ML20135E430
| ML20135E430 | |
| Person / Time | |
|---|---|
| Site: | Dresden, Quad Cities  |
| Issue date: | 02/28/1997 |
| From: | Pulsifer R NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20135E152 | List: |
| References | |
| NUDOCS 9703070069 | |
| Download: ML20135E430 (16) | |
Text
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UNITED STATES p
s NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. SteeHoot 1
COM40NWEALTH EDISON COMPANY m
MIDAMERICAN ENERGY COMPANY j
DOCKET NO. 50-254 00AD CITIES NUCLEAR POWER STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE i
Amendment No.172 License No. DPR-29 1.
The Nuclear Regulatory Comission (the Commission) has found that:
I A.
The application for amendment by Commonwealth Edison Company (the licensee) dated September 20, 1996, as supplemented January 21, 1997, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth iri 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Coisnission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B. of Facility Operating License No. DPR-29 is hereby amended to read as follows:
9703070069 970220 PDR ADOCK 05000254 p
l B.
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.172, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMISSION For bert
. Pulsifer, Project Manager Project Directorate III-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: February 28, 1997 f
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UNITED STATES p
i s
j NUCLEAR REGULATORY COMMISSION
- y.....j WASHINGTON, D.C. 30006 40M l
1 COPMONWEALTH EDISON COMPANY AND MIDAMERICAN ENERGY COMPANY DOCKET NO. 50-265 0UAD CITIES NUCLEAR POWER STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 168 License No. OPR-30 1.
The Nuclear Regulatory Comission (the Commission) has found that:
A.
The application for amendment by Commonwealth Edison Company (the licensee) dated September 20, 1996, as supplemented January 21, 1997, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth 1n 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in co,mpliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B. of Facility Operating License No. DPR-30 is hereby amended to read as follows:
1 I
i l
t
l B.
Technical Soecificatim The Technical Specifications contained in Appendices A and B, as revised through Amendment No.168, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COPMISSION h F.-
bert M. Pulsifer, Project Manager Project Directorate III-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: February 28, 1997
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ATTACHMENT TO LICENSE AMENDMENT NOS.172 AND 168 FACILITY OPERATING LICENSE NOS. DPR-29 AND DPR-30 DOCKET NOS. 50-254 AND 50-265 Revise the Appendix A Technical Specifications hy removing the pages identified below and inserting the attached pages. The revised pages are identified hy the captioned amendment number and contain marginal lines indicating the area of change.
REMOVE INSERT VIII VIII 3/4.6-19 3/4.6-19 3/4.6-20 3/4.6-20 3/4.6-21 3/4.6-21 3/4.6-21a 3/4.6-21b 3/4.6-21c, 3/4.6-21d*
B3/4.6-5 B3/4.6-5 B3/4.6-6 B3/4.6-6 B3/4.6-7 83/4.6-7 B3/4.6-8 f
B3/4.6-8 B3/4.6-9 t
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TABLE OF CONTENTS TOC i
1 LIMITING CONDITIONS FOR OPERATION ANO SURVEILLANCE REQUIREMENTS SECTION f1fai 31.4 &
PRIMARY SYSTEM BOUNDARY l
i 3/4.6.A Recirculation loops........................................ 3/4.6-1 l
3/4.6-3 3/4.6.5 Jet Pumps 3/4.6.C Recirculation Pumps........................................ 3/4.6-5 3/4.6-6 3/4.6.D 6dle Recirculation Loop Startup 3/4.6.E Safety Valves............................................ 3/4.6-7
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3/4.6.F Relief Valves............................................. 3/4.6-8 3/4.6.G Leakage Detection Systems.................................. 3/4.6-10 3/4.6.H O perational Lea kage....................................... 3/4.6-11 3/4.6 13 3/4.6.1 Chemistry Table 3.6.11, Reactor Coolant System Chemistry Limits 3/4.6.J Specific Activity............................. '............. 3/4.6 16 3/4.6.K Pressure / Temperature Limits................................. 3/4.6-19 Figure 3.6.K-1, Pressure-Temperature Limits for Pressure Testing - Valid to 18 EFPY Figure 3.6.K-2, Pressure-Temperature Limits for Pressure Testing - Valid to 20 EFPY Figure 3.6.K-3, Pressure Temperature Limits for Pressure Testing - Valid to 22 EFPY Figure 3.6.K-4, Pressure-Temperature Limits for Non-Nuclear Hestup/Cooldown - Valid to 22 EFPY Figure 3.6.K-5, Pressure-Temperature Limits for Critical Core Operations - Valid to 22 EFPY 3 %.6.L Reactor Steam Dome Pressure................................ 3/4.6 22
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3/4.6.M Main Steam Line Isolation Valves.............................. 3/4.6 23 3/4.6.N Structural integrity........................................ 3/4.6 24 3/4.6.0 Residual Heat Removal - HOT SHUTDOWN....................... 3/4.6 25 3/4.6.P Residual Heat Removal - COLD SHUTDOWN...................... 3/4.6 27
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i OUAD CITIES - UNITS 1 & 2 Vill Amendment Nos.172 and 168
PRIMARY SYSTEM BOUNDARY PT Limits 3/4.6.K 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS K.
Pressure / Temperature Limits K.
Pressure / Temperature Limits The p'rimary system coolant system
- 1. During non-nuclear heatup or cooldown, temperature and reactor vessel metal and pressure testing operations, at least temperature and pressure shall be limited as once per 30 minutes, specified below:
- a. The rate of change of the primary
- 1. Pressure Testing:
system coolant temperature shall be determined to be within the heatup
- a. The reactor vessel metal temperature and cooldown rate limits, and and pressure shall be maintained within the Acceptable Regions as
- b. The reactor vessel metal temperature shown on Figures 3.6.K-1 through and pressure shall be determined to 3.6.K-3 with the rate of change of the be within the Acceptable Regions on primary system coolant temperature Figures 3.6.K-1 through 3.6.K-4.
s20'F per hour, or
- 2. For reactor critical operation, determine
- b. The rate of change of the primary within 15 minutes prior to the withdrawal system coolant temperature shall be of control rods and at least once per 30 s100*F per hour when reactor vess'el minutes during primary system heatup or metal temperature and pressure is
- cooldown, maintained within the Acceptable Regions as shown on Figure 3.6.K-4.
- a. The rate of change of the primary system coolant temperature to be
- 2. Non-Nuclear Heatup and Cooldown and within the limits, and low power PHYSICS TESTS:
- b. The reactor vessel metal temperature
- a. The reactor vessel metal temperature and pressure to be within the and pressure shall be maintained Acceptable Region on Figure 3.6 K-5.
within the Acceptable Regions as shown on Figure 3.6.K-4, and
- 3. The reactor vessel material surveillance specimens shall be removed and
- b. The rate of change of the primary examined, to det*~nine changes in system coolant temperature shall be reactor pressure vsal material s 100'F per hour, properties in accordonca with 10CFR Part 50, Appendix H.
QUAD CITIES - UNITS 1 & 2 3/4.6-19 Amendment Nos.172 and 168
PRIMARY SYSTr.M BOUNDARY PT Limits 3/4.6.K 3.6 - LIMITINQ COND TIONS FOR OPERATION 4.6 - SURVElLLANCE REQUIREMENTS
- ;. g_,
- 3. Nuclear Hestup snd Cooldown:
- a. The reactor vessel metal temperature and pressure shall be nmintained s.
In OPERATIONAL MODE 4 when the l
within the Acceptable Region as reactcir coolant temperature is:
shown on Figure 3.6.K-5, and
- 1) s113*F, at least once per
- b. The rate of change of the primary 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
system coolant temperature shall be
$100'F per hour.
- 2) 593*F, at least once per 30 minutes.
Within 30 minutes prior to and at least head botting studs are under tension.
once per 30 minutes during tensioning of the reactor vessel head botting studs.
APPLICABILITY:
At all times.
ACTION:
l With any of the above limits exceeded,
- 1. Restore the reactor vessel metal temperature and/or pressure to within the limits within 30 minutes riithout exceeding the applicable primary system coolant temperature rate of change limit, and
- 2. Perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural ir.tegrity of the reactor coolant systen' and determine that the reactor coolant system remains acceptable for continued operations l
within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or
- 3. Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
QUAD CITIES - UNITS 1 & 2 3/4.6-20 Amendment Nos. 172 and 168
i PRIMARY SYSTEM BOUNDARY PT Limits 3/4.6 K -
l FIGURE 3.6.K-1 1
l PRESSURE - TEMPERATURE LIMITS FOR PRESSURE TESTING - VAllD TO 18 EFPY 1
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UPPER VESSEL AND I 312 Peo 1 M WITS l.
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OUAD CITIES - UNITS 1 & 2 3/4.6-21 Amendment Nos.172 and 16E i
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PRIMARY SYSTEM BOUNDARY PT Limits 3/4.6.K l
FIGURE 3.6.K-2 l
PRESSURE - TEMPERATURE LIMITS FOR PRESSURE TESTING - VAllD TO 20 EFPY i
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QUAD CITIES - UNITS 1 & 2 3/4.6-21 a Amendment Nos.172 and II 8
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PRIMARY SYSTEM BOUNDARY PT Limits 3/4.6.K l
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FIGURE 3.6.K-3 1,
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PRESSURE - TEMPERATURE LIMITS FOR PRESSURE TESTING - VALID TO 22 EFPY
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i emetMUM REACTOR VESSEL IETALTEWWIATINtE (T) l l
QUAD CITIES - UNITS 1 & 2 3/4.6-21 b Amendment Nos.172 and 16E i
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PRIMARY SYSTEM BOUNDARY PT Limits 3/4.6.K I
FIGURE 3.6.K-4 i
l PRESSURE - TEMPERATURE LIMITS FOR NON-NUCLEAR HEATUP/COOLDOWN - VALID TO 22 EFPY l
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QUAD CITIES - UNITS 1 & 2 3/4.6-21c Amendment Nos.172 and 161 l
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PRIMARY SYSTEM BOUNDARY PT Limits 3/4.6.K FIGURE 3.6.K-5 PRESSURE - TEMPERATURE LIMITS FOR CRITICAL l-CORE OPERATIONS - VALID TO 22 EFPY 1400
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1200 I,.00
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l UNACCEPTABLE REWON
- Tempestum Change Rate 800 s10FF/hr 800 Ei t
'400
-BELTUNE AND NON-MLTUNE LIMITS 200 terumum canonuty Tempemane arf l ACCEPTABLE REGION
- l 0
O 50 100 150 200 250 300 350 ANNIMUM REACTOR VESSEL MALMTINtE (T)
QUAD CITIES - UNITS 1 & 2 3/4.6-21d Amendment Nos.172 and 168 I
- ~ -
PRIMARY SYSTEM BOUNDARY B 3/4.6 BASES 3/4.6.J Soecific Activity The limitations on the specific activity of the primary coolant ermure that the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid and whole body doses resulting from a main steam line failure outsioe the containment during steady state operation will not exceed small fractions of the dose guidelines of 10 CFR 100. The values for the limits on specific activity represent interim limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters, such cs site boundary location and meteoroiogicai ?~fitions, were not considered in this evaluation.
The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 0.2 microcuries per gram DOSE EQUIVALENT l-131, but less than or equal to 4.0 microcuries per gram DOSE EQUlVALENT l-131, cccommodates possible iodine spiking phenomenon whic:4 may occur following changes in THERMAL POWER. Information obtained on iodine spiking will be used to assess the parameters tssociated with spiking phenomena. A reduction in frequency of isotopic analysis following power changes may be perroissible if justified by the data obtained.
Closing the main steam line isolation valves prevents the release of activity to the environs should a steam line rupture occur outside containment. The surveillance requirements provide adequate cssurance that excessiva specific activity levels in the reactor coolant will be detected in sufficient nne to take corrective action.
3/4.6.K Pressure /Temocrature Limits All components.in the reactor coolant system are designed to withstand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Section 3.9.1.1.1 of the UFSAR. During startup cnd shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.
The pressure-temperature limit lines are shown, for operating conditions; Pressure Testing, Figures l
3.6.K-1 through 3.6.K-3 Non-Nuclear Heatup/Cooldown, Figure 3.6.K-4 and Core Critical Operation I
Figure 3.6.K 5. The curves have been established to be in conformance with Appendix G to 10 CFR Part 50 and Regulatory Guide 1.99 Revision 2, and take into account the change in reference nil-ductility transition temperature (RT ) as a result of neutron embrittlement. The adjusted r::ference temperature (ART) of the limiting vessel materialis used to account for irradiation effects.
Four vessel regions are considered for the development of the pressure-temperature curves: 1) the core b'eltline region; 2) the non-bettline region (other than the closure flange region and the bottom head region); 3) the closure flange region, and 4) the bottom head region. The beltline QUAD CITIES - UNITS 1 & 2 B 3/4.6-5 Amendment Nos.172 and 168
PRIMARY SYSTEM BOUNDARY B 3/4.6 l
j BASES 4
i i
region is defined as that region of the reactor vessel that directly surrounds the effective height of the reactor core and is subject to en RT, adjustment to account for radiation embrittlement. The non-beltline, closure flange, and bottom head regions receive insufficient fluence to necessitate an I
RT, adjustment. These regions contain components which include: the reactor vessel nozzles, closure flanges, top and bottom head plates, control rod drive penetrations, and shell plates that do i
not directly surround the reactor core. Although the closure flange and bottom head regions are non-beltline regions, they are treated asperately for the development of the pressure-temperature i
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curves to address 10CFR Part 50 Appendix G requirements.
Boltuo Temnerature The limiting initial RT of the main closure flanges, the shell and head materials connecting to these flanges, connecting welds and the vertical electroslag welds which terminate 1
immediately below the vessel flange is 237. Therefore, the minimum allowable boltup temperature is established as 83T (RT
+ 607) which includes a 60T conservatism required j
by the original ASME Code of construction.
i Fiaures 3.6.K-1 throuah 3.6.K-3 Praname Testina 3
As indicated in Figure 3.6.K-1 through 3.6.K 3 for pressure testing, the minimum metal temperature of the reactor vessel shellis 837 for reactor pressures less then 312 psig. This i
83Y minimum boltup temperature is based on a RTa of 237 for tM electroslag weld i
immediately below the vessel flange and a 607 conservatism requir'ed by the original ASME i
Code of construction. The bottom head region limit is established as 687, based on i
moderator temperature assumptions for shutdown margin analyses. At reactor pressures greater than 312 psig, the minimum vessel metal temperature is established as 1137. The 113T minimum temperature is based on a closure flange region RT of 23Y and a g0T i
conservatism required by 10CFR Part 50 Appendix G. Beltline curves as a function of vessel exposure for 18,20 and 22 effective full power years (EFPY) are presented to allow the use of i
the appropriate curve up to 22 0FPY of operation.
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Fioures 3.6.K-1 through 3.6.K 3 are goveming for applicable pressure testing with a maximum i
l E *stup/coc%wn rate of 207/ hour.
i Fiaure 3.6.K Non Nuclear Heatun/Cooldown I.
J j
Figure 3.6.K-4 applies during heatups with non-nuclear heat (s'.g., recirculation pump heat) and during cooldowns when the reactor is not critical (e.g., following a scram). The curve provides 1
2 the minimum reactor vessel metal temperatures based on the most limiting vessel stress. The maximum heatup/cooldown rate of 100T/ hour is applicable.
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f QUAD CITIES - UNITS 1 & 2 i
B 3/4.6 6 Amendment Nos.172 and 168 3
4 mm W **
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PRIMARY SYSTEM BOUNDARY B 3/4.6 i
BASES l
Fioure 3.6.K Core Critical Operation The core critical operation curve shown in Figure 3.6.K-5, is generated in accordance with 10CFR i
Part 50 Appendix G which requires core critical pressure-temperature limits to be 40T above any i
Pressure testing or non-nuclear heatup/cooldown limits. Since Figure 3.6.K-4 is more limiting, i
Figure 3.6.K-5 is Figure _3.6.K-4 plus 40T. The maximum heatup/cooldown rate of 100T/ hour is j
applicable.
i The actual shift in RT,an of the vessel material will be established periodically during operation by l
removing and evaluating, in accordance with ASTM E185-82 and 10CFR Part 50, Appendix H, l
j irradiated reactor vessel material specimens installed near the inside wall of the reactor vessel in the l
core area. The irradiated specimens are used in predicting reactor vessel material embrittlement. The operating limit curves of Figures 3.6.K-1 through 3.6.K-5 shall be adjusted, as required, on the basis d
of the specimen data and recommendations of Regulatory Guide 1.99, Revision 2.
3/4.6.L Reactor Steam Dome Pressure The reactor steam dome pressure is an assumed initial condition of Design Basis Accidents and transients and is also an assumed value in the determination of compliance with reactor pressure vessel overpressure protection criteria. The reactor steam dome pressure of s 1005 psig is an initial condition of the vessel overpressure protection analysis. This analysis assumes an initial maximum reactor steam dome pressure and evaluates the response of the presisure relief system, primarily the safety valves, during the limiting pressurization transient. The deterrnination of compliance with the overpressure criteria is dependent on the initial reactor steam dome pressure; therefore, the limit on this pressure ensures that the assumptions of the overpressure protection analysis are conserved.
3/4.6.M Main Steam Line isolation Valves Double isolation valves are provided on each of the mairi steam lines to minimize the potential leakage
' paths from the containment in case of a line break. Only one valve in each line is required to maintain the integrity of the containment, however, single failure considerations require that two valves be OPERABLE. The surveillance requirements are based on the operating history of this type of valve.
The maximum closure time has been selected to contain fission products and to ensure the core is not uncovered following line breaks. The minimum closure time is consistent with the assumptions in the safety analyses to prevent pressure surges.
3/4,6,N Structural Inteority The inspection programs for ASME Code Class 1,2 and 3 components ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant.
QUAD CITIES - UNITS 1 & 2 B 3/4.6 7 Amendment Nos.172 and 161 l
l PRIMARY SYSTEM BOUNDARY B 3/4.6 l'
BASES The inservice inspection program for ASME Code Class 1,2 and 3 comgionents will be performed in 2
i accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the NRC pursuant to 10 CFR Part 50.55a(g)(6)(i).
3/4.6.0 Residual Heat Removal - HOT SHUTDOWN l
1 3/4.6.P Residual Heat Removal - COLD SHUTDOWN i
.I erradiated fuel in the reactor pressure vessel generates decay heat during normal and abnormal shutdown conditions, potentially resulting in an increase in the temperature of the reactor coolant.
This decay heat is required to be removed such that the reactor coolant temperature can be reduced in i
preparation for performing refueling, maintenance operations or for maintaining the reactor in cold shutdown conditions. Systems capable of removing decay heat 3re therefore required to perform these functions.
A single shutdown cooling mode subsystem provides sufficient heat removal capability for removing core decay heat and mixing to assure accurate temperature indication, however, single failure considerations require that two subsystems be OPERABLE or that alternab. methods capable of decay heat removal be demonstrated and that an alternate method of coola'nt mixing be in operation. An OPERABLE RHR shutdown cooling subsystem consists of one OPERABLE RHR pump, one heat exchanger, and the associtated piping and valves. The two subsystems have a common suction j
source and are allowed to have a common heat exchanger and common discharge piping. Therefore, to meet the Limiting Condition for Operation, both pumps in one loop or one pump in each of the two j
loops must be OPERABLE. Since the piping and heat exchangers are passive components that are assumed not to fail, they are allowed to be common to both subsystems (the ability to take credit for a common heat exchanger and discharge piping only applies to the SDC mode of RHR).
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l QUAD CITIES - UNITS 1 & 2 B 3/4.6-8 Amendment Nos. 172 and 168