ML20212N250
| ML20212N250 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 08/22/1986 |
| From: | George Thomas PUBLIC SERVICE CO. OF NEW HAMPSHIRE |
| To: | SEACOAST ANTI-POLLUTION LEAGUE |
| References | |
| CON-#386-497 OL-1, NUDOCS 8608280141 | |
| Download: ML20212N250 (56) | |
Text
7 n-r-
v-WND'90tutmny DOCKETED USNRC Dated:
August 22, 1986 116 AUS 27 20:39 l
UNITED STATES OF AMERICA 0FFICE OF it s.t iAny NUCLEAR REGULATORY COMMISSION 00CKETlHG A SEPVICf.
I BRANCH before the ATOMIC SAFETY AND LICENSING BOARD
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In the Matter of
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PUBLIC SERVICE COMPANY OF
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Docket Nos. 50-443-OL~7 NEW HAMPSHIRE, et al.
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50-444-OL--/
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'On-site Emergency Planhing (Seabrook Station, Units 1 and 2) )
and Safety Issues
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APPLICANTS' ANSWERS TO SEACOAST ANTI-POLLUTION LEAGUE'S INTERROGATORIES AND REQUESTS FOR DOCUMENTS TO THE APPLICANTS Interrogatories
.SAPL Supplemental Contention 6 (Formerly 'JH-10) 1)
List all documentary,or other materials the Applicants may employ in this proceeding to support its position (s)
With respect to this contention.
In addition to listing such documents and other materials,_ provide a copy of all of them pursuant to 10 C.F.R. $ 2.741.
The listing of documents and materials which may be employed in this proceeding to support our position with respect to this contention are provided in Exhibit 1-A.
Those documents not specifically provided with each response are available for your review at Seabrook Station or at the s
960828C141 860322 PDR ADOCK 05000443 G
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offices of Yankee Atomic Electric Co. (YAEC) in Framingham, MA.
To arrange inspection of these and all other documents referenced herein call W.J. Daley, Jr. at (603) 474-9521, Ext. 2057.
2)
State the names and provide the curriculum vita (e) of any person or p rsons relied upon to substantiate in whole or in part the Applicants' position (s) with respect to this contention.
The names of those persons relied upon to substantiate in whole or in part our position with respect to this conten. tion are:
T.B.
- Sheridan, E.A.
- Sawyer, W.G. Alcusky, J.L.
- Peterson, L.A. Walsh.
The associated curriculum vita (e) for these individuals are provided in Exhibit 1-B.
3)
Identify any person or persons the Applicants may call as a witness on this contention, and, if the information has.not been provided in respense to question 2,' provide curriculun vita (e) of said person or persons.
In addition to those persons identified in response to
,e may call G.S. Thomas as a witness.
question-(2) above, w
The curriculum vita (e) of this individual is provided in FSAR Appendix 13A.
4)
Provide a summarization of the proposed testimony, views or positions of all persons named in response to interrogatories (2) and (3) above that may be presented by the Applicants in this proceeding.
A summarization of the proposed testimony, views or positions of all persons named in response to interrogatories (2) and (3) above is provided in Exhibit 1-C.
r 5)
State the specific basis and references to documents which the persons named in response to interrogatories
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(2) and (3) above may rely upon or reference regarding this contention.
See Responses to Interrogaries Nos. 1 and 4.
6)
State with specificity the reasons why the Applicants believe.that stack monitor and steam generator (or steam line) radiation need not be added to the Safety Parameter Display System (SPDS) until prior to restart following the first refueling outage.
Steam generator and plant vent stack radiation will be included on the SPDS prior to exceeding 5% power.
These variables are not required prior to exceedi.ng 5% power since the radioactive source generated by the limited operation at low power cannot result in significant radioactive releases that would require offsite response.
These parameters are
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monitored in the control room.
See FSAR,Section 7.5 for details.
7)
Do the Applicants hold that the period of operation prior to the first refuelang outage is any_ safer than any other period of operation, and if so, upon what basis or bases?
No, we do not hold that the period of operatien prior to the first refueling out. age is safer than any othe period of I
operation.
8)
For the following parameter displays A through E, state how long it would take to install the parameter display,
.the cost, any technical problems that might be encountered in its instsllation, and Applicants' full justification for not having installed these parameter displays on the SPDS console to date:
a)
RHR flow b)
Containment Isolation c)
Containment Hydrogen Concentration d)
Steam Generator (or steam line) Radiation m
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Stack Monitor It would take approximately three months for design and installation.
The cost would he less than $200,000 and the only technical problems that might be incurred would be computer Input / Output availability for valve position.
As to justification of items a) - c) see SBN-987.
(Attachment 1, page 9); as to items d) and e) see answer to No. 6 above.
i 9)
Have Applicants'provided for operator protective equipment to be kept in the control room?
If not, state why not.
Is such protective equipment to be kept in any other location?
If so, state where.
We have not provided operator protective equipment to be kept in the control room or other locations for use by the control room operators.
Control Room habitab'ility is maintained under all operating conditions by the ventilation system.
For details, see FSAR Section 9.4.1.
- 10) List all actions or requirements for licensing that the Applicants are deferring until the first refueling outage.
All actions or requirements for licensing that have been deferred until the first refueling outage with respect to the performance of the DCRDR are identified in Exhibit 1-D.
- 11) Have Applicants performed a complete system functicn and task analysis and subsequent comparison of results of the analysis with the control room inventory as is required by NUREG-0737, Supplement 1?
Yes, such an analysis and comparison has been. performed.
- 12) Why, when in SBN-839 Applicants committed to review the control room furnishings for HEDs "at least 120 days
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PTLF", are Applicants now seeking to defer this action until prior to startup from the first refueling outage?
A final review of the control room furnishings could not be completed "at least 120 days PTLF" because all the e
control room furniture was not in the control room at that time.
.A final review has now been performed, some mirlor
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HEDs developed and resolutions, recommended.
These resolutions will be implemented before initial criticality.
- 13) Why, when in SBN-839 Applicants said that operator protection equipment and emergency equipment storage would be reviewed at'least 120 days PTLF and any HEDs submitted for NRC review, are Applicants se6 king to defer this action until prior to startup from the first refueling outage.
A final' review has been performed, some minor HEDs have been developed, resolutions recommended, and resolutions will be implemented before initial criticality.
_14) Why are Applicants unable to complete banding of indicutors prior to fuel load?
In a dynamic system, there is usually some slight difference between design values of operating parameters and the actual values determined once the system is operating.
We believe it is desirable to obtain the actual running values for the parameters before banding the indicators.
It is, therefore, necessary for the plant to have run at full power before these values can be accurately banded on the indicators.
- 15) State why Applicants have not planncd to correct
. color-related HEDs in the, Video Alarm System prior to fuel load and have instead sought to low power test the plant before making the correction.
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The colcr-related HEDs in the Video Alarm Systems will be c'orrected' prior to initial criticality.
- 16) Are reactor coolant system vents able to be remotely, operated from the control room as NUREG-0737 II.B.1 requires?
Provide a description of the related displays and controls.
Yes.
A description of reactor coolant systems vents and compliance with NUREG-0737, II.B.1 is discussed in FSAR Eections 1.9, 5.2.2.8, and 5.2.6.
- 17) Is there positive indication of reactor coolant system relief and safety valves in the control room as NUPEG-0737 II. D.3 requires?
Provide a description displays and controls.
Yes.
A description of reactor coolant system relief and safety valve indication and compliance with NUREG-0737, II.D.3 is discussed in FSAR -Sections 1.9, 5.2.2.8, and Table 7.5-1, Item D.11.
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- 18) Are the additional accident monitoring instrumentation
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and assocated displays and controls added to the control room as required by NUREG-0737 II F.1?
j Yes.
Compliance with NUREG-0737, II.F.1 is discussed in FSAR Sections 1.9, 6.2.2.5, 6.2.5, Table 7.5-1, 11.5.2.1.j, 11.5.2.1.k, 12.3.4, and 12.5.3.
- 19) Describe the, types.and locations of displays and alarms to be added.to the control roon that are related to the instrumentation'for detection of inade'qute core cooling required by NUREG-0737 II.
F.2.
t See FSAR Subsection 4.4.6.5 and Table 7.5-1.
In l
addition, a complete description of Seabrook's l
' Instrumentation for Detection of Inadequate Core Cooling" was submitted to the NRC via PSNH Letter (SBN-952), dated February 24, 1986.
- l
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- 20) Describe the Applicants' incorporation of the lessons learned from the Salem ATWS event in the Detailed-Control Room Design Review.
As part.of the DCRDR, the adequacy of the displays and controls to support the ATWS Procedure were reviewed.
No HEDs were found.
21') Provide a. description of how the Applicants have assured the following relative to the SPDS:
a) ~ that appropriate parameters are displayed b) that it is isolated from safety systems c) that it vill provide reliable and valid data d) that it incorporates good human engineering practice a) phe bases for parameter selection is presented in SBN-920 (Attachment 1, Section 4) as modified bf SBN-987'(Attachment 1, page 9).
b)
Isolation from safety systems is provided by qualified isolators.
c)
Data validation and reliability are discussed in SBN-987 (Attachment 1, pages 3 through 8).
i d).
Human engineering practices are discussed in SBN-987 (Nttachment 1, pages 2 and 3).
- 22) Provide the conclusions of the SFDS Verification and Validation (V&V) Program.
The test results of the V&V program demonstrated the effectiveness of man / machine interface in addressing and mitigating emergency conditions in the plant.
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The SPDS critical safety function status trees represent a precise and consistent mechanism to monitor the plant status.'
They are suitable for computer based or manual i
evaluation.
Implementation i,s straightforward.and well understood by the operators.
23 ) Provide a diagram of the control rpom displaying the location of the SPDS console relative to the Main Control Board with distance (s) indicated in appropriate units of measure.
Also indicate the dimens;.ons of the i
SPDS console (height, etc').
See Exhibit 1-E.
4 NECNP CONTENTION I.B.2 1)
List all documentar y or otlier mEterials the Applicants may employ in this proceeding to support its position (s) with respect to this contention.
In addition to listing such documents and other materials, provide a copy of nll of them pursuant to 10 C.F.R. 5 2.741.
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The documents ietied upon in rocponse to interrogatories are referenced with each interrogatory response.
Tho'se documents not spec ifically provided with each response are available for your review at seabrook Station or at the
' office of United Engineers & Constructors (UE&C) in Philadelphia, PA.
2)
State the names and provide the curriculum vita (e) of any person or persons relied upon to substantiate in i
whole or in part the Applicants' position (s) with res.oect to this contention.
l The names of those persons relied upon to substantiate, in whole or in part, our posicion with respect to this 4
contention are R.
- Bergeron, W.J.
- Cloutier, J.M.
- Salvo, G.
Tsouduros, and N.K. Woodward.
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e The associated curriculum vita (e) of these individuals are provided in Exhibit 2-A.
3)
Identify:any person or persons the Applicatns may call as a witness on this contention, and, if the information han not been provided in response to question 2, provide citrriculum vita (e i of said person or persons.
The names of those persons who may be called as a witness during these proceedings regarding NECNP Contention I.B.2 are J.M.
- Salvo, G.S.
Thomas, and N.K. Woodward.
The curriculum vita (e),of G.S.
Thomas is provided by FSAR Appendix 13A.
4)
Provide a sumnarization of the propoved testimony, views or positions of all persons named in respense to interrogatories (2) and (3) above tha,t may be presented by the Applicants in this proceedings.
a)
The testimony will explain what environmental qualification is and why it is being performed.
A I
suntuary cf NHY's standard (post-accident operability time) which was used at the time of the 1983 testimony, will be presented.
A discussion on t
the present 100 day accident duration and its applicability at Seabrook will also be presented.
All equipment which does not meet the 100 day requiremer.t, will be identified with a technical justification, which demonst. rates that the equipment performs its safety function well within the equipment's qualified duration.
Responses to all concerns raised by NECNP in their submittal, dated July 2, 1986 regarding environmental
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S qualification will be presented.
We will concludo that all equipment important to safety complies
.with 10 CFR 50 Appendix A GDC-4 and 10 CFR 50.49.
b)
In addition to those docu'ments relied upon to
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respond to these interrogatories and any documents specified in the interrogatories the Applicants may make use of,thone documents identified in Exhibit 2-B.
5)
State.the specific bases and references to documents which the peraons named in response to interrogatories (2) and (3) above may rely upon or reference regarding this contention.
See Responses to Interrogatory Nos. 1 and 4.
6)
Provide the analysis supporting conformance 5;ith Reg.
Guide 1.75.
USNRC Regulatory Guide 1.75 endorses with certain clarification IEEE Standard -384-1974 which sets forth the criteria utilized in the Seabrook design for separation of Class 1E equipment and circuits.
The overall compliance to Regulatory Guide 1.75 has been addressed in FSAR Chapter E and has been reviewed and accepted by the NRC.
The reference in SSER-5 Section 3.11.3.1 at 3-17 to submittal of analysis supporting conformance to Regulatory Guide 1.75 pertains to supplemental information for certain specific electrical separation configurations.
The Applicants have submitted the supplemental information to the NRC (Reference SBN-979 and SBN-1107).
The NRC has
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indicated the acceptability of the analysis in SSER-5 Soction 8.3.3.3 at 8-2.
4-7)
For the following, state how long it-would take to -
install the environmentally qualified equipment, the cost, any technical probl' ems tnat might be encountered
- in installation and Applicants' full justification for not having installed the equipment to date:
a)
Monitor for containment sump water temperature b)
Moni or for accumulator tank pressure c)
Monitor for accumulator' tank level d)
Monitor f.or primary coolant radiation level e)
Monitor for narrow range sump water level Justification for not having installed environmentally qualified indication for those parameters specified in the interrogatory is pr.ovided in the Seabrook FSAR Section 7.5, Appendix 7A.
In discussions uith the NRC Staff concerning Regulatory Guide 1.97 deviations, the Staff indicated that deviations on accumulator pressure or level and containment sump water temperature were unacceptable.
The applicant, therefore, has committed to install and have operational, prior to the restart after the first refueling, environmentally qualified accident monitoring equipment for safety injection j'
accumulator tank level, pressure and containment sump water temperature.
The acceptability of this commitment 1s reflected in SSER r Section 3.11.3.1 at 3-19.
The' Staff evaluati'on in SSEF-5 concluded that sufficient just fication was provided for not having environmentally qualified i
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parameters for primary coolant radiation level and narrow range containment sump water level.
It is estimated that the cost of installing an environmentally qualified accumulator tank level or pressure indicating system for the four accumulator tanks is approximately $100,000.00.
Installation of this type of equipment in a nuclear power generating station is accomplished with the reactor in a shutdown condition.
Since much of the necessary environmentally. qualified equipment requires long l'ead time (22 wks) to purchase, the installation of this equipment is scheduled to be installed duririg the first refueling outage.
It i.7 estimated that'the cost of installing environmentally qualified equipment to monitor containment sump water temperature is $40,000.00.
The installation of this type of equipment in a nuclear power generating station also is accomplished with the reactor in a shutdown condition.
Installation, times require that the reactor be shut down for an extended period and, therefore, this equipment is scheduled to be installed during the first i
refueling outage.
8)
Describe how Applicants would. perform post accident energy balance calculations without knowledge of the sump water temperature In an accident monitoring situation there is no need.to perform a post-accident energy balance calculati~on.
These calculations have been already performe.d by UE&C and s
m Westinghouse for design basis accidents which are described in FSAR chapters 6 and 15.
From the design basis accident analysis, we have conservatively determined the maximum sump water temperature and assared that plant design and component operation will be maintained throughout the governing design basis events.
The sump temperature is governed by a maximum injection water temper' ture from the Emergency Core Cooling System a
(ECCS), which originates from the Refueling Water Storage Tank (RWST).
The RWST water temperature is limited by Technical Specification to a maximum value of 989 F at Seabrook.
In addition, the accident analysis only takes
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credit for minimum sump water cooling capability through the RHR and Containment Spray Heat Exchangers.
These analyses provide conservative assurance that the plant can safely shut down for any design basis accident event and therefore there is no requirement for containment sump water temperature as an accident monitoring parameter.
9)
Describe how Applicants would perform net positive suction head calculations for safety system pumps during the recirculation mode without containment sump water temperature.
There is no need to perform a NPSHA calculation in a i
post-accident recirculating mode of operation.
These calculations have already been performed for all design basis events described in the FSAR ' Chapters 6 and 15, in order to assure the plant design can conservatively 4
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withstand all design basis accidents.
Again the governing 4
pcrameter which ensures NPSHA for safety system pumps is the
. mas.imum temperature Technical Specification on the RWST inventory and there is no need for containment sump water temperature as an accident monitoring parameter.
- 10) What unce rtainties are taken into account in the Seabrook tycle 1 nuclear design calculated shutdown margin at end of life?
The calculated end of Cycle 1 shutdown margin includes an allowance of 10% of the worth of all rods inserted less the most reactive stuck rod, and a 0.1% (delta K/K) uncertainty in doppler effect.
These uncertainties were used in calculating end of Cycle,1 shutdown margin provided in SBN-1090.
- 11) Describe the as.sumptions used in determining submergence potential in the containment building.
The calculation that determined the maximum containment flood level following a LOCA is UE&C Nuclear Calculation
- 4.3.22-07F Revision 5.
It incorporated the following assumptions to maximize the flood level.
1.
The initiating break was a design basis LOCA which i
produces the largest break area.
The break was assumed at the lowest pipe break location, which.
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causes the maximum reactor coolant to drain.
2.
The refueling canal was assumed to trap no water.
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The reactor coolant water inventory was asaumed to be at the upper level limits of the Pressurizer prior to the bre,ak.
4.
8200 GPM of water was assumed to continue to be pumped into containme,nt from RWST following the "Lo-Lo" signal switches from injection to recirculation operating mode.
5.
All water volumes were expressed on an equivalent mass basis at specific minimum design temperature.
This maximized the subcooled liquid height in i
containment.
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6.
The volume of water in the RHR recirculation piping loop was added to the containment flooding volume.
- 12) Describe the plant maintenance / surveillance program related to aging of electrical equipment'.'
All equipment qualification test report documentation has been reviewed and evaluated to identify all specific equipment qualification maintenance requirements necessary for the maintenance of qualified life.
This evaluation defined all EQ maintenance requirements, (i.e.,
frequencies and intervals) for implementation into the Station Maintenance Program.
The Seabrook Station Maintenance Program is a computer database tracking and scheduling program.
This program is designed to identify, track and schedule all maintenance activities and functions required t
to ensure that the qualified life of all safety-related,
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equipment is maintained.
For a detailed description of the Seabrook Station-Maintenance Program refer to Chapter 5 of the previously submitted report on environmental qualification of electrical equipment important to safety (submitted via SBN-886).
- 13) Have corrective measures been taken for all the deficiencies identified by the NRC staff's environmental qualification audit of February 25, 26 and 27, 1986?
Yes.
In SBN-1127 the NRC was informed that Seabrook Station had completed its EQ Program and had resolved all NRC audit observations.
The Appiicant submittals regarding EQ between the date of the audit and SBN-1127 which in some degree were related to the observations and comments made by the NRC Staff during their environmental quali'fication audit are SBN-988, s
-998, -1005, -1006, -1024, -1030, -1031, -1061, -1081, -1090 and -1126.
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' Ge(rge S.' Thomas STATE OF NEW HAMPSHIRE i
Rockingham, ss.
Augustf64, 1986-Then appeared before me the above subscribed George S.
Thomas and made oath that he is the'Vice President-Nuclear Production of Public Service Company of New Hampshire, authorized to execute the foregoing answers on behalf of the e
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applicants and that these answers are true and correct to the best of his knowledge and belief.
Before me, c-, n-i.
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i EXHIBIT 1-A 1.
PSNH Letter (SBN-274), dated May 11,1982, "Seabrook Station Control Room Design Review Preliminary Report", J. DeVincentis to F. Furaglia.
2.
USNRC Generic Letter (82-33), dated December 17, 1982, S2pplement I to NUREG-0737, " Requirements for Emergency Response Capability".
3.
USNRC NUREG-0700, dated September, 19ol, " Guidelines for Control Room Design Reviews '.
4.
PSNH Letter (SBN-499), dated April 14,1983, " Response to Generic Letter 82-33; Supplement 1 to NUREG-0737", J. DeVincentis to D.
Eisenhut.
5.
PSNH Letter SBN-530), dated July 7, 1983, "Seabrook Station Control Room Design Review", J. DeVincentis to G. Knighton.
6.
PSNH Letter (SBN-544), dated August 11, 1983, "Seabrook~ Station Control Room Design Review In-Progress Audit", J. DeVincentis to G.
Knighton.
7.
USNRC Letter, dated January 12, 1984, "Results of In-Progress Audit of Seabrook Station Detailed Control Room Design Review", G. Knighton to R. J. Harrison.
S.
PSNH Letter (SBN-701), dated July 30, 1984, " Response to DCRDR In-Progress Audit; SER Outstanding Issue #19, NUREG-0737, Item I.D.1", J. DeVincentis to G. Knighton.
9.
PSNH Letter (SBN-748), dated January 7,1985, " Detailed Control Room Design Review; SER Outstanding Issue #19, NUREG-0737, Item I.D.1", J. DeVincentis to G. Knighton.
10.
PSNH Letter (SBN-839), dated July 17,1985, " Supplemental Infor-mation as a Result of Continued Detailed Control Room Design Re-view (DCRDR) at Seabrook Station", J. DeVincentis to G. Knighton.
11.
USNRC NUREG-0896,-Supplement No. 3, Section 18, dated July,1985,
" Safety Evaluation Report, Related to the Operation of Seabrook Station Units 1 and 2".
12.
PSNH Letter (SBN-914), dated December 27,1985, " Detailed Control Room Design Review (DCRDR) at Seabrook Station (SER Outstanding Issue No. 19)" J. DeVincentis to V. S. Noonan.
13.
PSNH Letter (SBN-948), dated, February 20,1985, " Detailed Control Room Design Review (DCRDR) at Seabrook Station (SER Outstanding Issue No. 19", J. DeVincentis to V. S. Noonan.
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EXHIBIT l-A (Continued) 14.
USNRC NUREG-0896, Supplement No. 4, Section 18, " Safety Evaluation Report, Related to the Operation of Seabrook Station Unita 1 and 2".
15.
Nuclear Engineering Services Agreement between Yankee Atomic Electric Company and Thomas B. Sheridan Associates, previously submitted in response to Interrogatory No. NH 10.2.
16.
Seabrook Station FS AR.
17.
McCorm:lck, E. J., and Sanders, M. S., Human Factors in Engineering
'and Design.
18.
USNRC NUREG/CR-1580, Human Engineering Guide to Control Room Evaluation.
19.
USNRC NUREG/CR-3415, Nuclear Power Plant Control Room Task Analyses:
Pilot Study for Boiling Water Reactors.
r 20.
USNRC NUREG-0835, Human Factors Acceptance Criterla for the Safety Parameter Display System (Draft).
- 21. 'INP0 83-026, Control Room Design Review Implementation Guideline.
'l 22.
EPRI NP-309, Human Factors Review of Nuclear Pcwer Plant Control Room Design.
23.
EPRI NP-1982, Evaluation of Proposed Control Room Improvements Through Analyses of Critical Operator Decisions.
24.
In addition, a file has been developed containing a large number of memos, letters, and notes of meeting which deal with the study, review, and critique of the main control panel design.
This file is labeled 'MCB - Seabrook, Human Factors, 199.99.29" and is main-tained by Seabrook Project, Yankee Atomic Electric Company, 1671 Worcester Road, Framingham, MA 01701.
25.
PSNH Latter (SBN-987), dated April 2,1986, " Request for Additional Information Concerning Safety Parameter Display System for Seabrook Station", J. DeVincentis to V. S. Noonan.
26.
Westinghouse Letter, dated June 24,1986, " Notification of NRC of Shutdown LOCA Information Provided Utilities",'E. P. Rahe to J. M. Taylor.
27.
Westinghouse Owners Group Emergency Response Guidelines Executive Volume Revision 1, September 1, 1983.
28.
WCAP-10599 - Emergency Respdnse Guidelines Validation Program Final Report, June 1984.
29.
PSNH Letter (SBN-920), dated January 6,1986, "NUREG-0731 Task I.D.2, Plant Safety Parameter Display Console", J. DeVincentis to V. S. Noonan S
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EXHIBIT l-B Curriculum Vita (e) 4 4
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March, 1985 Summaty Resume of THOMAS B. SHERIDAN Thomas B. Sheridan was born in Cincinnati, Ohio, December 23, 1929. He attended Purdue University (B.S. 1951) and, a.f te r two years in military se'rvice (Aeromedical (aboratory, Wright Patterson Air Force Base, Ohio) attended the University of California, Los Angeles (th s.1954) and M. I.T. (Sc.D 1959).
His doctoral, program was interdependental between systems engineering and psychology, with one year spent in cross-registration at Harvard University.
For most o f his career, 'Dr. Sheridan has remained at M.I.T., where in 1970 he became Professor of Mechanical Engineering and more recently Professor of Engineering and Applied Psychology.
He heads the Man-Machine Systems Laboratory and teaches both graduate and undergraduate sub'jects in Man-Machine Systems.
He is a Faculty Associate of the M.I.T.
Science, Technology and Society Program.
He helped develop a new interdepartmental graduate degree program in Technology and Policy, and has. taught the core seminars for that program.
He has also taught control, design and other engineering subjects.
He has served as visiting faculty member at the University of California, i
Berkeley, Stanford University, and the Technical University of Delf t, Netherlands.
D r.. Sheridan's research has been on mathematical models of human operator and socio-economic systems, on man-computer interaction in piloting aircraft and in supervising undersea and industrial robotic systems, and on computer graphic technology for 'information searching and group decision naking.
He is author, with W.R.
Fe r re ll, of Man-Machine Systems: Information, Control and Decision Models of Human Performance, M.I.T. Press, 1974, 1981 (published in Russian, 1980) and co-editor of a 1976 Plenum Press book, Monitoring Behavior and Supervisory Control.
He is a fellow of the Institute of Electrical and Electr,onics Engineers, was formerly editor of the IEEE Transactions on Man-Machine Systems, is past president of the IEEE Systems, Man and Cybernetics Society, chaired the IEEE Committee on Technology Forecasting and Assessment.
He is also a fellow of the Human Fac to rs Society, and in.1977 received their Paul M.
Fitts Award for contributions to education.
He is Associate Editor of Automatica and on the Editorial Advisory Boards of Computer Aided Design and Robotics and Computer-Integrated Manufacturing.
He is listed in Who's Who in America and other Who's Whos.
Dr.
Sheridan has served on the Accident Prevention and Injury Control Study Sections-of the National Institutes of Health', the NASA Life Sciences Advisory Committee, the NSF Automation P.esearch Council, the NASA Study group on Robotics, the U.S. Congress OTA Task Force on Appropriate Technology, and the NSF Advisory Committee on Applied Physical, Mathematical and Biological Sciences.
He is chairman of the National Research Council Committee on Human Factors and served on the NRC Ad Hoc Committee on Aircrew-Vehicle Interaction and two advisory pancis of the NRC Marine Board.
His industrial consulting activitics have included: The General Motors Corp.
(auto safety); General Electric Co. (telemanipulators); C.S. Draper Laboratory (design of astronaut interface for Apollo guidance system, industrial robots); Biodynamics, Inc. (bioned'ical and human factors); Public Broadcast Service (TV audience feedback);
National Bureau of Standards (industrial robots); Group Dialog Systems, Inc.'(group meeting and decision technology); Northrop Aircraft (pilot workload); Babcock and Wilcox Co. (industrial instrumentation); Lockheed, General Physics, American Electric Power, Consumers' Power. Gibbs and Hill, Virginia Electric Power, General Public Utilities, Stone and Webster, the BWR Owners' Group, Brookhaven National Laboratory, Ycnkee Atomic, and Electric Power Research Institute (man-machine aspects of nuclear pla r.t safety).
Dr. Sheridan is married to the former Rachel Rice and has four children.
l
Edward A. Sawyer Manager Engineering Services Group Mr., Sawyer received his Bachelor of Science degree in Electrical Engineering from Northeastern University in 1965.
In 1985, he received a Master of Science degree in Fir ^
rotection Engineering from Worcenter Polytechnic Institute.
Mr.. Sawyer's kork experience includes employment by the New England Power Company, where he worked in the Electrical and Mechanical Maintenance sections at the Brayton Point Power Station.
He also spent three years in the Peace Corps, working as a Professor of Electrical Engineering with the State University of Chile, South America.
Mr. SnVyer began his experience in the nuclear industry with Yankee. Atomic Electris dompany in 1968.
He was assigned as an engineer in the Project' Group respQGsible for the design and construction of the Vermont Yankee Nuclear F699r Plant.
He held various assignments and* responsibilities ac a part of the Project Group, including a one year assignment at the Vermont Yankee Plant during Start-up Testing, where he worked in the Electrical and Mechanical Maintenance Department.
In 1972, Mr. Sawyer became Electrical Project Engineer on the Seabrook Nuclear Power Station Project, and in 1974 was promoted to Project Manager on tire Central Maine Power Company -Nuclear Power Project.
In 1976, Mr. Sawyer became Yankee's Fire Protection Coordinator, responsible for the overall preparation and implementation of the fire prevention and protection programs for the Yankee Plant at Rowe, Vermont Yankee, Maine Yankee and Seabrook.
In 1983, he was promoted to Principal Engineer, with additional responsibilities in the area of human factors review of plant control rooms.
He became the Review Team Leader for the Seabrook r
DCRDR, and subsequently became Management Team Chairman for the Rowe DCRDR and Management Team Member for the Vermont Yankee DCRDR.
In 1983, Mr. Sawyer became Manager of the Engineering Services Group within the Plant Engineering Department.
This group provides Yankee Atomic Electric Company with engineering support in the areas of human factors, fire protection, training, plant performance and other generic engineering areas.
l l
e-WILLIAM G. ALCUSKY t
SENIOR ELECTRICAL ENGINEER B.S. Electrical Engineering, Northeastern University M.S. Management, Wqrcester Polytechnic Institute t
Registered Professional Electrical Engineer - Massachusetts Mr. Alcusky has been a member of the Seabrook Control Room Design, Review team since its initial formation. He has been involved in the development of the program for the Seabrook Control Room Design Review, the efforts to identify Human Engineering Deficiencies (HEDs), and the evaluation of alternatives to correct HEDs. He has also participated in earlier Control Room Design Reviews under the BWR Owners Group Program. He is currently a member of the IEEE, Subcommittee on Human Factors and Control Facilities.
Mr. Alcu' sky joined Yankee in 1979 working on Instrumentation and Control upgrades and Human Factors reviews up to the present time.
W l
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y
SB 1 & 2 Amandmant 58 FSAR April 1986 RESUME:
Jerry L. Peterson s
SUMMARY
OF QUALIFICATIONS:
Sixteen years experience in the operation of a nuclear power plant, both commercial and military.
Experience includes:
procedure writing, initial startup, full power operations, refueling and minor maintenance and testing of various nuclear power plant systems.
EXPERIENCE:
April 1982 to Present Shift Superintendent, Public Service Company of New Hampshire Functions as the on shift representative of plant management and supervises all station operations required for safe, efficient and dependable service.
SIA qualified.
95 July 1979 to April 1982 Unit Shift Supervisor, Public Service Company of New Hampshire Directly responsible for the overall safety and efficiency of the unit to which assigned and for directing the actiwities of personnel assigned t'o that' unit.
January 1973 to July 1979 Power Authority of the State of New York Indian Point Unit No. 3 - Reactor Operator Responsible for shift operations, and supervisor of support groups,Lof a 1000 megaw'att electric nuclear station.
Also responsible for acceptance testing associated with nuclear equipment, instrumentation and systems.
Obtained Unit No. 3 operators license January 1977, Docket No. 55-4974.
Indian Point Unit No. 3 - Nuclear Plant Operator Pesponsible for the safe and competent operations of clie I
nuclear and conventional systems for the Unit No. 3 facility. Was involved during the construction phase with flushes, hydrostatic tests, equipment and system operational
~
tests, fuel assembly inspection and acceptance.
4' i
.__ ~-- _ _-
SB 1 & 2 Amandm2nt 58 FSAR April 1986 Jerry L. Peterson Page 2 Consolidated Edison of New York, Inc.
New York, New York Indian Point Unit No.2 - Reactor Operator Completed the formal reactor training program at the con Edison Indian Point Facility.
During this time became intimately feniliar with all components of the primary and secondary systems, including:
the design, purpose, limitations and normal / emergency procedures. Also had extensive training on the Indian Point Simulator including full power operations and emergency casualty training.
Obtained Unit No. 2 license January 1975.
Indian Point Unit No. 2 - Nuclear Plant Ooerator Responsible for the safe and competent operations of the nuclear and conventional systems of the Unit No. 2 start-up program, up to and including the initial core loading.
United States Navy 1966 to 1972 Assigned to the USS John C. Calhoun, (SSBN630) as an Elect.ronic Technician.
Qualified as a Reactor Operator and as an Engineer Watch Supervisor. Was involved with a complete overhayl and refueling of the nuclear power plant on the Calhoun.
EDUCATION AND TRAINING:
April 1985 B.S. Degree, N.Y. Regents October 1984 SRO License, Seabrook Station, Unit 1 January 1983 to September 1984 Seabrcok License Training Program October 1982 STA Qualified 1973 to 1977 Con Edison Training for Nuclear Plant Operator and Reactor Operator.
1966 to 1968 Electronic Technician School, U.S. Naval Nuclear Power School 1964 to 1966 Edinboro State College 47 13D-54 4
t e
<ws we
SB 1 & 2 Amendment 58 FSAR April 1986 RESi21E OF QUALIFINATIONS LAURINCE A. WALSH Have worked in the nuclear field since early 1961.
Previous job held covered the full range of operations. With two commercial and one Navy new construction plants behind me, Seabrook construction is no stranger. Will complete training necessary to hold any management position in a Nuclear Co= plex.
POSITION Operations Manager EDUCATION St. Thomas Grammar Graduated 1,955 Sacred Heart High School Graduated 1959 Navy Schools:
Basic Electronics and Electricity School Inte*----"a* cation Technicians School Submarine School Nuclear Power Training School Oxygen Generator Operations & Maintenance H2 Analyter Operations & Maintenance Vibration' Analysis Motion Projection Operation & Maintenance Connecticut Yankee Startup Training Course Central Maine Vocational Institute Instructor Training Maine Yankee Startup Training Course Central Maine Vocational Institute Technical Writing American Management As sociation Communications Course Psychology 401 University of New Hampshire Four credit hours Mathematics 1211 Meephis State University Three credit hours Physics 2511 Memphis State University Four credit hours Physics 2512 Memphis State University Four credit hours Nuclear Physics Memphis State University Phys. 4110 Three credit hours Reactor Physics Memphis State University Phys. 4210 Three credit hours Instrumentation Memphis State University Tech. 2411 Three credit hours Chenistry Memphis State University Ch em. 1010 Three credit hours Radiation Protection Memphis State University Biol. 4080 Three credit hours Csiculus I Memphis State University Math 1321 Four credit hours Calculus II Memphis State University Mach 2321 Four credit ho rrs Additionally, Memphis State University has been contracted for an addi-
~
tional 51 credit hours of Shif t Technical Advisor courses.
Management training - PSNH 280 hours0.00324 days <br />0.0778 hours <br />4.62963e-4 weeks <br />1.0654e-4 months <br /> 13D-10
l SB 1 & 2 Amendment 58 FSAR April 1986 S
L. A. WALSH - RES"ME Psge Twc b
EXPERIENCE 1956 to 1958 Worked in shipping, receiving, display and advertising for' local department store.
(Grive, Bisset & Holland, Waterbury, Connecticut) 1958 to 1959 Worked in an eyelet manufacturing shop as a machine operator and started an apprentice program for tool making.
(Westbury Mfg. Co., Waterbury, Connecticut) 1959 to-1961 Af ter completion of service schools associated with my rate, served aboard USS Sea,0vl (SS408) for approximately nine months.
Duties performed while aboard were operat. ion and nain-tenance of all com=unications systems, electronic compasses and ships batteries.
1961 to 1962 Af ter completing nuclear power training school, qualified as a reactor operator at-the SIC training prototype at Windsor, Connecticut.
(Conbustion Engineering PUR) 1962 to 1964 Served as a member of the nuclear engineering crew aboard the USS Laf ayette (SSBN616) thrcugh initial construction, sea trials, comnissioning and operations.
Among other duties, I wa.s also a qualified reactor operator on the SSW plant and also filled the duties of the spare parts petty officer.
1964 to 1966 Was employed by Connecticut Light and Power Company at their 600 MUe fossil fuel generating station in Devon, Connecticut.
While awaiting startup crew assignment at Connecticut Yankee Atomic, served as operating assistant.
1966 to 1970 Transferred to Connecticut Yankee Atomic Power Company and was a member of the staf f during construction and startup of the station.
Was elected business agent of the local union and received AEC reactor operating licenses OP-2438 and OP-2438-1.
1970 to 1978 Transferred to Maine Yankee Atomic Power Company to assist.with plant acceptance from the NSSS and AE.
Assisted with instruc-tion of prospective licenses holders and formulated initial procedures for plant operations.
Promoted to Assistant Department Head for Plant Operations and also represented con-pany as Project Engineer for a backfitted system to complement the cooling water outlet dif fuser.
While at 9aine Yankee, I held an NRC Senior Operating License 4 SOP 1693, SOP 1603-1 and SOP 1693-2, the latter being current until August 1973.
197.8 Transf erred to Public Service Company of New Hannshire.
Have 4
to Present hired a staf f for management 'of' the Operations Department and Unit I licensed operators. Have staged the effort for proce-dure developnent and an nidway through plant syst, ens design review.
I 3D ll...
m
EXHIBIT 1-C
SUMMARY
OF PROPOSED TESTIMONY
, 1.
Overall Position In our-j udgement, (1) the Seabrook DCRDR has been conducted in accordance
> w'ith the highest standards of human factors engineering practice and the requirements of Supplement 1 to NUREG-0737, and (2) the SPDS meets the requirements of Supplement I to NUREG-0737. The controls, displays, and other modification added or planned to be added to the control room as a result of the DCRDR will, in our opinion, reduce the potential for operator error.
2.
Use of NUREG 0737, Supplement 1 NUREG-0737, Item I.D.1, states that all licensees and applicants for operating licenses will be required to conduct a detailed control room design review (DCRDR) to identify and correct design deficiencies.
It lists guidel.ines documents issued or to be issued by the NRC, and dicusses the NRR review of the licensee's submittal. NUREG-0737, Item I.D.2 states that each licensee shall install a safety parameter display system (SPDS).
Supplement 1 to NUREG-0737 (generic letter 82-33) was issued to present a distillation of the requirements and provide additional clarification for l
five items on NUREG-0737, including the DCRDR and SPDS.
Therefore, Supplement 1 to NUREG-0737, not the NUREG itself, presents the requirements that have to be met to provide an acceptable Detailed Control Review program.
For this reason, NHY designed its program to meet the requirment of Supplement 1 to NUREG-0737.
3.
Requirements for Performance of DCRDR s
According to Supplement 1 to NUREG-0737, the basic requirements for conducting a DCRDR are:
Formation of, a multi-disciplined team to perform the review Performance of a function and task analysis to. identify control room operction tasks and to determine instrumentation and controls (16C) requirements Comparison of requirements against an inventory of Control Room l
instrumentation and controls Performance of' a control room survey to identify deviations from accepted human f actors principles j
Assessing which Human Engineering Deficiencies are significant and should be corrected.
I Selection of design improvements that will correct those discrep-ancies.
l l
Verification that the selected design improvements will provide l
the necessary correction I
Verification that these improvements do not create new HEDs.
l
EXHIBIT 1-C
( Conti nutd) 4.
Seabrook Response to Supplement I to NUREG-0737 Requirements NHY has met all of the above requirements.
(a) A multi-disciplined team, consisting of Program Manager, Human' Factors consultant, Instrumention and Control Engineer, and Station Shif t Superintendent, was formed. This team was expanded as necessary to include other plant operators and other engineering disci.plines for specific reviews.
(b) An extensive and comprehensive function and task analysis was performed to determine Instrumentation and Control requirements.
'( c) The I&C requirements developed as a result of this function and task analysis were compared agains't the actual Control Rocm instrumentation and controls.
(d) A control room survey and review was performed to identify deviations of the control room instrumentation and controls f rom accepted human factors principles.
(e) The potential Human Engineering Deficiencier. (HEDs) were assessed and prioritized. These activities Included survey, review, and assessment of all displays and controls, including S,DDS displays; and prioritization of all HEDs.
(f) Design, procedere and/or' training changes were selected to correct significant HEDs.
(g) During the assessment process, the Review Team verified that the selected design, procedure and/or training changes resolve the HEDs.
(h) A review will be performed by the Review Team to verif y that the improvements do not create new HEDs. This review has been per-formed on the chcnges made to date.
(i) The NRC Staf f, in Supplement 4 to the Seabrook SER, has concluded that "PSNH has conducted a DCRDR for Seabrook Station that satisf actorily meets the requirements of Supplement I to NUREG-0737."
5.
Issues of Material Fact (SAPL's Response to Applicant's Motion for Summary Disposition or Contention SAPL Supplement 6 (formally NH-10)
The following " issues of material fact" will be discussed on detail.
III..The Video Alarm. System (VAS) color coding scheme should be modified to be consistent with other control room CRT's prior to fuel load.
IV.
A preliminary evaluation of control room environment ought to be accomplished prior to fuel load.
a EXHIBIT l-C (Continutd)
V.
Control room f urnishing HED's and operator protective equipment and emergency equipment storage HED's should be.
resolved prior to loading -fuel as the. Applicants committed to in SBN-839.
6.
Conclusion Based upon all of the foregoing, it is our opinion that a DCRDR in
. compliance with NUREG-0737 1.D.1, has been conducted for Seal,r'ook and that theJdisplays and controls added, or to be added, to the control room as a result of the DCRDR do not increase the potential for operator error aad in. fact reduce it.
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9 4
9 8
5 W
f 5
2 s
EXHIBIT ID V.
GENERIC HUMAN ENGINEERING DISCREPANCIES G.
Indicators 3.
. Technical Specification alarm levels and-operating bands are not shown.
Resolution Operating plant management will determine what, if any, markings are necessary.
These will be shown on the indicators.
Priority 2C, s
Examples:
LI-4079 (CST level)
LT-2607 4.
Dual range scale meters sometimes use different scale units (e.g.,
300 psi, 3 x 102 psi).
Resolution (Submitted in SBN-701)
Some dual indicators have a different notation on the same indicator.
This will be changed so that the notation is the same on a single a
dual-scale indicator, if the variable is the same.
It will not be changed if the variables are different.
Priority 3C.
J.
Recorders 3.
Normal and abnormal ranges are not indicated.
Resolution Operating plant management will determine what, if any, markings are necessary.
These will be shown on the recorders.
Priority 2C.
K.
Lighting i
1.
There is a glare on indicators, recorders, and CRTs.
I s
6
e-j.
Resolution i
A study will be made of ths glare problem and jo the methods that can be used to reduce it.
Some P
of this work can be done n'ow, some must wait until the control Room is. complete.
Priority 3C.
VI. SPECIFIC HUMAN ENGINEERING DISCREPANCIES
-B.
Panel C Fr'ont 8.
The RCP ammeters have too insensitive a scale to
- distinguish li,mits on current.
They need narrow range.
Resolution Procedures should be reviewed to insure this is necessary.
Band the normal' range after this is determined during plant start-up.
If it is
. determined after start-up that a narrow-range indicator is necessary, it will be added.
Priority 3C.
Revised R'esolution (Submir,ted.in SBN-701) i The required function 15 to provide an 4
approximate indication that the RCP motors are drawing full load current.
The actual current drawn by the motor will vary with bus voltage fluctuations, the number of RCPs running, the temperature of the primary coolant, and the as-built characteristics of the motor.
The present indicators have a range of 0-400 amps; 5
they are sufficiently accurate for the required function.
The normal range will be determined during start-up and banded on these indicators (reference HED V.G.3).
Priority 3C.
4 I
i C.
Panel D Front 9.
Adjustment of boration-dilution is difficult without inadvertent alteration of count when window is raised; also, decimal point is not indicated; need to specify units (gallons).
l Pesolution This is a continuous operation, with a potential for error.
We will either rework window or look for another type of controller.
Priority 2C.
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-_.. - -n.n-ve,w-e
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D.
Panel E Front 9.
The main turbine impulse chamber pressure is mislabeled, possibly should be SG reference level.
Resolution I
For-Seabrook, SG reference level is set at a constant 50%.
This indication is not needed.
Delete the TG impulne chamber pressure and correct hierarchical labeling.
Priority 3A.
Revised Resolution v
The indication is not necessary and will be removed.
Priority 3C.
The implementation schedule originally proposed
,is being changed because:
(a) Since the operator does not use this meter, there is no safety concern; (b) At this time, the demarcation labeling and touch-up painting in this area of the board have been completed; and (c) Changes in other areas of the MCB.would make.
it extremely difficult to make this modification and to revise the completed demarcation and labeling per the original implementation schedule.
13.
There is a need for DP meter plua/minus 0 - 300 psi to go with PISO7 and PISO8.
Need better way to watch SG feed and main stream delta P.
Requireme'nt is to hold the programmed delta P across feed regulator valves.
This is especially needed if auto system fails.
Original Resolution
~ The feedwater control valve controllers presently have a delta P indicator with a C-100%
scale.
The scale will be changed to delta P in psig, resolving this problem.
Priority 2A.
Revised Resolution i
The feedwater* control valve controllers presently have a delta P indicator with a 0-100%
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_ m,_ _ _. _ _ _
scale.
The scale shall be changed to delta P in psig.
Priority 2C.
The implementation originally proposed is being revised because:
(a) The operators'can maintain the required delta P across the feedwater control valve using the 0-100% scale; (b) Other instrumentation (e.g., SG water level) would be used oy the operator if the feedwater control valve was in the manual mode; (c) Based on (a) and (b), we do not believe there is a safety concern. associated with delaying implementatsion; and (d)' Vendor problems, as well as ongoing changes in other areas of the MOB, make this modification difficult to complete per the original implementation schedule.
16.
There is no narr~ow range indicated on wide-range SG scale.
t Resolution The indicator will be banded.
Priority 2C.
18.
There is no low-range MEW flow indication for SGs for start-up conditions.
1 Resolution The need for this will be assessed during.
start-up.
If the automatic level control on the bypass does not work as designed, then an indicator will be considered.
Priority 2C.
E.
Panel F Front 12.
Turbine megawatt has poor scale.
Re_ solution This is a vendor-supplied meter.
There are other megawatt indications'that the operator can refer to.
Leave as is.
Priority 3E.
P +
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~,, - - -., - -,
t
,evised Resolution R
The scale will be changed to provide readable indication.-
Priority 3C.
15.
The turbine speed indicator should have 1800 marked on scale.
Resolution This will be marked on scal 4.
Priority 3A.
Revised Resolution (Submitted in SBN-7Cl)
The critical speeds will be determined from the start-up procedure.
These control speeds and 1300 rpm will be marked on the scale.
Priority 3C.
20.
There is no turbine generator emergency bearing oil pur5p header pressure indication available in the Control Room.
Resolution This pressure and others are available locally.
The need for indication in the Control Room will be assessed during the first operating cycle.
If changes are needed, they will be made.
Priority 3C.
s
-K.
Panels H and I Rear 7.
Turbine supervisory or equivalent computer display should be more accessible to operator on start-up on front of panel.
Numbering scheme seems poor.
How much can be done on CRT display, and is CRT-trending better for this purpose?
RPM does not agree with RPM meter on turbine control panel.
Cannot read vibration with sufficient accuracy.
Turbine supervisory needs general review.
Resolution This review will be done after plant start-up.
Any required changes will be made later.
Priority C.
L.
Hard-Wired Alarms 1.
Grouping of Tiles Within Boxes 1 6
On Panel B, the tiles for the two trains do not correspond.
There are two Train B tiles.ou the A train box which should be me.ved to the B train box.
Ihe SCCW alarms on the Panel H box ought to be on the Panel F box.
On Panel H the11 eft-hand box is 4 x 6, the right one is 6 x 6.
At'least the left-most 4 x 6 portion of the right-hand box should correspond 4
to the left-hand 4 x 6 box.
4 On all annunciator boxes, some alarmc come up on panels that are separated from the controls and displays associated with the alarm.
Additionally, the wording on some tiles does not indicate the correct parameter alarmed.
Resolution A g'eneral review has been undertaken to determine correct grouping ard wording on the tiles, and necessary changes have been mada.
Priority 2A.
Additional Clarification (Submitted in SBNa949)
' The grouping, placement, and~ wording was reviewed.
The resolutions identified as a i
result of this review were prioritized (i.e.,
2A, potentially affecting safe plant operatien; 3A/3C, operator convenience).
Those resolutions prioritized 2A and 3A have been implemented.
Those identified as 30 will be implemented by the end of the first refueling.
The following is a list of annunciator changes planned for the first refueling:
1.
Relocate alarms for:
(Priority 3C) i a.
Containment IA Pressure Low b.
PCCW Head Tank B - Level Low c.
PCCW Train B - Flow Low 2.
Revisions of the input (Priority 3C) logic to add Train B inputs to alarms for:
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a.
FWST Level Lo-Lo b.
Containment Pressure Hi-1 C.
Containment Pressure Hi_3 3.
Combinc:. the following two (Priority 3C) alarms into one alarm:
a.
RCP Cooles Train A PCCW Flow Low b.
RCf Coolers Train B PCCW Flow Low 4.
Add alarm for SSPS general (Priority 3C)
, arning alarm, both trains - reactor trip.
w N.
Commen Alarm System 2.
The silence, acknowledge, and reset push buttoms are duplicated on the keyboard and are insufficiently discriminaole.
Each set is dedicated to either the VAS or the backup annunciator requiring additional operator actions to acknowledge alarms that come in on both systems.
In addition, two keystrokes are required to acknowledge VAS alarms from the keyboard.
Resolution The existing push-button stations with silence, acknowledgs, and reset will be wired to interface with both systems.
For operator convenience, the acknowledge push button will also silence the horn.
Priority 2A.
The silence push button will be made distinctive from the other two push buttons.
Silence, acknowledge, and reset push buttons shall be distinguishable from each other by color'as well as location relative to the others.
The silence push button shall be gray, the acknowledge magenta, and the reset white.
Priority 3C.
4.
The test station for the hard-wired annunciator boxes on Panel H is too far from the acknowledge station for efficient operation.
l Resolution The test push button will be noved to a position nearer the acknowledge station.
Priority 3C. r S
W 6
Q.
HEDs Identified During Ongoing Revi'ew 3.'
The' MSR indicators and associated control switches are not consistently laic; out left te
- right, Resolution The two indicators.will be changed to provide proper correlation with the control switches.
Priority 3C.
5.
Westinghouse process control indicators are not laid out in a left-to-right orientation, Ch.
I, Ch. II, Ch..III, and Ch. IV, for channel failure analysis and consistent orientation to match the existing bistable status light arrangement.
'Zhis includes controlling signal switch p:.acement.
Resolution A coded sticker will be placed on or near the indicators to indicate the correct channel.
The controlling switch will be put under the channel it controls and will be demarcated.
Priority 3C.
7.
For motor-operated valves, there is no way to tell if the motor thermal overload has actuated.
In this situation, the indicating light (s) remains or., but the valve will not respond to actuation signals.
Resolution This problem should be addressed in a manner to provide the operator with information that the thermal overload has actuated.
The human factors team will review the resolution to insure it does not create another HED.
Priority 2C.
R.
F' ire Panel - NRC Item, Appendix, Part C, Item 8.24; Seabrook Item VI.K.4 1.
Fire panel alarms from various plant areas are randomly grouped on the Control Room fire alarm panel.
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s e
Resolution To aid the operator in quickly discerning which building has an alarm, the alarms will be grouped by building and by elevation within the building.
The alarm labels will clearly indicate fire panel number and the building where the panel is located.
Priority 3C.-
t S.
RDMS - NRC ' tem, Appendix, Part C, Item 3.3; Seabrook Item VI.A.19 10.
The labeling on several.of the push buttons does not accurately describe the displays called up on the CRT.
(Submitted in SBN-914) a.
Push buttons are labeled Grid 1, Grid 2, and Grid'3.
The information displayed are the area, process, and airborne monitors.
- b. Push buttons are labeled Grid 4, Grid 5, and Grid 6.
These are for Unit 2 information, not,used in Unit 1.
- c. The selection controls for the computers are labeled Primary / Alternate.
The computers are labeled System 1/ System 2.
- d. A push button is labeled Alarm Acknowl' edge, but it does not perform this function.
Resolution a.
Change the push button labels to read:
Area, Process; and Airborne.
- b. Make these push buttons blank.
c.
Change the controls to read:
System' 1/ System 2.
i
- d. Change the push button to read:
System
~ Acknowledge.
(
Priority 3C.
T.
HEDs Resulting froa Procedure Walk-Through (Submitted in SBN-839) 1.
In many casec in the procedures, the operator cannot read the indicator to the accuracy
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presently-implied by the procedure, and no
-tolerance is allowed in the value to be read.
Resolution (3ubmitted in SBN-839)
The "setpoints" for operator action in the procedures are the same as the applicable instrument setpoints.
The instrument
-setpoints are specified to a high accuracy to meet the accident analysis.
Since operator action duou not require the fast response required to meet an accident analysis, the "setpoint" for operator action does not require the same accuracy as the instrument set point.
We will indicate the approximate location of the procedure "setpoint" as part of the indicator banding.
Priority 2C.
The operators will be instrudted that the procedure "setpoints" do not have to be read as exact values.
Priority 2A.
Revised Resolution (Submitted in SBN-948)
In order to preclude any concern of reading the indicator (i.e., what accuracy), we will, as part of the indicator banding, identify (i.e., mark) the procedure setpoint.
This banding of indicators is to be done by the end of first refueling per a previous commitment.
It should be noted that the "setpoints" for operator action in the procedures are the same as the applicable instrument setpoints.
The instrument setpoints are specified to a high accu' racy to meet the accident analysis.
Since operator action does not require the fast s
response required to meet an. accident i
analysis, the "setpoint" for operator action
~
does not require the same accuracy as the instrument setpoint.
Therefore, marking setpoint or banding will be adequate for operator action.
2.
During the check of critical safety function, the operator must, check containment high-range radiation.
The only presently available monitor is on CP-180A and B, the RDMS lE panels.
The operator must move from the left side of the board to the far right position of 4
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the RDMS panels to read this indicator.
The importance of this indication suggest.s that it should be in the front of the MCB, on Panel' BF.
Resolution t
Containment high-range radiation indicators will be added on Panel BF.
In the interim, the containment high-range radiation indication can be obtained by the operator manning the right side of the board, by the primary operator through the station computer workstation on the MCB er the RDMS CRT within the primary operating u_ea.
Priority 2C.
U.
Use of Color on CRTs 2.
The useuaf the color magenta ts an alarm in the Graphics System does not fit into the
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logical color scheme developed for the CRTs in the Control Room.
(Submitted in SBN-914)
Resolution Delete the use of the color magenta'for alarms on the Graphics System.
This information is available in other places, and is net needed.
Priority 3C.
4.
On the graphics CRT, a yellow question mark indicates that uncertain information is pre.~ent with respect to a component or' valve.
On the RDMS CRT, magenta indicates that uncertain information is present.
(Submitted in SBN-948)
Resolution The graphics CRT will be changed.
The indication for uncertain information will become a magenta cuestion mark.
This will make the graphics CRT consistent with the RDMS CRT.
Priority 3C.
A_dditional HEDs Resulting From On-Going Review 1.
Removal of the BIT from the plant has caused a problem with the status light panels.
The status lights have blanks where there were once lights.
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Resolution The status light panels that have been affected need to be regrouped to allow pattern recognition.
Prierity 2C.
'1 5
- 2.
The Unit 2 vacuum recorders have pen labeling that is not concistent with the, rest of the board recorders.
Resolution Remove the inconsistent labels.
Priority 3C.
4.
On Panel D, Controller CS-HCV-123 is not permanently-and correctly labeled.
Resolution Correct the labeling.
Priority 3C.
5.
Recorder RC-LR-460 is hot labeled.
Resolution Add a label.
Priority 3C.
7.
There have been many changes on labels on the control board.
In all cases, the labels are i
easily readable.
However, in some cases different type font or stroke widths have been used.
'The labels should be made consistent with respect to style, stroke, width, etc.
Resolution 4
Make the labels consistent across the main control board and RSS panel.
Priority 3C.
8.
Currently, the operator does not have sufficient information on the switch escutcheon to know if the valve requires him to hold the switch in,'the open or close position until the valve makes its ne:tt
- position, i.e.,
a valve that has no seal-in circuit.
Nesolution On valves that have no seal-in circuit (used for throttling), indicate it on the switch j
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escutcheon with the words, " Throttle Open" or
" Throttle Close."
Priority 2C.
3.
-On Panel D, the SI actuation s' witch has'a mislabeled. escutcheon.
It says " trip" in the center position.
Resolutf :en
-Correct the labeling.
It should say
" activate" in the right-hand position.
Priority 2A.
Z.
Control-Display Integration Findings 3.
Panel DF has two controle but no indication of the status of flow and chiller temperature.
The meters needed by the operators Lr on Back g
Panel'BR-1 and require a trip away from the front panel to see the n.eters (NRC D5.61).
1.
Response
(See NRC Item 6.12, Seabrook Item VI.C.6.)
Indicators are available behind the board.
Indication will be added to the computer for the operator.
Priority 3C.
9.
The Control Room environment during operat$on
.should be evaluated (NRC A.1.0.B).
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Response
A great deal of work has been dons in the auditory single area with respect to alarms in the Control Room.
HEDe have been developed and submitted to NRC for r.eview.
A preliminary HVAC evaluatien for the area has i
been done, and no HFns have been found.
It is our intention to complete the final 1
j evaluation of the auditory ' area when the plant is fully opctational with a normal level of sound available', and that the HVAC will be l
fully reviewed after a yearly cooling and heating cycle has been experienced, again, with the plant operational.
Much of the review work has been done, or will be done, before plant operation, but the final evaluation cannot be conpleted until.after plant operation.
Priority 3C.
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s Exhibit 2-A Curriculum Vita (e) t t
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Richard Bergeron Instrumentation & Controls Engineering Supervisor Education BS Marine Engineering, Maine Maritime Academy, May 1969 Mr. Bergeron joined Public Service Company of New Hampshire in May 1982 as Senior 16C Engineer in the Engineering Services Department. His areas of responsibility include coordination of I6C Engineering activities for the Station Staff, Construction and Startup interface activities, as well as, various special projects.
Mr. Bergeron was recently appointed to the position of Instrumentation & Controi Supervisor in the Engineering Depart-ment.
For the past four years Mr. Bergeron has also been assigned as the Station Staff Representative on the E-luipment-Qualification Task Force.
He has been responsible for the coordination and review of the Equipment Qualification Program, as well as, coordinating the implementation of the Station Equipment Qualification Program.
Mr. Bergeron came to Public Service Company of New Hampshire from Stone
& Webster Engineering Corporation where he was employed f rom 1972-1982.
He held the position of Principle Instrument Applicati,on Engineer respon-sible, for specifying, purchasing and design review of electron and pneumatic instrumentation control systems.
Mr. Bergeron is also experienced in the scheduling and preparation of Logic Diagrams and System Descriptions which define the functional control concepts. He was also assigned as a task member to assist in the development and preparation of the 79-01B equipment qualification submittal for Duquesne Light Company.
Between 1969 and 1972 was employed by Cuff 011 Corporation as an engineer in their Marine Engineering Division. There he was responsible for the operation and Maintenance cf Marine Power Plants.
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William J.
Cloutier, P.E.
Lead Systems Engineer Mr. Cloutier received his Bachelor of Science degree in Mechanical Engineering from Norwich University in 1971 and his Master of Science degree in Mechanical Engineering from Worcester Polytechnic Institute in 1973.
He is a registered professional engineer in the state of Michigan.
From 1973 to 1977, Mr Cloutier worked at Stone and Webster Engineering Corporation on various projects.
He worked in the Boston office as a support engineer for system design on the Millstone 3 Nuclear Project and as a staff engineer in the Engineering Mechanics Division assigned to the North Anna 1 Nuclear Project.
He worked on the North Anna Project doing High Energy Line Break (HELB) analysis and seismic design.
Mr. Cloutier worked on field assignments for Boston Edison Company and the Power Authority of New York as a test and startup engineer for Mystic 7 and Storia 6 fossil power plants.
In 1977, Mr. Cloutier joined General Electric Company as a Reliability / Field support engineer servicing Moisture Separator / Reheaters at various nuclear power stations.
From 1978 to 1983, Mr. Cloutier worked at Consumers Power Company in Michigan as a senior engineer.
He was assigned to the technical support group for the testing and startup of the Campbell 3 coal power plant.
He also managed the testing and startup of a variety of modification projects to support operating nuclear and fossil plants.
In 1980, Mr.
Cloutier was assigned to the Midland Nuclear projects in the Safety and Licensing Department.
He worked'as a senior engineer managing special projects related to the soil settlement problems at the plant site and Equipment Qualification.
This included resolving NRC issues regarding buried safety class piping and safety class water storage tanks, which resulted in testimony prepared for the Atomic Committee for Reactor Safety (ACRS) and the Atomic Safety Licensing Board (ASLB).
i In 1983, Mr. Cloutier joined Yankee Atomic Electric Company as a senior engineer assigned to the Seabrook Nuclear Project.
He is currently arsigned as the Lead Systems Engineer for the Seabrook Project.
Mr. Cloutier is an associated member of the American Society of Mechanical Engineers and a member of the National Society of Professional Engineers.
JOSEPN M. SALVO SENIOR MECHANICAL ENGINEER EDUCATION B.S.
Mechanical Engineering, Northeastern University, June 1973 M.S.
Mechanical Engineering, Northeastern University, June 1974 M.S.
Applied Management, Lesley College, May 1986 PROFESSIONAL AFFILIATIONS / REGISTRATIONS Registered Professional Engineer (Mechanical), Commonwealth of Massachusetts Registered Professional Engineer (Mechanical, HVAC), State of New Hampshire American Society of Mechanica] Engineers Mr. Salvo joined the Yankee Staff in February,1982 as a member of the Systems Group, assigned to the Seabrook Project. His areas of responsibility included coordination of the HVAC portions of the project, construction interface activities, FSAR preparation and review, as well as various special projects.
For the past year Mr. Salvo has been responsible for the coordination and review of the Equipment Qualification Program for Scabrook.
Mr. Salvo came to Yankee from Gaulin Corporation, where he was employed from 1977 to 1982. He held the position of Senior Engineer responsible for the design, analyses and performance testing of all pumps for government (U.S. Navy) and nuclear applications. He also coordinated the seismic and environmental testing of this equipment when required.
Between June 1973 and March 1977 Mr. Salvo was employed by Stone and Webster as an engineer in the Engineering Mechanics Division. There he was responsible for computer modeling and finite element analysis.
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George Trouderos From 1959 to 1962, Mr. Tsouderos was associated with the U.S. Army He served as the construction inspector, Coprs of Engineers in Iran.
responsible for maintaining surveillance over construction of certain portions of the Dezful Airfield built by the U.S. Army Corps of Engineers for the Iranian Air Force.
From 1962 to 1963, Mr. Tsouderos served as the Resident Engineer for for the the U.S. Army Corps of Engineers on a construction project U.S. Air Force in Irakion, Crete. The construction project included an extension to existing power plant f acilities and various supporting buildings.
In 1967 Mr. Tsouderos recieved his Bachelor of Sciene in Electrical Engineering f rom Lowell' Technological Institute, Lowell, Massachusetts.
Af ter receiving his degree in 1967, Mr. Tsouderos joined the Electrical His and Control Design Group of the Yankee Atomic Electric Company.
primary responsibilities have been la the area of electrical. engineering in connection with the various Yankee operating plants, as well as new pl a nt s.
He served as Cognizant Engineer during the replacement of the emergency diesel generators at Connecticut Yankee. He has participated in the start-up of Vermont Nuclear Power Plant, and during 1973/1974, he served as Project Engineer on the conversion of the 115 kV Coolidge Line to 345kV, at Vermont Yankee Nucient Power Plant.
In July 1976, Mr. Tsouderos was promoted to Senior Engineer in the Electrical Engineering Group with responsibility for providing expertise and services in all areas of electrical engineering for Vermont Yankee, Mr. Tsouderos has been Maine Yankee, and Yankee Nuclear power Plants.
intimately involved with the electrical design of the Seabrook Station.
In July 1978, Mr. Tsouderso was promoted to Senior Electrical Engineer, and in July 1982 to Principal Engineer in the Electrical Engineering In this position he was responsible for the technical performance Group.
and. quality of engineering activities of the Group, and in assisting the group manager in the day-to-day running of the Group.
In September 1983, Mr. Tsouderos assumed the responsibilities of Lead In this position, he is Electrical Engineer for the Seabrook Project.
responsible for the activities of Electrical Engineering associated with the Seabrook Project.
Mr. Tsouderos is a member if IEEE and as a member of the IEEE Working Group on Batteries, participated in the preparation and review of various IEEE Standards.
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RESUME NEWELL K. WOODWARD
SUMMARY
OF QUALIFICATIONS:
Eighteen years experience in the commercial and naval nuclear power industry.
Extensive experience in nuclear power plant design, con-struction, licensing, operation, and refueling.
Supervision and i
direct participation in the design, procurement, installation, test, operation, and repair of nuclear power plant electrical and mechanical i
equipment.
PROFESSIONAL EXPERIENCE:
March, 1980 to present:
Impell Corporation, Melville, New York Supervising Engineer Supervision and approval of electrical and mechanical environmental i
equipment qualification (E.Q.) work at Impell for utility clients with nuclear power plants. Project Engineer for the preparation of electrical and mechanical equipment E.Q. Files for Seabrook Station.
October, 1976 to March, 1980:
Combustion Engineering, Inc. - Power Systems Division, Windsor, CT NSSS Design Engineer Responsible for the design, technical, and licensing support of a variety of special projects relating to the Reactor Coolant, Steam Supply, Main Feedwater, Condensate, and Auxiliary Feedwater Systems.
These projects included the analysis, design, and operation of system j j,
}l backfits to meet regulatory requirements, system design reviews with j
respect to regulatory requirements, and utility requested design evaluations.
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1974 to 1976:
fi Morrison-Knudsen Co., Inc. - Energy Systems and Services Division, i
i West Milton, NY Senior Quality Control Inspector (1975 - 1976)
Shift Refueling Engineer (1974 - 1975)
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NEWELL K. WOODWARD RESUME (CONT'D) 1974 - 1976:
(CONT'D)
Direct involvement and supervision of the refueling, overhaul, and modification of 3 daval Prototype Pressurized Water Reactor Plants.
Responsible for monitoring and certifying reactor plant servicing procedures, nuclear and non-nuclear welding, brazing, system testing, pressure vessel and fuel lif ting and handling equipment, and nuclear plant materials.
Certified to perform non-destructive testing in accordance with NAVSHIPS 250-1500-1, with civilian certification to ASNT/TC-1A, Level II.
Extensive experience in the theory, main-tenance, and operation of automatic Omega seal welding and cutting equipment.
1968 - 1974:
United States Nevy - U.S.S. Ulysess S. Grant (SSBN-631)
Nuclear Trained Engineering Watch Supervisor Responsible for the supervision of 10 technicians who operated the mechanical, electrical, and reactor controls equipment of a Naval Nuclear Submarine Propulsion Plant.
In addition, as Machinery Division Leading Petty Officer was responsible for the supervision and training of 15 technicians who maintained the mechanical equip-ment associated with the reactor plant and the steam propulsion plant.
EDUCATION AND TRAINING:
Bachelor of Science, Charter Oak College, 1980 University of Hart ford, Four Semesters, Colgate University, Three Semesters Mitchell College, Three Semesters U. S. Naval Nuclear Power and Prototype Training Various Hil'itary schools encompassing the operation, maintenance, and troubleshooting of specific nuclear plant and submarine systems and equipme nt.
i Professional Atfiliations:
Pi Tau Sigma, National Honorary Mechanical Engineering Fraternity American Nuclear Society Standards Committee ANS 56.3.
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' o EXHIBIT 2-B 1.
EQUIPMENT QUALIFICATION FILES o
EQ File No. 113-06-01 113-03-01 113-17-01 113-18-01 113-20-01 171-01-01 172-01-01 248-36-01 252-38-01 2.
USNRC DOCUMENTS Regulatory Guide 1.75, Rev. 2, dated September 1978, " Physical o
Independence of Electtic Systems" Regulatory Guide 1.89, Rev.1, dated June 1984, " Environmental o
Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants" Regulatory Guide 1.97, Rev. 3, dated May 1983, " Instrumentation o
for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident" NUREG-0588, Rev.1, dated July 1981, " Interim Staf f Position on o
Environmental Qualification of Sarety-Releted Electrical Equip-cient" o
NUREC-0896, Supplement No. 5, dated July 1986, " Safety Evaluation Report Related to the Operation of Seabrook Station, Units 1 and 2" o
10CFR50.49 4.
INDUSTRY STANDARDS o
IEEE Standard 323-1974, " Qualifying Class IE Equipment for Nuclear Power Generating Stations" ICCU Standard 323A-1975, Supplement to forward of IEEE o
Standard 323-1974 o
IEEE Standard 384-1974, " Independence of Class IE Equipment and Circuits"
.. o DOLKETED USNRC CERTIFICATE OF SERVICE I,
Thomas G.
- Dignan, Jr., one of the attorneys f16 AME 27 40 39 Applicants herein, hereby certify that on August 22, 1986, I made service of the within document by depositing c pgggp g q3g; thereof with Federal Express, prepaid, for delivery hiqR;,& SLFv:Cf.
where indicated, by depositing in the United States ma11SRANCH first class postage paid, addressed to):
Administrative Judge Sheldon J.
Stephen E. Merrill, Esquire Wolfe, Esquire, Chairman Attorney General Atomic Safety and Licens,ing George Dana Bisbee, Esquire Board Panel Assistant Attorney General U.S. Nuclear Regulatory Office of the Attorney General Commission 25 Capite) Street East West Towers Building Concord, NH 03301-6397 4350 East West Highway Bethesda, MD 20S14 Dr. Emmeth A. Luebke Dr. Jerry Harbour Atomic Safety and Licensing Atomic Safety and Licensing Board Panel Board Panel U.S. Nuclear Regulatory U.S.
Nuclear Regulatory Commission Commission East West Towers Building East West Towers Building 4350 East West Highway 4350 East West Highway Bethesda, MD 20814 Bethesda, MD 20814 Robert Carrigg, Chairman Richard A. Hampe, Esquire j
Board of Selectmen Hampe and McNicholas i
Town Office 35 Pleasant Street l
Atlantic Avenue Concord, NH 03301 North Hampton, NH 03862 Andrea C.
Eerster, Esquire Sherwin E. Turk, Esquire Diane Curran, Esquire Office of the Executive Legal Harmon & Weiss Director Suite 430 U.S.
Nuclear Regulatory Commission 2001 S Street, N.W.
Tenth Floor Washington DC 20009 7735 Old Georgetown Road Bethesda, MD 20814
- Atomic Safety and Licensing Robert A. Backus, Esquire Appeal Board Par.el Backus, Meyer & Solomon U.S. Nuclear Regulatory 116 Lowell Street Commission P.O. Box 516 Washington, DC 20555 Manchester, NH 03105
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- Atomic Safety and Licensing Mr. Ed Thomas Board Panel FEMA, Region I U.S. Nuclear Regulatory 442 John W. McCormack Post Commission Office and Court Hottse Washington, DC 20555 Post Office Square Boston, MA 02109 Paul McEachern, Esquire Carol S.
Sneider, Esquire Matthew T.
Brock, Esquire Assistant Attorney General Shaines & McEachern Department of the Attorney General 25 Maplewood Avenue One Anhburton Place, 19th Floor P.O.
Box 360 Boston, MA 02108 Portsmouth, NH 03801 Gary W. Holmes, Esquire Mr. Peter J. Matthews Holmes & Ells Mayor 47 Winnacunnet Road City Hall Hampton, NH 03841 Newburyport, MA 01950 Mrs. Sandra Gavutis Mr. Calvin A. Canney Chairman, Board of Selectmen City Manager RFD 1 - Box 1154 City Hall Kensington, NH 03827 126 Daniel Street Portsmouth, NH 03801
- Senator Gordon J. Humphrey Mr. Angie Machiros U.S.
Senate Chairman of the Washingtor, DC 20510 Board of Selectmen (Attn:
Tom Burack)
Town of Newbury Newbury, MA 01950
- Senator Gordon J.
Humphrey Mr.
J. P. Nadeau 1 Pillsbury Street Selectmen's Office Concord, NH 03301 10 Central Road (Attn:
Herb Boynton)
Rye, NH 03870 Mr. Thomas F.
Powers, III Mr. William S.
Lord Town Manager Board of Selectmen Town of Exeter Town Hall 10 Front Street Friend Street Exeter, NH 03833 Amesbury, MA 01913 H.
Joseph Flynn, Esquire Brentwood Board of Selectmen Office of General Counsel PFD Dalton Road Federal Emergency Management Brentwood, NH '03833 Agency 500 C Street, S.W.
Washington, DC 20472
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-Philip Ahrens, Esquire Judith H.
Mizner, Esquire Assistant Attorney Ge neral Silyergate, Gertner, Baker Department of the Attorney General Fine, Good & Minner Augusta, ME 04333 88 Broad Street Boston, MA 02110
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b [fh Thoritus'G. Dignan, Jr.
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- = Ordinary U.S. First Class M' ail.
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