Table Of ContentSAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION
RELATED TO AMENDMENT NO. 242 TO RENEWED FACILITY OPERATING
LICENSE NO. NPF-49
DOMINION NUCLEAR CONNECTICUT, INCORPORATED
MILLSTONE POWER STATION, UNIT 3
DOCKET NO. 50-423
TABLE OF CONTENTS
1.0 INTRODUCTION………………………………………………………...................................- 1 -
1.1 Application.................................................................................................................- 1 -
1.2 Background...............................................................................................................- 2 -
1.3 Licensee’s Approach..................................................................................................- 3 -
1.4 Plant Modifications.....................................................................................................- 3 -
1.5 Method of NRC Staff Review…………………………...…….……………………………- 4 -
2.0 EVALUATION..................................................................................................................- 6 -
2.1 Materials and Chemical Engineering..........................................................................- 6 -
2.2 Mechanical and Civil Engineering............................................................................- 28 -
2.3 Electrical Engineering..............................................................................................- 47 -
2.4 Instrumentation and Controls...................................................................................- 59 -
2.5 Plant Systems..........................................................................................................- 61 -
2.6 Containment Review Considerations.......................................................................- 96 -
2.7 Habitability, Filtration, and Ventilation....................................................................- 107 -
2.8 Reactor Systems....................................................................................................- 115 -
2.9 Source Terms and Radiological Consequences Analyses.....................................- 188 -
2.10 Human Performance............................................................................................- 227 -
3.0 FACILITY OPERATING LICENSE AND TECHNICAL SPECIFICATION CHANGES...- 232 -
4.0 REGULATORY COMMITMENTS................................................................................- 242 -
5.0 RECOMMENDED AREAS FOR INSPECTION............................................................- 242 -
6.0 STATE CONSULTATION............................................................................................- 242 -
7.0 ENVIRONMENTAL CONSIDERATION.......................................................................- 242 -
8.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION..............- 242 -
9.0 CONCLUSION.............................................................................................................- 247 -
10.0 REFERENCES.......................................................................................................... - 248 -
Attachment: List of Acronyms
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION
RELATED TO AMENDMENT NO. 242 TO RENEWED FACILITY OPERATING
LICENSE NO. NPF-49
DOMINION NUCLEAR CONNECTICUT, INCORPORATED
MILLSTONE POWER STATION, UNIT 3
DOCKET NO. 50-423
1.0 INTRODUCTION
1.1 Application
By letter dated July 13, 2007,1 as supplemented by additional letters,2 Dominion Nuclear
Connecticut, Inc. (DNC), licensee of Millstone Power Station, Unit 3 (MPS3), submitted the
application, “Dominion Nuclear Connecticut, Inc., Millstone Power Station Unit 3, License
Amendment Request, Stretch Power Uprate,” to the U.S. Nuclear Regulatory Commission
(NRC).
The proposed license amendment would allow an increase in the maximum authorized power
level from the current licensed thermal power (CLTP) level of 3,411 megawatts thermal (MWt) to
3,650 MWt, and make changes to the facility operating license and technical specifications, as
necessary, to support operation at the stretch power level, which is an increase of
approximately 7 percent. The proposed increase in power level is considered a stretch power
uprate (SPU).
The supplemental letters dated January 10 (4 letters), January 11 (4 letters), January 14,
January 18 (5 letters), January 31, February 25 (2 letters), March 5, March 10 (2 letters), March
25, March 27, April 4, April 24, April 29, May 15, May 20, May 21, July 10, and July 16, 2008,
provided additional clarifying information that did not expand the scope of the initial application
and did not change the NRC staff’s original proposed no significant hazards consideration
determination as published in the Federal Register on January 15, 2008 (73 FR 2549).
1
DNC Letter (07-450) to the NRC, “Dominion Nuclear Connecticut, Inc., Millstone Power Station Unit 3 License
Amendment Request, Stretch Power Uprate,” dated July 13, 2007 (Agencywide Documents Access and
Management System (ADAMS) Accession No. ML072000386).
2
Supplemental Letters dated: July 13, 2007 (ML072000281); September 12, 2007 (ML072570061); November 19,
2007 (ML073230976); December 13, 2007 (ML073480240); December 17, 2007 (ML073520051); January 10, 2008
(ML080100600, ML080100604, ML080100606, ML080100611); January 11, 2008 (ML080110695, ML080140077,
ML080170495, ML080580476); January 14, 2008 (ML080140570); January 18, 2008 (ML080220506, ML080220527,
ML080220530, ML080220541, ML080280375); January 31, 2008 (ML080320308); February 25, 2008 (ML0805
60392, ML080560615); March 5, 2008 (ML080660108); March 10, 2008 (ML080710377, ML080710391); March 25,
2008 (ML080850894); March 27, 2008, ((ML080880268); April 4, 2008 (ML081430014); April 24, 2008
(ML081150679); April 29, 2008 (ML081200643); May 15, 2008 (ML081360625); May 20, 2008 (ML081420443);
May 21, 2008 (ML081420824); July 10, 2008 (ML081930274); and July 16, 2008 (ML081990112).
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1.2 Background
MPS3 uses a 4-Loop, closed cycle, pressurized-water reactor (PWR) type nuclear steam supply
system (NSSS) furnished by Westinghouse Electric Corporation and a turbine-generator
furnished by the General Electric Company (GE). The architect-engineer was Stone & Webster
Engineering Corporation. MPS3 has a sub-atmospheric reactor containment.
The site, approximately 500 acres in area, is on the north shore of Long Island Sound and on
the east side of Niantic Bay. It is located in the Town of Waterford, Connecticut, about 3.2 miles
west-southwest of New London and about 40 miles southeast of Hartford. The surrounding
area is primarily residential with some commercial and industrial uses.
The construction permit for MPS3 was issued on August 9, 1974. The full-term operating
license, as well as the full-power license, were issued on January 31, 1986, for operation at
3,411 MWt. MPS3 was at 100 percent power on April 17, 1986, and entered commercial
operation on April 23, 1986.
In 2001, Millstone Power Station (MPS), Units 1, 2 and 3 operating licenses were transferred
from Northeast Nuclear Energy Company to DNC. DNC is an indirect wholly-owned subsidiary
of Dominion Energy, which is in turn owned by Dominion Resources, Inc. (DRI). Virginia Power,
which is the licensed owner and operator of the North Anna and Surry Nuclear Stations, is also
a subsidiary of DRI.
DNC filed for renewal of the MPS2 and MPS3 operating licenses in January 2004. In July 2005,
the NRC issued NUREG-1437, “Generic Environmental Impact Statement for the Renewal of
Nuclear Power Plants,”3 Supplement 22, for license renewal for the two units. The staff’s review
and its acceptance of the license renewal application (LRA) are documented in staff’s SE report,
NUREG-1838.4 The renewed operating license for MPS2 now expires on July 31, 2035, while
the MPS3 renewed operating license now expires on November 25, 2045. MPS1 permanently
ceased operation on July 21, 1998, and is currently being decommissioned.
A new permissive (P-19) will be added to monitor low reactor coolant system (RCS) pressure
during the fall 2008 refueling outage and before implementation of the SPU amendment. The
P-19 permissive is designed to permit the cold leg injection valves to open automatically upon
receiving a safety injection signal. The permissive will be derived utilizing the existing low
pressurizer pressure reactor trip two out of four bistable trip logic and will have the same set
point as that function. Within the MPS3 solid state protection cabinets, the signal will be
separated from the reactor trip function logic to develop the low RCS pressure, cold leg injection
permissive. The cold leg injection permissive relay contacts will be placed in series with the
safety injection relay contacts in the control logic for the cold leg injection valves, to permit them
to open automatically upon receiving both the safety injection signal and the cold leg injection
permissive. Using the low pressurizer pressure reactor trip bistable trip logic helps to maintain
diversity from the low pressurizer pressure safety injection bistable trip logic to the extent
possible.
3 ADAMS Accession No. ML051990002
4 “Safety Evaluation Report Related to the License Renewal of the Millstone Power Station, Units 2 and 3, Docket
Nos. 50-336 and 50-423, Dominion Nuclear Connecticut, Inc.,” (ADAMS Accession No. ML053270483).
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1.3 Licensee’s Approach
The licensee's application for the proposed SPU follows the guidance in the Office of Nuclear
Reactor Regulation’s (NRR’s) Review Standard (RS)-001, “Review Standard for Extended
Power Uprates,”5 to the extent that the review standard is consistent with the design basis of the
plant. Where differences exist between the plant-specific design basis and RS-001, the
licensee described the differences and provided evaluations consistent with the design basis of
the plant.
Since MPS3 has a renewed license, DNC performed an evaluation of the SPU impact on the
extended period of plant operation. The purpose of the evaluation was to identify which, if any,
structures, systems, and components (SSCs) warranted additional aging management action
because of new aging effects due to the changes in the operating environment resulting from
SPU or the addition of, or modification to, components relied upon to satisfy SPU operating
conditions. SSCs relied upon for achieving the license renewal scoping objectives were
evaluated within the structure or system that contains them. DNC also evaluated the potential
impact of the proposed SPU on license renewal Time-limited Aging Analyses (TLAA).
Specifically, the evaluation considered any new aging effects or increases in degradation rates
potentially created by the new SPU operating parameters.
The licensee reviewed the MPS3 design in accordance with the July 1981 Edition of the
“Standard Review Plan (SRP) for the Review of Safety Analysis Report for Nuclear Power
Plants,” dated July 1981,(NUREG-0800)6, Section 5.3.1, Rev. 1. As noted in the updated final
safety analysis report (UFSAR) Section 3.1, the design bases of MPS3 are measured against
the NRC General Design Criteria (GDC) for Nuclear Power Plants, of Title 10 of the Code of
Federal Regulations (10 CFR) Part 50, Appendix A, as amended through October 27, 1978.
The licensee plans to implement the SPU in one step. The licensee plans to make minor
modifications necessary to implement the SPU during the refueling outage (RFO) in fall 2008.
Subsequently, the plant will be operated at 3,650 MWt starting in Cycle 13.
1.4 Plant Modifications
The licensee has determined that several plant modifications are necessary to implement the
proposed SPU. The following is a list of modifications that the licensee proposes to complete
during the fall 2008 RFO:
(1) Replace the turbine for the main feedwater pump;
(2) Modify ductwork to provide additional ventilation cooling in the condensate pump
area for the turbine building heating, ventilation, and air-conditioning (HVAC)
system;
(3) Provide control building auto initiation of pressurized filtration following a control
building isolation (CBI) signal for the control building ventilation;
5
ADAMS Accession No. ML023610659
6 ADAMS Accession No. ML033580033
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(4) For the turbine generator, provide the following: (a) control valve position
demand against lift settings for the valve position cards; (b) changes to power
load imbalance circuits; (c) sensor rescaling for steam pressure changes; (d)
instrument scaling; and (e) main control board meter scale changes;
(5) Increase the piping design temperature between residual heat removal (RHR)
and component cooling water (CCW) heat exchanger for the CCW system;
(6) For the instrumentation and controls (I&C) systems, provide set point changes to
the following: (a) balance-of-plant (BOP) systems; (b) feedwater pump; (c)
pressurizer level control; (d) electronic filter on the T signal; (e) pressurizer
hot
relief tank (PRT) level alarm; (f) condenser steam dump trip valve control; and
(g) P-8 set point change;
(7) Pipe support modifications for the condensate system, feedwater system, CCW
system, and containment recirculation;
(8) Provide a permissive for opening cold leg injection valves for the emergency core
cooling system (ECCS);
(9) Provide instrument loop rescaling for the following: first stage turbine pressure;
and
(10) Deletion of automatic rod withdrawal capability for the rod control system.
The NRC staff’s evaluation of the licensee’s proposed plant modifications is provided in
Section 2.0 of this safety evaluation.
1.5 Method of NRC Staff Review
The NRC staff used previously-approved SPUs, along with RS-0017 for guidance. An extended
power uprate (EPU) review includes the following areas: materials and chemical engineering;
mechanical and chemical engineering; electrical engineering; I&C; containment review
considerations; habitability, filtration, and ventilation; reactor systems; source terms and
radiological consequences analyses; human performance; health physics; risk evaluation; and
power ascension and test plan. As described in this memorandum, an SPU includes the same
areas as an EPU except for health physics, risk evaluation, and power ascension and test plan.
The NRC staff reviewed the licensee's application to ensure that: (1) there is reasonable
assurance that the health and safety of the public will not be endangered by operation in the
proposed manner; (2) activities proposed will be conducted in compliance with the
Commission's regulations; and (3) the issuance of the amendments will not be inimical to the
common defense and security or to the health and safety of the public. The purpose of the NRC
staff’s review is to evaluate the licensee’s assessment of the impact of the proposed SPU on
design-basis analyses. The NRC staff evaluated the licensee’s application and supplements.
7 RS-001, "Review Standard for Extended Power Uprates," dated December 2003 (ADAMS Accession Number
ML033640024)
- 5 -
In areas where the licensee and its contractors used NRC-approved or widely accepted
methods in performing analyses related to the proposed SPU, the NRC staff reviewed relevant
material to ensure that the licensee/contractor used the methods consistent with the limitations
and restrictions placed on the methods. In addition, the NRC staff considered the affects of the
changes in plant operating conditions on the use of these methods to ensure that the methods
are appropriate for use at the proposed SPU conditions. Details of the NRC staff's review are
provided in Section 2.0 of this safety evaluation.
An audit of the Rod Withdrawal at Power (RWAP) overpressure analysis supporting the SPU
was also conducted. The result of the audit is discussed in section 2.8 of this safety evaluation.
Independent NRC staff calculations were performed in relation to the following topics:
• The staff performed an independent calculation of the end of life (EOL) upper-
shelf energy (USE) values for the MPS3 reactor vessel (RV) beltline materials
using the limiting 54 effective full-power year (EFPY) neutron fluence value for
the one-quarter of the RV wall thickness (1/4T) location for the SPU conditions.
• The staff performed an independent calculation of the EOL RTPTS values for
MPS3 using the 54 EFPY neutron fluence value for the clad-metal interface
location of the vessel at SPU conditions.
• The NRC staff used the RADTRAD computer code to perform independent
confirmatory dose evaluations.
The results of the calculations are discussed in Section 2.0 of this safety evaluation.
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2.0 EVALUATION
2.1 Materials and Chemical Engineering
2.1.1 Reactor Vessel Material Surveillance Program
Regulatory Evaluation
The RV material surveillance program provides a means for determining and monitoring the
fracture toughness of the RV beltline materials to support analyses for ensuring the structural
integrity of the ferritic components of the RV. The NRC staff’s review primarily focused on the
effects of the proposed SPU on the licensee’s RV surveillance capsule withdrawal schedule.
The NRC’s acceptance criteria are based on: (1) General Design Criterion (GDC)-14, insofar as
it requires that the reactor coolant pressure boundary (RCPB) be designed, fabricated, erected,
and tested so as to have an extremely low probability of rapidly propagating; (2) GDC-31,
insofar as it requires that the RCPB be designed with margin sufficient to assure that, under
specified conditions, it will behave in a nonbrittle manner and the probability of a rapidly
propagating fracture is minimized; (3) 10 CFR Part 50, Appendix H, which provides for
monitoring changes in the fracture toughness properties of materials in the RV beltline region;
and (4) 10 CFR 50.60, which requires compliance with the requirements of 10 CFR Part 50,
Appendix H. Specific review criteria are contained in SRP Section 5.3.1 and other guidance
provided in Matrix 1 of RS-001.
Technical Evaluation
The NRC’s regulatory requirements related to the establishment and implementation of a
facility’s RV materials surveillance program and surveillance capsule withdrawal schedule are
given in 10 CFR Part 50, Appendix H. By reference, 10 CFR Part 50, Appendix H invokes the
guidance in American Society for Testing and Materials (ASTM) Standard Practice E185,
“Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels.” ASTM
Standard Practice E185 provides guidelines for designing and implementing the RV materials
surveillance programs for operating light-water reactors, including guidelines for determining RV
surveillance capsule withdrawal schedules based on the vessel material predicted transition
temperature shifts (ΔRT ).
NDT
The surveillance capsule withdrawal schedule shown in Table 2.1.1-3 of the SPU Licensing
Report (LR) was prepared in terms of EFPY of plant operation with a projected design life of 32
EFPY. To date, three surveillance capsules were withdrawn, and the neutron fluence
projections were updated using the third surveillance capsule. The licensee stated that the
neutron fluence value obtained from the latest (third) surveillance capsule exceeded the
projected neutron fluence value for the license renewal period (54 EFPY) under the SPU
conditions. Therefore, the licensee concluded that the current surveillance capsule withdrawal
schedule is still valid for the SPU conditions, and it meets the intent of ASTM E185. Consistent
with the requirements specified in paragraph 7.6.2 of the ASTM E185, the licensee stated that
the next capsule (standby capsule Z) will be withdrawn when the projected neutron fluence
value will not exceed two-times the projected value at 54 EFPY. The licensee concluded that its
neutron surveillance program meets the requirements specified in 10 CFR Part 50, Appendix H,
and that this program adequately monitors neutron-induced embrittlement in low alloy steel RV
base metals and their associated welds.
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The staff reviewed the licensee’s RV surveillance program under SPU conditions and finds it
acceptable. This acceptance is based on:
(1) The neutron fluence value obtained from the latest surveillance capsule exceeds
the projected neutron fluence value for the license renewal period (54 EFPY)
under the SPU conditions. However, this value is less than two-times the
projected 32 EFPY vessel fluence, and therefore, it complies with the
requirement specified in paragraph 7.6.2 of the ASTM E185. Hence, the
implementation of SPU does not affect the surveillance capsule withdrawal
schedule.
(2) Consistent with the requirements of paragraph 7.6.2 of the ASTM E185, the
licensee stated that the next capsule (stand by capsule Z) will be withdrawn
when the projected neutron fluence value will not exceed two-times the projected
value at 54 EFPY.
(3) The licensee’s surveillance capsule program complies with the requirements
specified in 10 CFR Part 50, Appendix H and 10 CFR 50.60.
Conclusion
The NRC staff has reviewed the licensee's evaluation of the effects of the proposed SPU on the
RV surveillance withdrawal schedule and concludes that the licensee has adequately addressed
changes in neutron fluence and their effects on the schedule. The NRC staff further concludes
that the RV capsule withdrawal schedule is appropriate to ensure that the material surveillance
program will continue to meet the requirements of 10 CFR, Part 50, Appendix H, and 10 CFR
50.60, and will provide the licensee with information to ensure continued compliance with GDC-
14 and GDC-31 in this respect following implementation of the proposed SPU. Therefore, the
NRC staff finds the proposed SPU acceptable with respect to the RV material surveillance
program.
2.1.2 Pressure-Temperature Limits and Upper-Shelf Energy (USE)
Regulatory Evaluation
Pressure-temperature (P-T) limits are established to ensure the structural integrity of the ferritic
components of the RCPB during any condition of normal operation, including anticipated
operational occurrences (AOOs) and hydrostatic tests. The NRC staff’s review of P-T limits
covered the P-T limits methodology and the calculations for the number of EFPYs specified for
the proposed SPU, considering neutron embrittlement effects and using linear elastic fracture
mechanics. The NRC’s acceptance criteria for P-T limits are based on: (1) GDC-14, insofar as
it requires that the RCPB be designed, fabricated, erected, and tested so as to have an
extremely low probability of rapidly propagating fracture; (2) GDC-31, insofar as it requires that
the RCPB be designed with margin sufficient to assure that, under specified conditions, it will
behave in a nonbrittle manner and the probability of a rapidly propagating fracture is minimized;
(3) 10 CFR Part 50, Appendix G, which specifies fracture toughness requirements for ferritic
components of the RCPB; and (4) 10 CFR 50.60, which requires compliance with the
requirements of 10 CFR Part 50, Appendix G. Specific review criteria are contained in
SRP Section 5.3.2 and other guidance provided in Matrix 1 of RS-001.
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Technical Evaluation
2.1.2.1 USE Value Calculations
The NRC staff’s criteria for maintaining acceptable levels of USE for the RV beltline materials of
operating reactors throughout the licensed lives of the facilities is provided in 10 CFR Part 50,
Appendix G. The rule requires RV beltline materials to have a minimum USE value of 75 ft-lb in
the unirradiated condition, and to maintain a minimum USE value above 50 ft-lb throughout the
licensed period of operation of the facility, unless it can be demonstrated through analysis that
lower values of USE would provide acceptable margins of safety against fracture equivalent to
those required by Appendix G of Section XI of the American Society of Mechanical Engineers
Boiler and Pressure Vessel Code (ASME Code). The rule also mandates that the methods
used to calculate USE values must account for the effects of neutron irradiation on the USE
values for the materials and must incorporate any relevant RV surveillance capsule data that
are reported through implementation of a plant’s 10 CFR Part 50, Appendix H RV material
surveillance program.
The licensee discussed the impact of the approximate 7 percent SPU on the USE values for the
RV beltline materials in Section 2.1.2 of the Attachment 5 of the DNC’s SPU LR, dated July 13,
2007. In this section, the applicant stated that all RV beltline materials have a USE greater than
50 ft-lb through the EOL, 54 EFPY, as required by Appendix G to 10 CFR Part 50. Table 2.1.2-
4 of the SPU LR provides the predicted USE values for MPS3 beltline materials, based on the
neutron fluence value equivalent to 54 EFPY. In NUREG-1838, “Safety Evaluation (SE) Report
Related to the License Renewal of the Millstone Units 2 and 3,”8 the staff, reviewed and
approved the use of a neutron fluence value for 54 EFPY, which exceeds the neutron fluence
value under SPU conditions at 54 EFPY for the limiting beltline material as shown in Table
2.1.2-4 of the SPU LR. In request for additional information (RAI) CVIB-07-002, dated October
29, 2007,9 the staff requested that the licensee explain why the neutron fluence value at 54
EFPY under SPU conditions is lower than the staff-approved value as shown in the NUREG-
1838. In its response to RAI CVIB-07-002, dated November 19, 2007,10 the licensee stated that
the projected neutron fluence value approved by the staff in NUREG-1838 was based on the
analysis of MPS3 surveillance capsule X. The projected neutron fluence value under SPU
conditions at 54 EFPY was obtained from the analysis of the more recent MPS3 surveillance
capsule W, which projected a lower value under SPU conditions at 54 EFPY. The methodology
used to project the neutron fluence values for surveillance capsules X (WCAP-1540511) and W
(WCAP-1662912) adhered to the guidance in Regulatory Guide (RG) 1.190, “Calculational and
Dosimetry Methods for Determining Pressure Vessel Neutron Fluence,”13 or its precursor Draft
RG, DG-1053, “Calculational and Dosimetry Methods for Determining Pressure Vessel Nuetron
Fluence.”14
8 ADAMS Accession No. ML053270483
9 ADAMS Accession No. ML072960179
10 ADAMS Accession No. ML073230976
11 WCAP-15405NP, “Analysis of Capsule X from the Northeast Nuclear Energy Company Millstone Unit 3 Reactor
Vessel Radiation Surveillance Program” Westinghouse Electric Company, LLC by E. Terek at al, May, 2000.
12 WCAP-16629NP, “Analysis of Capsule W from the Dominion Nuclear Connecticut Millstone Unit 3 Reactor Vessel
Radiation Surveillance Program” Westinghouse Electric Company, LLC by F.C. Gift at al, September, 2006.
13 ADAMS Accession No. ML010890301
14 ADAMS Accession No. ML003777844
Description:242 TO RENEWED FACILITY OPERATING. LICENSE NO. NPF-49. DOMINION
NUCLEAR CONNECTICUT, INCORPORATED. MILLSTONE POWER STATION
