ML20217Q663
| ML20217Q663 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 04/30/1998 |
| From: | Stephen Dembek NRC (Affiliation Not Assigned) |
| To: | Carns N NORTHEAST NUCLEAR ENERGY CO. |
| References | |
| TAC-M98916, NUDOCS 9805080273 | |
| Download: ML20217Q663 (12) | |
Text
Mr. Neil S. Carns 2 i
S:nior Vics Presidint April 30, 1998
?and Chisf Nuctrar Offic:r,
N North' east Nuclear Energy Com'pany l
c/o Ms. Patricia A. Loftus l
-Director - Regulatory Affairs'
^ P. O. Box.128
. Waterford, CT 06385
SUBJECT:
REVISIONS TO TECHNICAL SPECIFICATIONS BASES, MILLSTONE NUCLEAR POWER STATION, UNIT 1 (TAC NO. M98916)
L 4
7
~
3
Dear Mr. Carns:
l e
By letters dated April 29 and Septembeh 3,1997, Northeast Nuclear Energy Company (NNECO) provided the NRC with changes to Technical Specifications (TSs) Bases Sections 3.6,3.7.B, and 4.7. NNECO provided the TS Bases pages to the NRC for
. information only.'
T As you are aware, the TS Bases are'not part of the TS'as defined by 10'CFR 50'36.
. Changes to the TS Bases may voluntarily be made in accordance with the provisions of
- 10 CFR 50.59. -Should the proposed change involve an unrevi,ewed safety question
.j pursuant to 10 CFR 50.59(a)(2), or involve a change in the interpretation of g
implementation of the TS (i.e.,-constitute a TS change), then the proposed change is to be provided.to the staff pursuant to the provisions of 10 CFR 50.59(c) and 10.CFR 50.90 for.
- prior NRC review and approval.
The TS Bases you provided are hereby returned to you and should be inserted in the TS to J ensure the NRC staff and NNECO have identical TS Bases pages. The staff did not
- perform an evaluation of your TS Bases revisions and staff concurrence with the revisions is not implied by this letter. The staff may review the evaluations that support these TS Bases revisions during the next inspection of Millstone Unit 1's implementation of 10 CFR 50.59.
Sincerely, l signed by:
Origina h50 Stephen Dembek, Project Manager Special Projects Office.- Licensing DO 5000245 P
POR Office of Nuclear Reactor Regulation Docket ' o. 50-245
~
N
. Endlosure: As stated l
. cc w/ encl:' i e next page 9
DISTRIBUTION:
%; PUBLIC PMcKee ACRS
?
SDembek GHill (2)
I kf* (S g g$ @ @f
' SFO-L Reading ~
LBerry _.
JDurr, RI
- SPO Reading -
WBeckner, TSB
- WTravers '
\\ l DOCUB5ENT NAME: G:\\DEMBEK\\M98916.LTR h
.; ; To rocs 4ve e copy of thle document,insgeste in the box: 'CJ fopy without attachment / enclosure "E" = Copy with attachment / enclosure
- N*,=,Ng copy 0FFICE' SPO-L:PM
,_l SPO-L:LA-V,/D f/
SPLS:5C l
ENCO BC(A) i; d
SPO J M //
l m
NAME S0enbek WhX.LBerry f{J /
LMarsh I'
ESullivan 6M '
PMc/e5'I 10/7 /97 I
10/$f97 I T I.st i 1$/(g97 0
@ /9T DATE 10/K/97 M"
3 l
, ; $0[y-0FFICIAL RECORD COPY n.a o
REVISED TECHNICAL SPECIFICATIONS BASES FACILITY OPERATING LICENSE NO. DPR-21 i
DOCKET NO. 50-245 Replace the following pages of the Appendix A, Technical Specifications Bases, with the attached pa29s. The revised pages contain vertical lines indicating the areas of change.
Remove Insert B'3/4 6-1 B_3/4 6-1
~B 3/4 7-5 B 3/4 7-5 8 3/4 7-6 B 3/4 7-6 B 3/4 7-7 B 3/4 7-7 B 3/4 7-8 B 3/4 7-8 B 3/4 7-9 B 3/4 7-9 B 3/4 7-10 B 3/4 7-10 B 3/4 7-11 B 3/4 7-11 i
9 3.6 PRIMARY SYSTEM BOUNDARY BASES A.
Thermal Limitations The reactor vessel. has been analyzed for thermal conditions encountered during heatup and cooldown operations conducted within the specified differential temperatures and rate limits. Heatup and cooldown operations throughout plant life at uniform rates of 100*F/hr were considered in the temperature range of 100*F to 546'F and were shown to be within the requirements for stress intensity and fatigue limits of Section III of the ASME Boiler and Pressure Vessel Code (1965 Edition). The allowable number of reactor vessel closure bolt pre-loading cycles is 80. The allowable number of heatup and cooldown cycles during the service lifetime is 260.
B.
Pressurization _Temnerature B.I.a Inservice Pvdrostatic and Leak Tests Operating limits for the reactor vessel pressure and temperature during normal heatup and cooldown, and during inservice hydrostatic and leak testing were established using 10 CFR 50 Appendix G, January 1992, and Appendix G, of the 1992 Addenda of Section XI of the ASME Boiler and Pressure Vessel Code. For the purpose of this analysis the reference temperature, RTuor, of the reactor vessel material is based on the impact test data taken in accordance with the requirements of the Code to which the reactor vessel was designed and manufactured. For the reactor vessel beltline region, a RTuor of 137.2*F was calculated for 32.0 EFPY based on surveillance capsule 300', results.
For the remainder of the reactor vessel, a RTuor of +40'F was used as this is the maximum NOT temperature (NDTT) permitted by the reactor vasal purchase specification.
Figure 3.6-1 establishes the minimum temperature for hydrostatic and leak testing required by the ASME Boiler and Pressure Vessel Code,Section XI.
Test pressures for inservice hydrostatic and laak testing required by ASME Section XI are a function of the testing temperatures and-the component material. Accordingly, the maximum hydrostatic test pressures will be 1.1 times i.he operating pressure or about 1139 psig for a reactor coolant temperature greater than 100*F.
Figure 3.6-2 provides limitations for plant heatup and cooldown when the reactor is agi critical. The thermal limitations consider maximum heatup and cooldown rates of 100'F/hr in any one-hour period.
Figure 3.6-3 establishes operating limits when the core is critical.
These limits include a margin of 40'F as required by 10 CFR 50 Appendix G.
Fast neutron irradiation affects the fracture toughness of the reactor vessel material.
In order to prevent non-ductile failure, two types of information are needed:
a) a relationship between the r.hange in RTuor end the accumulated fast neutron fluence, and glstoneUnit1 B3/46-1 Revised by NRC Letter
3.7 CONtAINNENT SYSTEMS BASES containment is normally slightly pressurized during periods of reactor operation assuring no air in-leakage through the primary containment. However, at least once a week, the oxygen concentration will be determined as added assurance.
- 7..
Containment Hiah-Ranae Radiation Monitors The containment high-range radiation monitors (CHRRM) ensure that adequate information is available to monitor and assess containment radiation levels during and following an accident. Area Radiation Monitor (ARM) #12 at the control rod drive removal hatch will be utilized as the preplanned alternate method of monitoring j
containment high-range radiation in the event that the CHRRMs are not available. Due to inaccuracies' induced by cable bias under specific conditions, the CHRRMs do not meet the low and accuracy requirements specified in Regulatory Guide 1.97, " Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant and Environs During and Following an Accident". Under conditions of high temperature and low radiation, the operator would not rely on the CHRRMs for indication of containment radiation to assess an accident. Since these inaccuracies do not have an a?ierse affect on the ability of the operator to assess accidents,'the intent of Regulatory Guide 1.97 is met. This capability is consistent with the recommendations of NUREG-0737, " Clarification of TMI Action Plan Requirements," dated November, 1980.
8.
Containment Pressure Monitors The containment pressure etnitor ensures additional information is available to monitor and assess the coritainment pressure from
-5 psig to at least 3 times containment design pressure. This capability is consistent with the recommendations of NUREG-0737,
" Clarification of TMI Action Plan Requirements," dated November 1980.
8.
Standby Gas Treatment Systems The standby gas treatment system is designed to filter and exhaust the reactor building atmosphere to the stack during secondary containment isolation conditions.
Both standby gas treatment system fans are designed to automatically start upon automatic containment isolation and l to maintain the reactor building pressure to the. design negative pressure so that all leakage should be in-leakage.
Each of the two fans has 100 percent capacity.
High efficiency particulate absolute (HEPA) filters are installed before and after the charcoal adsorbers to minimize potential rele'ase of NILLSTONE UNIT.'1 B3/47-6
[y*
d Le, lh5 d AP i
l 3.7 CONTAIMENT SYSTEMS BASES particulates to the environment and to prevent clogging of the iodine adsorbers. Operation of the fans significantly higher than design flow (11210 cfs) could change the removal efficiency of the HEPA filters and charcoal adsorbers. HEPA filters are purchased at a 99.97% removal efficiency for particulate larger than 0.3 micron.
The HEPA in-place DOP leak test. ensures filter bypass leakage is less than 15 (removal 199%). The'1eak test is required for the annual surveillance, when a HEPA filter is replaced, or after structural maintenance that could affect the HEPA or seal integrity. This test is not required when a charcoal sample is taken, or charcoal trays are replaced.
The charcoal adsorber in-place halogenated hydrocarbon leak test ensures filter bypass leakage is less than 1% remoyal 199%, and must be performed for the annual surveillance,(when charcoal) trays are installed, or after structural maintenance that could affect the charcoal cell or seal integrity. This test is not applicable to HEPA filter replacement.
The laboratory analysis:of used (exposed or weathered) charcoal should indicate a methyl iodide removal efficiency of at least 95% to ensure sufficient adsor> tion margin until the next efficiency test is performed.
Installing new ciarcoal with a methyl iodide removal efficiency 199%, in lieu of sampling the used charcoal, satisfies the surveillance requirement to perform a sample analysis, since it provides the same degree of assurance that charcoal efficiencies are above the efficiency limit.
Only one of the two standby gas treatment trains is needed to clean up the reactor building atmosphere upon containment isolation.
If one train is found to be' inoperable, there is no immediate threat to the containment system performance, and reactor operation or refueling operation may continue while repairs are being made. During RUN, STARTUP/ HOT STANDBY and HOT SHUTDOWN, OPERABILITY of the standby gas treatment system is required. Standby gas treatment system OPERABILITY is also required during COLD SHUTDOWN or REFUELING when situations exist where a significant release of fission products can be postulated, such as moving the fuel cask, irradiated fuel or other loads in containment; or when performing CORE ALTERATIONS or operations with a potential for draining the reactor vessel when the vessel contains irradiated fuel.
During a REFUELING OUTAGE, wher. reactor c.colant temperature is less than or equal to 212' F and secondary contairment integrity is required, two off-site power sources (345 kV or 23 kV's and one emergency power source would provide an adequate and reliable source of power and allow diesel or gas turbine preventative maintenance.
Likewise, one source of offsite power (345 kV or 23 kV) and two emergency power sources provide an adequate and reliable source of power.
-NILLSTONE UNIT 1 83/47-6 RMt,ef I
4 3.7 CONTAINNENT SYSTEMS BASES C.-
Secondary Containment The secondary containment is' designed to minimia any ground' level release of radioactive materials which might result from a serious accident.
The reactor building provides secondary containment during reactor operation, when the drywell is sealed and in service; the reactor 1
building provides primary containment when the reactor is shutdown and the drywell is open, as during refueling. Because the secondary
)
containment is an integral part of the complete containment system,..
secondary containment integrity is required at all times that primary l
containment is recuired.
Secondary containment integrity is also required when activities-aaving the potential cf significant fission products release, such as movement of the fuel cask, irradiated fuel, or other loads in containment are performed.
Administrative controls ensure that loads moved in containment, which may result in significant release of fission products, are evaluated to determine if secondary containment is required.
If secondary containment is inoperable in RUN, STARTUP/ HOT STANDBY or HOT SHUTDOWN, it must be restored to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> time provides a period of time to correct the problem that is commensurate with the importance of maintaining secondary containment integrity in RUN,.
STARTUP/ HOT STANDBY or HOT SHUTDOWN. This time period also ensures that the
. probability of an accident (requiring secondary containment OPERABILITY) occurring during periods where secondary containment is inoperable is minimal.
If secondary containment cannot be restored to OPERABLE status within the required time, the plant must be brought to a condition in which the LCO does not apply. An additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> are allowed to achieve this condition. The allowed times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
D.
Primary Containment Isolation Valves Double isolation valves are provided on lines penetrating the primary containment. Closure of one of the valves in each line would be sufficient to maintain the integrity of the pressure suppression system.
Automatic initiation is required to minimize the potential leakage paths from the containment in the event of a loss of coolant accident.
- gIgLSTONEUNIT1 B3/47-7 ed[ydRCLettr
4.7 CONTAINNENT SYSTEMS BASES A.
Primary Containment The water in the suppression chamber is used only for cooling in the event of an accident; i.e., it is not used for normal operatinn; therefore, a once per shift check of the temperature, volume, and differential pressure is adequate to assure that adequate heat removal capability is present.
The interior of the suppression chamber is coated with paint to prevent rusting. The inspection'of the coating during refueling outage assures the coating is intact.
The interior of the drywell is coated with paint. The drywell interior l
surfaces will be inspected during each refueling outage to determine whether the coating is deteriorating such that it could be stripped off in the event of a LOCA. Such an event could lead to the release of debris l
that may affect operation of the ECCS.
The primary containment preoperational test pressures are based upon the calculated primary containment pressure response in the event of a loss of a coolant accident. The peak drywell pressure would be about 43 psig which would rapidly reduce to 25 psig within 30 seconds following the pipe break.
Following the pipe break, the suppression chamber pressure rises to 25 psig within lo reconds, equalizes with drywell pressure and therefore rapidly decays with the drywell pressure decay.
I The design basis loss of coolant accident was evaluated at the primary containment maximum allowable accident leak rate of 1.2% day at 43 psig.
The analysis showed that with this leak rate and a standby gas treatment system filter efficiency of 90% for halogens, 90% for particulates, the maximum whole body and maximum thyroid dose are within 10CFR100 limits.
Additional margin is achieved by establishing the allowable o>erational leak rate.
The operational limit is derived by multiplying tie allowable test leak rate by 0.75, thereby providing a 25% margin to allow for leakage deterioration which may occur during the period between leak rate tests.
The periodic retest schedule for with the requirements of 10CFR50, performing Type A tests is consistent Appendix J, paragraph III.D.
If two consecutive periodic Type A tests fail to meet acceptance criteria, 10CFR50, Appendix J, paragraph III.A.6(b) requires an increase in Ty>e A test frequency. However, if Type B and C leakage rates constitute t1e identified contributor to the failure of the two Type A tests, an exemption to the increased Ty e A test requirements of 10CFR50, Appendix J guidanceofparagraphIII.A.6b)mayberequestedbyfollowingthe Information Notic No. 85-71.
The combined leakage rate for all penetrations and valves subject to Type B and C tests is limited to 60% of the allowable test leak rate in accordance with 10CFR50 Appendix J.
NILLSTONE UNIT 1 B3/47-8 Revised by NRC Letter A ru 30, 1998 P
dated
~
l 4.7 CONTAINMENT SYSTEMS BASES The penetration and air purge piping leakage test frequency, along with the containment leak tests which is performed at a test pressure of at Whenever a doubl ), is adeq,ute to allow detection of leakage trends.
least 43 psig (P equipmenthatches,gasketedpenetration($rimarycontainmenthead e-and the suppression c amber access hatch isbroken and remade, the space between the gaskets is pressurized to) determine that the seals are performing properly. The test pressure of at least 43 psig is consistent with the accident analyses and the. maximum preoperational leak rate test pressure.
l Personnel air lock door seal testing is performed in accordance with 10CFR50, Appendix J requirements.
Monitoring the nitrogen makeup requirements of the inerting system provides a method of observing leak rate trends and would detect gross leaks in a very short time. This equi from service for test and maintenance,pment must be periodically removed but this out-of-service time will be
'kept to a practical minimum.
Surveillance of the suppression chamber-drywell vacuum breakers consists of operability checks, calibration of instrumentation and inspection of the valves.
The monthly operability tests are performed to check the capability of the disc to open and close and to functionally test the position indication system. This test frequency is justified based on previous experience and the fact that these valves are normally closed and are only open during tests or accident conditions.
The refueling outage surveillance tests are performed to check that the valve will perform properly during the accident condition and to verify the calibration of the position indication system. Measuring the force required to lift the valve assures that the valve will function properly during an accident.
Inspection of a select number of valves during each refueling outage assures that deterioration of the valve internals or misalignment of the disc does not impair the proper operation of the valve.
This test interval is based on equipment quality and previous equipment experience.
B.
Standby Gas Treatment System and C.
Secondary Containment Pressure drop across the combined HEPA filters and charcoal adsorbers of less than 7 inches of water,d adsorbers are not clogged by excessive at the system design flow rate, will indicate that the filters an amounts of foreign matter. Heater capability, pressure drop should be determined at least once per operating cycle to show system performance capability.
gLSTONE LMIT 1 83/47-9 edgRgeg
i O
4.7 CONTAINNENT SYSTEMS BASES The frequency of tests and sample analysis are necessary to show that the HEPA filters and charcoal adsorbers can perform as evaluated.
In-place leak tests of the charcoal adsorbers with halogenated hydrocarbon.
refrigerant shall be performed in accordance with ANSI /ASME N510-1980.
Iodine removal efficiency tests shall follow ASTM D3803-1989, as well as the conditions and requirements specified in the technical specification.
The charcoal adsorber efficiency test procedures should allow for the removal of one adsorber tray, emptying of one bed from the tray, mixing.
the adsorbent thoroughly and obtaining at least two samples or removal of a test canister.
Each sample should be at least two inches in diameter and a length equal to the thickness of the bed.
If test results indicate removal efficiency less than 95%, all adsorbent in the system shall be replaced with an adsorbent qualified to ANSI /ASME N509-1980. The replacement tray for the adsorber tray removed for the test should meet the same adsorbent quality. The charcoal filter air flow distribution test is required only if inlet piping, filter plenum, or outlet piping geometry is altered, or if maintenance is performed that alters flow distribution. The flow distribution test is no't applicable to the HEPA filters.
In-place leak tests of the HEPA filters with DOP aerosol shall be i
performed in accordance with ANSI /ASME N510-1980. Any HEPA filters found defective shall be replactJ with filters qualified pursuant to ANSI /ASME N509-1980. Although the SGTS design flow rate is 1100 SCFM, the DOP test at reduced flow rate is actually more sensitive because diffusion is the primary mechanism of small particle collection. The lower limit for test flow rate (500 SCFM) is based on test instrument sensitivity.
All elements of the heater should be demonstrated to be functional and operable during the test of heater capacity. Operation of the heaters i
will prevent moisture buildup in the filters and adsorber system.
Demonstration of the automatic initiation capability and operability of filter cooling is necessary to assure system performance capability.
Verifying that secondary containment access doors are closed ensures that
]
the infiltration of outside air of such a magnitude as to prevent maintaining the desired negative pressure does not occur. Maintaining secondary containment OPERABILITY, in accordance with the definition for SECONDARY CONTAINMENT INTEGRITY, requires verifying at least one door in each access opening is closed. The access openings covered by Surveillance Requirement 4.7.C.I.b are the normal and emergency double-door accesses on Reactor Building elevation 14' 6" and the railroad track bay access doors. The monthly frequency for the surveillance requirement is considered adequate in view of the other indications of door status that are available.
MILLSTONE UNIT 1 8 3/4 7-10 Revised by NRC Letter om dated April 30, 1998
~~
4.7 CONTAINMENT SYSTEMS BASES.
D.
Primary Containment Isolation Valvas Those large pipes comprising a portion of the reactor coolant system, whose~ failure could result in uncovering the reactor core, are supplied with automatic isolation valves (except those lines needed for emergency core' cooling system operation or containment cooling). The closure times specified in the Technical Requirements Manual (TRM) are adequate to prevent loss more coolant from the circumferential rupture of any of these lines outside the containment than from a steam line rupture.
Therefore, this isolation valve closure time is sufficient to prevent uncovering the core.
In order to assure that the doses that may result from a steam line break do not exceed the 10 CFR 100 guidelines, it is necessary that no fuel rod perforation resulting from the accident occur prior to closure of the-main steam line isolation valves. Analyses suggest that fuel rod clad-ding perforations would be avoided for main steam valve closure times, including instrument delay, as long as 10.5 seconds. However, for added margin, the Technical Specifications require a valve closure time of not greater than five seconds.
The TRM lists the automatic primary containment isolation valves and their respective closure times. Also, remote manual or manual primary containment isolation valves that could be opened at power are listed.
The TRM includes administrative controls for opening normally closed remote manual and manual primary containment isolation valves on an intermittent basis during power operation. Administrative controls ensure that containment integrity is maintained during the time that the valves are open.
The addition, deletion, or modification of any primary containment isolation valve or related information is reviewed under 10 CFR 50.59 and is approved by the Plant Operations Review Committee.
For reactor coolant system temperatures less than 212*F, the containment could not become pressurized due to a loss of coolant accident. These valves are highly reliable, have low service requirement and most are normally closed. The initiating sensors and associated trip channels are also checked to demonstrate the capability for automatic isolation. Ref.
Section 6.2 and Tables 6.2-4 and 5 of the UFSAR. The test interval of onceperoperatingcyc1gforautomaticinitiationresultsinafailure probability of 1.1 X 10' that a line will not isolate. More frequent testing for valve operability results in a more reliable system.
The main steam line isolation valves are functionally tested on a more frequent interval to establish a high degree of reliability.
The containment is ' penetrated by a large number of small diameter instru-ment lines. A program for periodic testing and examination of the excess flow check valves in these lines is performed in accordance with Specification 4.7.D.I.b.
MILLSTONE UNIT 1 B 3/4_7-11 Revised by NRC Letter dated April 30, 1998
Northeast Nuclear Energy Company Millstone Nuclear Power Station Unit 1 cc:
Lillian M. Cuoco, Esquire Mr. Wayne D. Lanning Senior Nuclear Counsel Deputy Director of Inspections Northeast Utilities Service Company Special Projects Office P. O. Box 270 475 Allendale Road Hartford, CT 0610 -0270 King of Prussia, PA 19406-1415 Mr. Kevin T. A. McCarthy, Director Mr. F.- C. Rothen Monitoring and Radiation Division Vice President - Work Services Department of Environmental Northeast Nuclear Energy Company Protection P. O. Box 128 79 Elm Street Waterford, CT 06385 Hartford, CT 06106-5127 Ernest C. Hadley, Esquire Mr. Allan Johanson, Assistant 1040 B Main Street Director P. O. Box 549 Office of Policy and Management West Wareham, MA 02576 Policy Development and Planning Division Mr. John Buckingham 450 Capitol Avenue - MS# 52ERN Department of Public Utility Control P. O. Box 341441 Electric Unit Hartford, CT 06134-1441 10 Liberty Square New Britain, CT 06051 Regional Administrator Region i Mr. David Amerine U.S. Nuclear Regulatory Commission Recovery Officer - Nuclear 475 Allendale Road Engineering and Support King of Prussia, PA 19406 Northeast Nuclear Energy Company P. O. Box 128 First Selectmen Waterford, CT 06385 Town of Waterford Hall of Records Mr. B. D. Kenyon 200 Boston Post Road President and Chief Waterford, CT 06385 Executive Officer Northeast Nuclear Energy Company Mr. D. M. Goebel P. O. Box 128 Vice President - Nuclear Oversight Waterford, Connecticut 06385 Northeast Nuclear Energy Company P. O. Box 128 Senior Resident inspector Waterford, CT 06385 Millstone Nuclear Power Station c/o U.S. Nuclear Regulatory Mr. J. P. McElwain Commission Recovery Officer - Millstone Unit 1 P.O. Box $13 Northeast Nuclear Energy Company Niantic, CT 00357 P. O. Box 128 Waterford, CT 06385
Northeast Nuclear Energy Company Millstone Nuclear Power Station Unit 1 cc:
l The Honorable Terry Concannon Co-Chair Nuclear Energy Advisory Council Room 4035 l
Legislative Office Building l
Capitol Avenue l
Hartford, Connecticut 06106 Mr. Evan W. Woollacott Co-Chair Nuclear Energy Advisory Council i
128 Terry's Plain Road Simsbury, Connecticut 06070 Citizens Regulatory Commission ATTN: Ms. Susan Perry Luxton 180 Great Neck Road Waterford, Connecticut 06385 Deborah Katz, President Citizens Awareness Network P. O. Box 83 Shelburne Falls, MA 03170 Little Harbor Consultants, Inc.
Millstone -ITPOP Project Office P. O. Cox 0630 Niantic, Connecticut 06357-0630 l
Mr. Daniel L. Curry Project Director Parsons Power Group Inc.
2675 Morgantown Road Reading, Pennsylvania 19607 Mr. Don Schopfer Verification Team Manager Sargent & Lundy 55 E. Monroe Street Chicago, Illinois 60603