manager and will follow plant emergency notification procedures.

Notification procedures wi 11 also list the new positions of the recovery organization.

The staff finds the applicant's response acceptable and considers this matter closed.

13.3.2. 14 Exercises and Drills The SER identified one open issue.

In response to the staff's concern, the applicant has described the policy on exercises and drills in Section 5.3 of the revised plan.

Detailed procedures for the conduct of exercises and drills are described in Corporate Emergency Plan Implementation Procedure CEPIP-18.

The revised plan states that exercises will be conducted under various weather conditions and that some will be unannounced.

It also provides for "free play" decisionmaking.

The Plant Emergency Coordinator, with the Director of Emergen-cy Preparedness, wi 11 determine corrective actions and will follow up on implementation.

On the basis of its evaluation of Section 5.3 of the revised plan, the staff finds the applicant's policy acceptable.

Shearon Harris SSER 1

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13.3.2. 15 Radiological Emergency

Response

Training The SER identified two open issues.

The applicant's response to these issues is as follows:

(1)

The applicant has provided details of the training to be provided to specific categories of emergency response personnel in the revised plan.

The plan specifies that initial training and annual retraining will be provided.

It also provides assistance with the training of offsite emergency response personnel according to the radiological emergency

plan, and supplemental training as related to Shearon Harris.

The Emer-gency Plan Training Program is described in detail in Plant Emergency Procedure PEP-403 and in the Plant Emergency Plan Training Program Lesson Plans.

(2)

The applicant has specifically addressed the training of Damage Control Team personnel in Section 5.2.3.7 of the revised plan.

On the basis of its evaluation of the applicant's

response, the staff finds the applicant's training program acceptable.

13.3.3 Conclusions On the basis of its review of the Shearon Harris Nuclear Power Plant Emergency Plan in the SER and the review of the plan revisions, the staff concludes that, on satisfactory completion of those items identified in Sections 13.3.2.4 (1) and 13.3.2.7 of this report, the Emergency Plan will provide an adequate planning basis for an acceptable state of emergency preparedness.

After the staff has reviewed the findings and determinations made by the Federal Emergency Management Agency on the adequacy of state and local emergency re-sponse plans and after it has reviewed the revisions to the applicant's Emer-gency Plan, the staff wi 11 provide, in a supplement to the SER, its overall conclusions as to whether the state of onsite and offsite emergency prepared-ness provides reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency.

13.6 Ph sical Securit 13.6.3 Proposed License Condition The following proposed license condition was inadvertently omitted from the SER:

The licensee shall fully implement and maintain in effect all provisions of the Commission approved physical security, guard training and qualifi-cation and safeguards contingency plans, including amendments made pursu-ant to the authority of 10 CFR 50.54(p).

The approved plans, which contain Safeguards Information as described in 10 CFR 73.21, are entitled, "Shear on Harris Nuclea'r Power Plant Security Plan," Revision 1, dated July 5, 1983; "Shearon Harris Nuclear Power Plant Safeguards Contingency Plan," Revision 1, dated July 5, 1983; "Shearon Harris Nuclear Power Plant Security Personnel Training and qualification Plan," Revision 1, dated July 5, 1983.

Shearon Harris SSER 1

13-6

18 HUMAN FACTORS ENGINEERING As discussed in the

SER, Supplement 1 to NUREG-0737 requires the applicantto complete a detailed control room design review (DCRDR) before an OL can be is-sued for Shearon Harris.

The applicant departed from the NRC recommendation to provide a program plan early in their review process, followed by a summary report at the completion of the DCRDR.

The applicant submitted a combined program plan and summary report by letter dated December 7,

1982.

These were reviewed against the requirements of Sup-plement 1 to NUREG-0737.

Consultants from Lawrence Livermore National Labora-tory (LLNL) assisted the staff in the review and prepared the program plan comments and the in-progress audit report.

A human factors engineering in-progress audit of the control room design review was performed at the site August 15 through 19, 1983.

18. 1 General The applicant's submittal included elements of a program plan that was devel-oped and implemented in January 1981 and revised in September 1981 before the DCRDR requirements of NUREG-0737 Supplement 1 and the guidelines of NUREG-0700 were issued.

The submittal included a description of tasks that have been com-pleted and a summary report of the control room design review that noted the findings and corrective actions taken to date.

Before, during, and after the audit the applicant provided a number of support-ing documents.

These included

~

sample human engineering discrepancy (HED) sheets a sample HED status information report

~

a sample HED listing

~

the Shearon Harris Unit 1 control room design evaluation records file index

~

a representative set of interim main control boards (MCB) panel drawings the applicant's letter LAP-83-426, dated September 27, 1983, which con-tained additional information requested by the NRC audit team Available at the audit site were the control board simulator, which is signifi-cantly different from the MCB, which is still under construction.

The appli-cant's operating and engineering personnel and the applicant's human factors consultant from the Essex Corporation were available on a daily basis during the audit.

Shearon Harris SSER 1

18-1

The audit included briefings, discussions, document reviews, and a brief review of the simulator and the incomplete MCB.

The emphasis was on evaluating the organization and processes of the applicant's DCRDR.

The applicant's submittal contains elements of a program plan that meets most of the basic requirements of NUREG-0737, Supplement 1, in addition to describ-ing the related activities completed at the time it was submitted.

However, specific areas of the work should be described in greater detail and additional documentation should be made available for audit.

The additional documentation should clarify methodology, task procedures, and objectives to ensure that the applicant understands the requirements and processes of a DCRDR.

18.3 ualifications and Structure of the DCRDR Team The applicant established a multidisciplinary review team for the DCRDR that included applicant and Essex Corporation personnel, representing a cross sec-tion of the required disciplines.

The qualifications of the applicant's review team members are in Appendix A of the applicant's supplemental summary report dated September 27, 1983.

A description of the review team composition was included in Attachment 1 and Appendix A of the applicant's letter of June 1,

1983.

The applicant's review team was divided into three groups (1) human factors evaluation (2) human. factors/operations support (3) project management/nuclear operations/plant engineering and design group The Human Factors Evaluation Group was composed of six intermediate-and junior-level human factors specialists from the Essex Corporation.

The group's re-sponsibility was to collect and reduce data and perform preliminary data analysis.

However, except for the group supervisor, the extent of each group member's participation is not clear.

The group supervisor's responsibility was to coordinate the evaluation activities, interface with the applicant s project

manager, and represent the human factors position at HED review meetings.

The Human Factors/Operations Support Group was composed of Essex Corporation home office personnel.

This group consisted of five senior-level human factors specialists, a nuclear engineer, two reactor operators, three junior-level hu-man factors specialists, a procedure specialist, and a photographer.

The over-all responsibility of this group was to review:

in-depth analysis; discrepancy definition; recommendations for resolution of discrepancies; data collection support; and operational and engineering analysis.

The group leader was not identified, and the extent of each member's participation is unclear.

The Project Management/Nuclear Operations/Plant Engineering and Design Group was composed of applicant personnel from various disciplines, representing the Nuclear Plant Engineering Department, System Planning and Coordination Depart-ment, Nuclear Operations Department, and Nuclear Operations Department Train-ing.

The group leader coordinated the activities of Essex, the applicant, Shearon Harris SSER 1

18-2

EBASCO (the architect/engineer),

and Mestinghouse (the nuclear steam supply system vendor).

The group leader also reviewed the overall progress of the DCRDR.

EBASCO and Westinghouse personnel provided any required review and com-ments on design philosophies, discrepancy

analysis, etc.

On the basis of its review, the staff finds that the review team, as described above, satisfies Supplement 1 to NUREG-0737.

18.4 Review Phase 18.4. 1 System Function and Task Analysis The applicant performed its task analysis, using NUREG-1580 guidelines, which differ from the NUREG-0700 guidelines.

(When NUREG-0700 was issued, the appli-cant's analysis was already in progress.)

The task analysis guidelines in NUREG-1580 are based on analyzing procedures to document operator actions, in-formation requirements, and controls and displays used in executing procedures.

The NUREG-0700 task analysis guidelines are based on a systematic top-down function/task analysis to determine information and control capability require-ments that can be objectively compared to the actual instrumentation and controls available in the control room.

In response to an NRC audit team request, the applicant stated in a letter dat-ed September 27, 1983, that a generic top-down task analysis--based on the iden-tification of event sequences, plant systems, and operator functions and tasks as recommended in NUREG-0700--was performed on the High Pressure Basic Version of the Mestinghouse Owners Group (WOG) emergency response guidelines (ERGs).

The staff review of the WOG guidelines indicates that they constitute a satis-factory start in identifying and describing tasks, but they are generic and not plant specific.

The guidelines do not identify information and control re-quirements in sufficient detail to allow a comparison to the control room in-strumentation and control room inventory.

There was no substantiating documentation on the details or completeness of the methodology used in the task analysis.

On the basis of its review, the staff finds that the applicant's task analysis process described in the summary report may not identify instrument and control requirements independent of existing instrumentation in the control room and may not ensure that the man-machine interface is complete.

Thus, the staff finds that the applicant must provide a more complete rationale and justifica-tion of the method for conducting task analysis for the staff to determine whe-ther the requirements of Supplement 1 to NUREG-0737 have been satisfied.

18.4.2 Control Room Inventory During the in-progress audit, the applicant stated that the EBASCO panel compo-nent list had been substituted for a control room inventory.

This list was used to check the existence of controls and displays on the MCB panel drawings to confirm the expected panel content versus need.

(The instrument and control needs were determined by EBASCO and the applicant utilizing their expertise and operating experience.)

Because of the list used and the fact that the WOG generic function and task analysis does not provide a detailed list of control Shearon Harris SSER 1

18-3

and display characteristics, an accurate comparison of control and display re-quirements with the inventory is not possible.

18.4.3 Control Room Survey The Shearon Harris control room survey was carried out through reviews of plant design documents, vendor documentation, and layout drawings.

The Shearon Har-ris simulator was used where applicable.

A to-scale paper mockup was also used as a basis for identifying HEDs and as a star ting point from which to relocate controls and diaplays.

The control room survey process described by the applicant in the summary re-sort and during the onsite audit satisfies the requirement of Supplement 1 to IUREG-0737.

8.5 Assessment and Im lementation Phase 3.5. 1 Assessment of HEDs

>e applicant's review team did not perform a formal, documented assessment of

Ds, as required by NUREG-0737, Supplement l.

Instead, the assessment was e of several, parallel, iterative processes executed during what the appli-nt called the redesign of the MCB.

~ initial HED identification process was completed on each panel in the con-

)1 room.

HEDs were screened by the onsite human factors consultant in con-iction with the applicant's review Project Manager.

HEDs that required cor-tion were then discussed (assessed) in batches in meetings attended by the licant's review team (which included the human factors consultant, the licant's review Project Manager, applicant operating and engineering person-and EBASCO and Mestinghouse personnel, as appropriate).

The applicant's ject Manager and the Essex human factors consultant attended all meetings.

!r members of the applicant's review team were present when their expertise needed.

Many redesign review meetings were held during which the control room was re-peatedly and iteratively redesigned.

The appl.icant stated that this process included discussion of many alternate solutions before the final solution to correct each HED was selected.

The applicant's policy was to correct every HED by designing it out of the sys-tem, with the objective of achieving an HED-free board.

If successful, the execution of this philosophy obviates the necessity to formally assess each HED for importance, potential safety consequences, cumulative impact of minor HEDs, determination of priorities, and setting implementation schedules.

However, for those instances in which an HED with safety significance cannot be designed out of the system, and a decision is made by the applicant not to correct the HED or to only partially correct it, the applicant must provide justification for the action taken.

Ihe staff finds that the assessment methodology'used by the applicant meets the "equirements of Supplement 1 to NUREG-0737.

'hearon Harris SSER 1

18-4

18.5.2 Selection of Design Improvements Alternate design improvements for each HED were considered during the appli-cant's redesign review, which was done at the same time as the redesign of the MCB.

The review team used NUREG-1580 and human engineering requirements speci-fications (HERS) developed for the applicant by consultants as the bases for the HED resolutions.

As described in the applicant's summary report, the rede-sign effort appears to have been similar to the effort involved in the determi-nation of a new design, The applicant stated that the MCB redesign review resulted in the removal of about 200 controls and the relocation of a significant number of unnecessary items from front-to-rear panels.

EBASCO and Westinghouse concurred with these changes.

Appendix D of the applicant's report "Recommended Control Room Equip-ment Arrangement" describes how the recommended panel arrangements were deter-mined; factors considered included desirability of equipment location, readability of displays and labels from various MCB locations, visual angles, adverse effect on the operator's short-term memory because of distances opera-tors had to move between equipment, discrimination problems caused by viewing distances, etc.

In addition, a quarter-scale mockup was used to study and con-firm rearrangement of the control board instrumentation.

The mockup was also used to verify and confirm labeling, demarcation, and annunciator arrangement.

The redesign effort included ongoing discussions with operators to verify that the MCB redesign details and HED corrective actions were compatible with opera-tional needs and left no discrepancies uncorrected.

This included verification that any equipment found missing would be installed before the redesign was considered complete.

On the basis of its review, the staff finds that the methods used for selection of design improvements to correct HEDs described in the applicant's summary report and during the onsite audit satisfy the requirements of Supplement 1 to NUREG-0737.

However, the applicant must submit for staff evaluation informa-tion concerning any HEDs identified during the verification process, their resolution, and an acceptable time schedule for implementing corrective actions.

18.5.3 Coordination of Control Room Improvements with Other Programs The Shearon Harris emergency operating procedures (EOPs) are being developed using modified procedures from the H.

B.

Robinson plant and the Shearon Harris simulator.

The WOG ERGs will be considered in the development of the Shearon Harris

EOPs, and the applicant has stated that the EOPs will be subject to task analysis before they are implemented.

The results of the task analysis will be integrated back into the EOPs and into the DCRDR.

The applicant then plans to complete the system function and task analysis and verify and validate the EOPs after they are implemented and the MCB is fully functional.

The applicant stated that any subsequent MCB modifications will be made following the same processes used in the MCB redesign.

Although the applicant described this coordination effort, the applicant did not provide information on coordinating the DCRDR activity with the

SPDS, RG 1.97, and the emergency response facilities.

Shearon Harris SSER 1

18-5

18.5.4 Proposed Schedule for Implementing Design Changes The applicant's report does not include information on implementation of design changes because the applicant's policy of correcting all HEDs during the redesign process before operation obviates the need for implementing design changes.

In letter LAP-83-156 (dated June 3, 1983) the applicant states that the verifi-cation and validation of the control room redesign after the WOG task analysis and generic guidelines are available, and the completion of the Shearon Harris EOPs verification and validation will reveal few, if any, MCB human factors concerns.

If any HEDs are identified, they will be corrected via the continua-tion of the applicant's redesign corrective action policy.

18.5.5 Justification for HEDs with Safety Significance that Are Left Uncorrected or Partially Corrected For the reasons discussed

above, the applicant's policy of correcting all HEDs identified during redesign of the MCB essentially satisfies this requirement"of Supplement 1 to NUREG-0737.

18.6 Conclusions The applicant is continuing its DCRDR and improving the redesigned MCB.

On the basis of its review of the applicant's reports and the results of the in-progress onsite audit, the staff has determined that the applicant is conduct-ing a DCRDR that will substantially meet the guideines of NUREG-0700 and 'the requirements of Supplement 1 to NUREG-0737, except that the applicant must (1)

Provide a detailed description of the Shearon Harris system function and task analysis.

(2)

Describe the process used to compare display and control requirements as determined by the function and task analysis, with the control'oom inventory.

(3)

Describe the process to be used to verify that the corrective actions achieved the desired improvement without introducing new HEDs into the control room.

(4)

Provide a supplementary summary report that addresses items not included in the summary report because they are not ready for review.

These items are identified below.

The NRC Audit Team reviewed the partially completed'CB and independently veri-fied many of the HEDs identified by the applicant;

however, the applicant's HEDs and those found by the applicant were not matched on a one-to-one basis.

For each of the HEDs listed in part A of the in-progress audit report, the ap-plicant should give the staff the status and proposed resolutions, as well as a

schedule for implementing corrective action.

This should be submitted at least 6 months before fuel load.

Any HEDs found must be corrected before fuel load or on a schedule approved by the staff.

Shearon Harn s SSER 1

18-6

The items listed below were not included in the applicant's report because they were not ready for review.

For the DCRDR to be complete, the applicant must provide the results of the evaluation of the following items so the staff can determine whether the requirements of Supplement 1 to NUREG-0737 have been satisfied:

(1) workspace (2) communications (3) remote shutdown panel (4) recorder panel (5)

CRTs (6) process computer and peripherals (7) annunciator systems The applicant should submit the additional information required and results of the evaluation of the items discussed above at least 6 months before the Shearon Harris licensing date so the staff has time to evaluate the submittal.

Shearon Harris SSER j.

18-7

19 REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS During its 285th meeting, January 12-14, 1984, the Advisory Committee on Reac-tor Safeguards (ACRS) reviewed the application of Carolina Power and Light Com-pany and the North Carolina Eastern Municipal Power Agency for a license to operate the Shearon Harris Nuclear Power Plant, Unit 1.

This application was also considered at the subcommittee meeting held on January 3-4,'1984, in Apex, North Carolina, at which time members of the subcommittee toured the plant.

Transcripts of each of these meetings are available at the Wake County Public Library, 104 Fayetteville Street,

Raleigh, NC 27601.

Copies of these tran-scripts are also available for review at the NRC Public Document Room at 1717 H

Street, NW, Washington, DC.

The ACRS issued a letter report January 16,

1984, a copy of which is attached as Appendix G to this report.

The ACRS concluded that if due regard is given to items identified in its Janu-ary 16, 1984, letter report (which are described below),

and subject to resolu-tion of the outstanding and confirmatory issues identified in the Shearon Harris SER and satisfactory completion of construction, staffing and pre-operational testing, there is reasonable assurance that the Shearon Harris Nu-clear Power Plant can be operated at core power levels up to 2775 MWt without undue risk to the health and safety of the public.

These items are (1)

Control Room Emer enc Air Recirculation S stem The Committee expressed its desire to be kept informed about the applicant's operational test of the control room emergency air recirculation system, in-cluding control room habitability during the recirculation mode.

The applicant has scheduled this test for January 1985.

The test procedure and test results will be made available to the ACRS.

(2)

Westin house 0-4 Steam Generators The Committee asked that it be kept informed about the operating experience of the Westinghouse D-4 steam generators that are being used at Shearon Harris, Unit 1 and other nuclear facilities.

The ACRS will be kept informed of the operating experience with these steam generators.

Commanche

Peak, Unit 1 is the lead plant with 0-4 steam generators.

An inspection of the steam genera-tors at Commanche

Peak, Unit 1 will be performed before it exceeds 1 year of full-power operation.

The results of the inspection will be made available to the ACRS.

(3)

S stematic Assessment of Licensee Performance SALP The Committee requested that the two scheduled SALP reports should be provided to the ACRS before full-power operation to support the reported improvement in the applicant's management.

The staff will comply with this request.

Shearon Harris SSER 1

19-1

The Committee received a letter from a member of the public that makes several allegations concerning quality assurance and other issues and asked that the staff investigate the allegations and provide a written report to the Commit-tee.

The staff is preparing a response to the allegations that will be issued in the near future.

The Committee recommended that specific attention be given to assurance of ade-quate seismic capability of the emergency ac power supplies, the dc power sup-plies and small components such as actuators and instrument lines that are important to the accomplishment of safe shutdown and decay heat removal.

The committee has also requested that specific attention be given to the adequacy of clearances between adjacent buildings.

The staff will, as part of its seis-mic audit, pay specific attention to the items above as recommended by the ACRS.

The staff will also address the'dequacy of the clearances between adjacent buildings and report its findings in a future supplement to the SER.

(6)

Essential Chilled Water S stem As requested by the Committee, the staff will provide a detailed discussion of the essential chilled water system in a future supplement to the SER.

Shearon Harris SSER 1

19-2

APPENDIX A CONTINUATION OF CHRONOLOGY OF NRC STAFF RADIOLOGICAL REVIEW OF SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 November 22, 1983 Letter from applicant transmitting response to draft SER open item regarding service water sampling.

November 23, 1983 Letter from applicant transmitting responses to draft SER open items.

November 28, 1983 Letter from applicant transmitting revised response to draft SER open item 172.

December 1,

1983 Letter from applicant forwarding support information for response to draft SER open item 275.

December 2,

1983 Letter from applicant transmitting response to request for additional information.

December 2,

1983 Issuance of Generic Letter 83-32, NRC Staff Recom-mendations Regarding Operator Action for Reactor Trip and ATWS.

December 6,

1983 Letter from applicant forwarding revised response to draft SER open item 31.

December 9,

1983 December 9,

1983 Letter to applicant regarding facility staffing survey.

Letter from Transamerica Delaval regarding diesel generators installed in nuclear power plants.

December 13, 1983 Issuance of Safety Evaluation Report.

December 15, 1983 Letter from Transamerica Delaval regarding users'roup meeting held November 30, 1983.

December 16, 1983 Letter from applicant forwarding response to facility staffing survey.

December 19, 1983 Issuance of Generic Letter 83-43 -- Reporting Require-ments of 10 CFR Part 50, Sections 50.72 and 50.73, and Standard Technical Specifications.

December 19, 1983 Issuance of Generic Letter 83-42 -- Clarification to Generic Letter 81-07 Regarding

Response

to NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants."

Shearon Harris SSER 1

Appendix A

December 20, 1983 Issuance of Generic Letter 83-44 Availability of NUREG-1021, "Operator Licensing Examiner Standards."

December 21, 1983 Letter from applicant counsel to ASLB advising of Unit 2 cancellation.

December 29, 1983 Letter from applicant forwarding "Evacuation Time Estimates for Plume Exposure Pathway Emergency Planning Zone."

January 3-4, 1984 January 5,

1984 January 5, 1984 January 6, 1984 ACRS Subcommittee meeting with staff and applicant.

Issuance of Generic Letter 84-01 --

NRC Use of the Terms, "Important to Safety" and "Safety Related."

Board Notification 84-004 -- Environmental gualification Briefing of Chairman by Sandia.

Issuance of Generic Letter 84-02 Notice of Meeting Regarding Facility Staffing.

January 10, 1984 Letter from applicant transmitting revised "Management Capability Report."

January 13, 1984 Issuance of Generic Letter 84-03 -- Availability of NUREG-0933 on Generic Safety Issues.

January 16, 1984 January 17, 1984 Letter from Advisory Committee on Reactor Safeguards.

Letter from applicant transmitting Amendment No.

11 to FSAR.

January 19, 1984 Meeting with applicant to discuss its design of the alternate shutdown system in the event control room is unavailable.

January 23, 1984 Letter from applicant forwarding "quarterly Data Report -- Seismic Monitoring Program, October December 1983."

January 24, 1984 Letter from applicant to IE regarding radiological aerial survey.

February 1, 1984 Issuance of Generic Letter 84-04 "- Safety Evaluation of Westinghouse Topical Reports Dealing With Elimination of Postulated Pipe Breaks in PWR Primary Main Loops.

February 6, 1984 Letter to applicant requesting additional information regarding Transamerica Delaval emergency diesel generators.

February 7, 1984 Meeting with applicant to discuss functional capability of Class 1 auxiliary piping systems.

Shearon Harris SSER 1

Appendix A

February 7,

1984 Letter from applicant transmitting additional infor-pation on minimum wall thickness of Class 1, 2, and 3

piping.

February 13, 1984 Board Notification 84-020 -- Report of Meeting of Representatives of the Transamerica Delaval, Inc.

(TDI) Emergency Diesel Generators Owners'roup.

February 14, 1984 Letter to applicant regarding deletion of home telephone

numbers, unlisted utility numbers, etc.

from emergency plans.

February 16, 1984 Meeting with applicant to discuss auxiliary piping systems.

February 20, 1984 Letter from applicant forwarding response to request for additional information from Materials Engineering Branch on low pressure turbine rotors.

February 27, 1984, Letter from Long Island Lighting Company forwarding "Investigation of Types AF & AE Piston Skirts" related to diesel review.

February 28, 1984 March 2, 1984 Letter from Long Island Lighting Company forwarding task descriptions for 16 known problems on diesel generators.

Letter from Long Island Lighting Company forwarding TDI Diesel Generator Owners'roup Program Plan.

March 6, 1984 March 7, 1984 Letter from applicant forwarding shift experience information.

Letter from,applicant forwarding information on fuel handling building retaining wall design.

March 7, 1984 Letter from applicant transmitting preservice inspection program implementing procedure and flow diagrams.

March 8, 1984 Letter from applicant transmitting Revision 2 to Emergency;,P 1 an.

March 12, 1984 Letter from Long Island Lighting Company transmitting TDI Diesel Generator Owners'roup Design Review Report on Connecting Rod Bearing Shells for Shoreham and Grand Gulf.

March 12, 1984 March 13, 1984 Letter from applicant transmitting "North Carolina Emergency

Response

Plan."

Letter from Long Island Lighting Company transmitting TDI Diesel Generator Rocker Arm Capscrew Stress Analysis Report.

Shearon Harris SSER 1

Appendix A

March 16, 1984 Letter from Transamerica Delaval responding to requests for information on R-4 series engines.

March 16, 1984 Letter from applicant requesting amendment to Construction Permit No.

CPPR-158 to extend construction completion date to March 1, 1986.

March 21, 1984 Meeting with applicant to discuss equipment qualifi-cation program, including both environmental and seismic considerations.

March 22, 1984 Letter from Transamerica Delaval forwarding copies of test results on first engine produced in each model line.

March 22, 1984 Letter from applicant forwarding information on monitoring of service water system.

March 23, 1984 Letter from Long Island Lighting Company transmitting TDI. Diesel, Generator Air Start Valve Capscrew Dimensional and Stress Analysis Report.

March 29, 1984 Letter from applicant transmitting Amendment No.

12 to FSAR.

March 30, 1984 Letter from Long Island Lighting Company transmitting TDI Diesel Generator Cylinder Head Stud Stress Analysis Report.

April 4, 1984 Meeting with applicant to discuss design of the retaining wall with the staff and the applicant.

April 4, 1984 April 9, 1984 April 9, 1984 Board Notification 84-072 -- Transamerica Delaval, Inc.

(TDI) Owners Group/NRC Meeting Transcript and Additional TDI Owners Group Information Submitted.

Letter from Long Island Lighting Company providing additional information concerning Owners Group Report on Types AF and AE Piston Skirts.

Letter from applicant forwarding revised response to Draft SER Open Item 136 concerning loss of component cooling water.

April 10, 1984 Letter from applicant regarding the requested con-struction permit extension.

April ll, 1984 Letter to J.

P.

McGaughy, Mississippi Power 8 Light, regarding preliminary assessment of two reports sub-mitted to the NRC by the Transamerica

Delaval, Inc.

(TDI) Owners Group.

Shearon Harris SSER 1

Appendix A

April 11, 1984 April 12, 1984 April 13, 1984 April 13, 1984 April 13, 1984 April 16, 1984 April 18, 1984 April 19, 1984 Letter to applicant forwarding request for additional information.

Letter from Long Island Lighting Company forwarding fatigue data for modular cast iron.

Letter from applicant forwarding seismic report for first quarter of 1984.

Letter from Long Island Lighting Company forwarding TDI Diesel Generator Supplement to Cylinder Head Stud Stress Analysis and Supplement to the Air Start Valve Capscrew Dimension and Stress Analysis.

Letter from Long Island Lighting Company forwarding calculations related to Rocker Arm Capscrew and Cylinder Mead Stud Reports.

Letter from Long Island, Lighting Company forwarding TDI Diesel Generator Report on Engine Driven Jacket Mater Pump Design Review.

Letter to applicant forwarding request for additional information.

Letter from Long Island Lighting Company forwarding copies of TDI Diesel Generator Report on Push Rods.

Shearon Harris SSER 1

Appendix A

APPENDIX B BIBLIOGRAPHY Carolina Power and Light Co.,

"Human Factors Design Report for Shearon-Harris Unit 1 Control Room," January 23, 1981; revised September 16, 1981, and April 14, 1983.

---, letter LAP-83-156, to H.

R.

Denton (NRC),

"Human Factors Design Evaluation Report for the Shearon-Harris Unit 1," June 1, 1983.

letter LAP-83-426, to H.

R.

Denton (NRC), "Supplement 1 to NUREG-0737, Detailed Control Room Design Review," September 27, 1983.

U.S.

Nuclear Regulatory Commission, NUREG-0700, "Guidelines for Control Room Design Reviews,"

September 1981.

NUREG-0737, Supplement 1, "Requirements for Emergency

Response

Capability" (Generic Letter No. 82-33),

December 17, 1982.

---, NUREG-0801, "Evaluation Criteria for Detailed Control Room Design Reviews,"

draft for comment, October 1981.

NUREG-1580, "Human Engineering Guide to Control Room Evaluation," July 1980.

Shearon Haarris SSER 1

Appendix B

APPENDIX D

ACRONYMS AND INITIALISMS AFW BOP CRT auxiliary feedwater balance of plant cathode ray tube EOP

'RFIS ERG ESCWS emergency operating procedures emergency response factors information system emergency response guidelines essential services chilled water system FRP FSAR functional restoration procedures Final Safety Analysis Report ICC ICCI inadequate core cooling inadequate core cooling instrumentation LMFW LOAC LOOP loss of main feedwater loss of alternating current loss of offsite power MCB main control board NSSS nuclear steam supply system PWR pressurized water reactor RVLIS reactor vessel level instrumentation system SER SMM SPDS safety evaluation report subcooling margin monitor safety parameter display system WOG Westinghouse Owners Group WPBCWS waste processing building cooling water system Shearon Harris SSER 1

Appendix D

APPENDIX E

PRINCIPAL STAFF CONTRIBUTORS This supplement is a product of the NRC staff.

The staff members listed below were principal contributors to this report.

Name Title Review Grou Marthe E. Harwell Section Chief Ray Ramirez Senior Human Factors Engineer George A. Schwenk Senior Nuclear Engineer Editorial Section Human Factors Branch Core Performance Gerald E.

Simonds Emergency Preparedness Specialist Emergency Preparedness Norman H. Wagner Mechanical Engineer Auxiliary Systems Shearon Harris SSER 1

Appendix E

APPENDIX G ACRS REPORT Shearon Harris SSER 1

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UNITED STATES NUCLEAR REGULATORY COMM(SSION ADVISORY COMMITTEE ON REACTOR SAF EGUA'RDS WASHINGTON, D. C. 20555 January 16, l984 Honorable Nunzfo J. Palladino

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

20555

SUBJECT:

ACRS REPORT ON THE SHEARON HARRIS NUCLEAR POWER PLANT During its 285th

meeting, January 12-14,

1984, the Advisory Committee on Reactor Safeguards reviewed the application of Carolina Power 8 Light Com-pany (CPAL) and the North Carolina Eastern Municipal Power Agency (the Applicants) for an operating license for the Shearon Harris Nuclear Power Plant.

The Shearon Harris Nuclear Power Plant will be operated by CP&L which also operates three other nuclear units.

The project was considered during an ACRS Subcommittee meeting in Apex, North Carolina on January 3-4, 1984.

Members of the Subcommittee toured the facility on January 3,

1984.

During its review, the Committee had the benefit of discussions with repre-sentatives and consultants of the Applicants, Westinghouse Electric Corpor-

ation, Ebasco

Services, Inc., the NRC Staff, and a

member of the public.

The Committee also had the benefit of the documents referenced.

The Commit-tee commented on the application for a permit to construct the Shearon Harris Plant fn reports dated March '8,

1972, January 17,

1973, and August 19, 1977.

On October ll, 1977 the Committee provided a

response to an inquiry regarding the resolution of ACRS Generic Items related to the Shearon Harris Nuclear Power Plant.

The Shearon Harris Nuclear Power Plant is located in Wake

County, North

Carolina, approximately 16 miles southwest of-the nearest boundary of

Raleigh, North Carolina.

Originally the Shearon Harris Nuclear Power Plant was to comprise four units.

However, only Unit 1 will be completed, with an estimated fuel load date of June 1985.

Units 3

and 4 were cancelled on December 18, 1981 and Unit 2 was cancelled on December 21, 1983.

The Shearon Harris Nuclear Power Plant uses a three-loop Westinghouse nuclear steam supply system with a rated core power of 2775 MWt.

The containment fs a large, dry, reinforced concrete structure.

During the Committee's consideration of this plant, the control room design was reviewed.

The Applicants informed us that they intend to perform an operational test of the control room emergency air recirculation system.

As a part of this exercise, control room habitability during the recirculation mode will be evaluated.

We wish to be kept informed.

Shearon Harris SSER 1

Appendix G

Honorable Nunzio J. Palladino 2

January 16, 1984 The Shearon Harris Nuclear Power 'Plant uses Westinghouse D-4 steam genera-tors.

Steam generators of this design have experienced tube degradation related to flow-induced vibrations in the preheater region.

Internal modifi-cations have been developed by Westinghouse which include expanding some steam generator tubes and directing some of the main feedwater flow through the auxiliary feedwa:er nozzle.

We expect to be kept informed regarding the operating experience of these steam generators.

The NRC Staff has previously identified management deficiences in CP5L's nuclear program.

These deficiencies are enumerated in the report (May 1983) of the most recent Systematic Assessment of Licensee Performance (SALP) con-ducted by the NRC Staff to assess CP&L's nuclear operations for the period January 1982 - January 1983.

CPSL has taken measures to improve management function and capability.

These include restructuring of the corporate or-ganization which will eventually result in a consolidation of CPEL's nuclear organization under one senior manager.

The restructuring also provides for a corporate level executive to be located onsite, as a member of involved site management, to ensure greater access to resource=

and to enhance the ability to'initiate new programs from the site.

These efforts are expected to cor-rect the past deficiencies.

Members of the Region II Staff reported orally during the meeting that significant improvement in performance has been observed since the last SALP inspection.

The Committee believes that written evidence of an improvement in CPEL's nuclear operations, which could, for

example,

.be reported in the two scheduled SALP reviews prior to fuel load should be available prior to full power operation.

We wish to be kept informed.

Subsequent to the meeting with the Applicants, we have received a letter from a

member of the public which makes several allegations concerning quality assurance and other issues.

We request that the NRC Staff investigate these allegations and provide a written report to the Committee.

The ACRS has on several occasions recommended that evaluations be made of the capability of light water nuclear power plants to be shut down safely in the event of an earthquake of greater severity and lower likelihood than the safe shutdown earthquake.

In a letter dated January ll, 1983, the ACRS made rec-ommendations concerning a possible broad approach to deal generically with the question of seismic margins.

In the meantime, for the Shearon Harris Nu-clear Powe'r

Plant, we recomnend

that, in addition to items already con-

sidered, specific attention be given to assurance of adequate seismic capa-bility of the emergency AC power supplies, the DC power supplies, and small components such as actuators and instrument lines that are important to the accomplishment of safe shutdown and decay heat removal.

We suggest also that specific attention be given to the adequacy of clearances between adjacent buildings.

During this review there was discussion of the reliability and the fracture resistance of the chilled water system.

The Applicants and the NRC Staff Shearon Harris SSER 1

Appendix G

Honorable Nunzio J. Palladino 3

w January 16, 1984 reported orally that the system is satisfactory in these respects.

The ACRS would like to receive a detailed discussion of the chilled water system in a supplement to the Safety Evaluation Report.

One of the confirmatory issues concerning this application is "turbine missiles."

Because of the nonoptimum orientation of the turbine relative to vital components in this plant, we recommend that a structured test program for evaluating overspeed protection of the turbine be prepared and submitted to the NRC Staff for review and approval before full power operation.

A number of items have been identified by the NRC Staff as Outstanding Issues.

There is also a set of Confirmatory Issues that awaits additional documentation.

We found no reason to believe that any of these issues will be especially difficult to resolve.

We recommend that they be resolved in a manner satisfactory to the NRC Staff.

The ACRS believes that, if gue regard is given to the items mentioned

above, and subject to satisfactory completion of construction,

staffing, and preoperational

testing, there is reasonable assurance that the Shearon Harris Nuclear Power Plant can be operated at core power levels up to 2775 Nt without undue risk to'he health and safety of the public.

Sincerely, Jesse C. Ebersole Chairman

References:

1. Carolina Power 5 Light Company, "Shearon Harris Nuclear Power Plant Units 1, 2, 3,

and 4, Final Safety Analysis Report,"

Volumes 1-20 and Amend-ments 1-10

2. U. S. Nuclear Regulatory Commission, "Safety Evaluation Report Related to the Operation of Shearon Harris Nuclear Power Plant, Units 1

and 2,"

USNRC Report NUREG-1038, dated November 1983

3. Letter from Wells Eddleman, Intervenor,

Subject:

Comnents on the Shearon Harris Nuclear Power Plant to the Advisory Committee on Reactor Safeguards, dated January 13, 1984 Shearon Harris SSER 1

Appendix G

wo The Advisory Committee on Reactor Safeguards 1.13.84 by Wells Eddleman, intervenor Many things can oompromise nuclear plant safety.

One is plant or corporate management.

Another is, in pfark Twain's words, "what, you know, that ain't so".

In 1979 (construction permit remand hearing on management capability, NRC Docket 50<<400),

the NRC staff "knew" that although CP8cL had had problems at its Brunswick plant, e.g.

staff turnover, lack of supervisory attention, so many problems that they couldn't all be given attention, large numbers of LZRb and long times to repair, things were going to be OK from now on because OPAL had straightened, out.

Although the Atomic Safety

and, Licensing Board in that'emand, expressed some doubts about some CPkL testimony and actions, it bought that judgment overall".

Prom 1979-83, Brunswick has continued to have high employee

turnover, numerous

problems, large numbers of LERs, long times to repair, etc.

In addition, numerous meltdown precursors at Brunswick were identified including the loss of 3 out of 4 RHR heat exchangers due to biofouling in 1981.

Most recently, CPS'efused to shut Brunswick down to check for pipe cracks in the summer of 1983.

On shutdown, numerous cracks were found up to 18" in length, and overlay welds (a "band-aid") were used. to put a unit back into service.

The better operating record OPAL 'claims for Brunswick recently does include the longest run without either unit shutting down, in the entire history of the plant:

6 weeks in fall 1983.

Brmaswick 1 and 2 have respectively the 2d worst and. worst design electrical rating capacity factors (DER CP) of any 2 BVYRs in the USA, below even that of Browns Perry 1 which was disabled by, fire in 1975.

OPAL has lots of charts and lists about management, but one test is results, and their results are poor enough tnat some explanation should be sough beyond NRC Staff's reassurances

~

One indication that' fairly obvious is that between the ACRS subcommittee meetings Jsn 3-4 and the Jan 12 presentation, the CBRL presentations were not updated gPmi directly.

Statistics onpperson years of experience which the Sub-,

committee folk had called "meaningless" were still included.

You can't expect the Company to do other than try to look good, but safety requires some deeper examination.

CMP, for example, who audited OPAL management and gave a rather favorable report, Shearon Harris SSER 1

Appendix G

had adopted "unusually effective approaches" to the construction of its Seabrook pl@at:

That plant is nor recognized as a model of waste in construction, using aearly 8000 workers to try to build a 2-unit plant.

The CMP auditors relied mostly on CP8cL-supplied statistics in auditing CP8:L, and PSNH may have fooled them about Seabrook.

As a C~olina consumer said, the CtdP report shows CPM "is well managed in all areas except results".

The CMP report also fails to assign responsibility for CP8Q's repeated foulups at Brunswick.

Brunswick was rushed. oa-line in 1974 for financial reasons.

J.A. Jones, a senior CP8:L executive in 1979, testified ia the CP remand that CP8:L made an all-out effort to get Brunswick 2 running by 12-31-74 in order to get tax benefits.

(You know TXI-2 was handled similarly).

Shearon Harris is now being "rushed",

3 shifts if necessary, to meet a 1986 commercial operavioa date, tho NRC Staff believed in a recent check of the schedule that they would be 6 months behind. schedule in going on-line.

The Harris pipe hanger situation does not reflect pd,11 oa CP8cL.

In September

1980, NRC inspector Maxvrell noticed a bad pipe hanger.

A sample of 400 were then reinspecteL (all having been previously approved),

and 95$ 'were found defective.

Then about 1700 pipe haagers were reinspected and of these, about 1/3 were founL OK, tho numbers of tiiem hsd, waivers (I have reviewed. thousands of pages of CPS'ipe hanger iaspectivn reports witnout yet finding a request for field change or waivers that was rejected).

About 1/3 were also found defective.

The remainiag fraction were such that from the blueprints (due to errors or lack of clarity) it could not be told whether the hangers were yelde4. to specification or not.

These weld designs are

made, as I understand it, by Bergen-

Paterson, reviewdd by Zbasco, then reviewed by CP8cL, then sent to Daniel International and welded.

At each of these

stages, unclear symbols should have been re)ected>

but were not for hundreds of pipe hangers that were welded (aad approved by inspectors) prior to this reinspection.

More pipe hanger problems aad support problems have been identified and CPEcL aow anticipates finishing its 10'einspection by October 1984, i.e, more tnan 4 years after the first problem found by the NRC inspector.

At Apex, the NRC Staff told the Subcommittee that although CP8t:L had numerous QA/qC violaticns, they had no safety significance because such violations happen at other plants.

Zimmer'?

Wolf Creek't Shearon Harris SSER 1

Appendix G

The unfortunate Qistory of NRC Staff coverups of defects at plants like Zimmer doesn't just undermine those NRC staffers who are dedicated to safety and quality.

Xt also means that it is not safe to rely on staff reassurances, expecially if safety significance is determined by whether other plants also have such violations, in which case there is no such safety significance according to NRC Staff. I'd like to see them try tnat argument on a tr~fic cop who pulled one of them for speeding.

There are about 12000 pipe hangers in the Harris design.

Hot all appear to be in place.

When COL retrained their hanger

welders, over 10$ failed, a test immediately after the retraining, by welding a defective hanger yet again.

NRC Staff has only inspected about 50 pipe hangers at Harris, according to discovery documents.

JYbre information is surely needed on tnis issue to sustain a recommendation that the plant oe found safe.

Zf b"ilt by idiots, no ingenious design or operation can save a plant from trouble.

Pipe hangers are very important to plant safe shutdown in earthquakes and various transients.

They snou1d. not be left to such sloppy practice as COL has evidenced to date.

Such w3.de-ranging QA failures as CP8cL has had, on pipe hangers indicate a breakdown of QA, and voluntarily stopping while they try to fix it is about as meritorious as stopping one's car for repairs after a wheel falls off.

There are other Harris problems.

Por example, the "fix"for its defective steam generators doesn't just include expanding lots of steam generator tuoes (which, by NRC rule, are not allowed to leak except for one rupture per, stem generator or low volume leakage.

Those tubes at Ginna who violated tnis rule have not been prosecuted to my knowledge). lt also involve a jury-rig feedwater arrangement, about 18'nrough the AZW nozzle, permanently.

The safety significance of events like the hiaine Yankee water hmmer of early 1983 on such jury-rigged arrangements does not appear to have been adequately analyzed.

COL's."plan" to find voids in pipes where water hmmers could start consists of having weQ3cth&ugh people look for leaks.

What x-ray vision they may use to see inside pipes is not clear.

Nor does CP&L appear tn have identified wnich piping and valves shoed be checked, for leaks because of possible water hemmer events those leaks might indicate are possible (e.g.

due to voids in pipes).

Shearon Harris SSER 1

Appendix G

OPAL's track record of being wrong xs weA.J. x~usrrarea oy the 3 holes deep down to bedrock that surround Harris ti-the sites of cancelled units 2,3 and, 4.

There's a fault under the Unit 1 containment indctive for 2.5 million years or more, CBcL's consultant

says, but it still weakens the rock there.

HRC seismick analysis apoears to be for acceleration only, not for fissures under the plant or parts of the plant mov'ng 2 ways at once.

And tho the unit 3 and 4 holes (and perhaps unit 2's will also) be filled in with dirt, they still areztt as strong as rock or the basemats and.buildings planned for those locations.

Earthquakes are also a very logical source of common-mode failure of sll 7 transmission lines to Harris, leading to total loss of offsite power (LOOP).

Terrorist attack a huge ice storm, a wind storm or hurricane can also lead to such coze on mode failures of these lines.

COL only gives a probability for tornado-caused common mode failure.

Fihat about an earthquake-induced crack in the cooling water supply lines?

='ven were the plant shut down, it still would need to be cooled, and without offsite power and w"ter that would be very difficult.

The spent fuel pool, lying on a ridge between the 2 holes farthest from Unit 1, and along a "ledge" by unit 1 to another hole, is in shaky shape even if the holes are filled with soil.

Has CP8cL or HRC staf adequately addressed the possibility of the pool cracking in an earthquake, losing cooling water due to cracked piping or loss of 'source in a quake, and so perhaps boiling dry?

hhat about cri icali.y as fuel bundlexs swing in a quake?

The Diablo Canyon experience shows that undiscovered geologicai feature or design errors can greatly raise the actual risk from earthquakes compared to initial analysis.

Concerning security, the essential water chillers need to be protected.

I won't say more on this, not wanting to give any terrorists advice.

I do note that we intervenors instructed our experts in security who reviewed the Harris security plan to only make a'contention if a serious problem in the plan were found.

Minor matters were to be suggestiors for ir.provement only.

I understand that 6 security contenticns were filed in secret f and that a list of auggesticns was also given to CPRL, which re jected all of them, perhaps not noting that they were suggestiors.

This situation deserves a c'oser look by t:he ACBS.

Shearon Harris SSER 1

Appendix G

CPM has been fined repeatedly by NRC for poor practices.

'Their Brunswwick plant has drawn very low SALP ratings.

Indeed, at the time of the T3.'I-2 accident, Ty3I was rated higher in virtually every area evaluated by NRC, than was Brunswick.

This says 2 things:

don't trust a high rating, and look into Brunswick.

(Data in Board Exhibit-8 of the 1979 remand on CP8cL management capability.)

CPRL has dragged their feet on fire protection at Brunswick

~d taken 16 months to put an indicator light from a vital water-tight door protecting ZCCS components from flooding, to the control room at Brunswick.

The Robinson plant has a very low "site stringency" or level of safety requirements (Bd. Zxh. 8, op cit) and thus isn' such a good performer as its record might seem to indicate it has looser requirements which lead to fewer violations.

Good managemen

'PAL's control room design leaves much to be desired at Harris.

/he moved cabinets actually are in each other's way (there is a contention admitted on this),

anL the SPDS is inadequate (content'cns have bean filedon this and CP&L h~s been ordered to produce more SPDS information to me in Pebruary.

Sp far, C2&L has not been forthcoming with information like its detaileL Human Engineering Requirements Specifications for lights and noise.

One SPDS analysis OPAL has submitted has the RCS light GREEN ("safe") during a large LOCA, wnich doesn't apoear very bright at all.

The COL management capability contention will go to trial in late summer or early fall as now scheduled.

The emergency plan is expected out in mid-Pebruary (delayed fro-Decemoer) end will likely be the subject of numerous contentions.

All these matters (snd those above) merit review. I hope this little information is of some use to you+ though i.t covers only a few of the issues the ACRS must Leal with about Shearon Harris l.

In closing, I'd like to thank the AORS for its review of Harris.

ACRS has the repMation of being the boost knowledgeable group associated with the NRC, but of being rather shy to state problems specifically in its letters on plants.

We members of the public depend on you as a line of defense of our safety.

Though moat of you may be "1000 miles away" like the proverbial nuclear construction worker when and if Harr's 1 starts up, many people will be closer and will have to live with the plant and C2&L!s running of it.

Xour investigatic and your comments can be of great help to these people.

I believe CPEcL went through its whole presentation with but minor mention of people around the plant.

Those people are real, end they depend on the ACPS to c~h problems CBcL and NRC Staff do not.,;:"~~i"~elle Zdd]caen Shearon Harris SSER 1

Appendix G

6 F,

v I NRC FORM 33S 16 631 3 TITLE AND SUBTITLE V S NUCLEAR REGVLATORY COMMISSION BIBLIOGRAPHIC DATA SHEET REPORT NUMBER lAIRPnhd ot TIOC, hdd Voi IYo, ih hnhl NUREG-1038 Supplement:

No.

1 2 Lthvt bihn4 4

RECIPIENT'S ACCESSION NVMBER Safety Evaluation Report Related to the Operation of Shearon Harris Nuclear Power Plant, Unit 1

6. AVTHORISI 5 DATE REPORT COMPLETED MONTH June 2 DATE REPORT ISSUED MONTH June YEAR 1984 YEAR 1984 6 PERFORMING ORGANIZATIONNAME AND MAILINGADDRESS llntftdh Zvo Cochl Division of Licensing Office of Nuclear Reactor Regulation U.S ~ Nuclear Regulatory Commission Washington, D.C, 20555

9. PROJECTITASKIWORK VNIT NUMBER 10 FIN NUMBER I I SPONSORING ORGANIZATION NAME AND MAILINGADDRESS llnhivCC Zto CoCkl 124 TYPE OF REPORT Same as 8. above Safety Evaluation Report 126 PERIOD COVEREO llnhlvhvt dhihll 13 SVPPLEMENTARY NOTES Docket No. 50-400 I~ ABSTRACT QGG vvohdh oh ICIII Supplement No. 1 to-the Safety Evaluation Report for the application filed by Carolina Power and Light Company and North Carolina Eastern Municipal'ower Agency for a license to operate the Shearon Harris Nuclear Power Plant, Unit 1 (Docket No. 50-400), located in Wake and Chatham Counties, North Carolina, has been prepared by the Office of Nuclear Reactor Regulation of the U;S. Nuclear Regulatory Commission.

This supplement provides mor'e recent information regarding resolution of some of the open items identified in the Safety Evaluation Report.

This supplement also provides and discusses the recommenda-tions of the Advisory Committee on Reactor Safeguards in its report on Shearon Harris, dated January 16; 1984.

154 KEY WORDS AND DOCVMENT ANALYSIS ISO, OESCRIPTORS 16 AVAILABII.ITYSTATEMENT Unlimited 11 SECURITY CLASSIFICATION IThilItpohll 16 NUMBER OF PAGES Unclassified 19 SECURITY CLASSIFICATION llhih IMPtl Unclassified 20 PRICE.

46,6, GOTBBIGIEBT PRIIITIBG OPTICS I 1966 ~21-297/4621