ML20085B635
| ML20085B635 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 07/30/1991 |
| From: | Capra R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20085B637 | List: |
| References | |
| NUDOCS 9108020324 | |
| Download: ML20085B635 (19) | |
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5c Ee NUCLEAR REGULATORY COMMISSION WAsHINotoN, D C,20666
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BALTIMORE,,G,AS AND ELECT,R1,C, Cpf,PA,NJ DOCKET NO 50-317 CALVERT CL1FFS NUCL E,AR,,P,0,WE,R, PLANT,U,N,1,T N,0,1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.158 License No. DPR-53 1.
The Nuclear Regulatory Commission (the Comission) has found that:
A.
The application for amendment by Baltimore Gas and Electric Company (thelicensee)datedMay 24, 1991, 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 in 10 CFR Chapter I; B.
The facility will o)erate in conformity with the application, the provisions of tie Act, and the rules and regulations of the Comission; C.-
Thereisreasonableassurance(i)that-the'activitiesauthorized 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 common 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 2.C.(2) of Facility Operating License No. OPR-53'is hereby amended to read as follows:
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PDR ADOCK 05000317 d)0 P-PDR
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. r (2) Technical Specifications The ' -hnical Specifications contained in Appendices A ans,, as revised through Amendment No.158, 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 COMMISSION bc
.f h Y
h Robert A. Capra, Director Project Directorate 1-1 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
Jrly 30,1991 F
&* M ou 3s y%
UNITED STATES h' "j /
E j
NUCLEAR REGULATORY COMMISSION i
WASHINGTON. D C. 20566 BALTitiORE GAS AND ELECTRIC COPPAtlY DOCKET NO. 50-318 C ALVERT CLJ,FJ5, !ipCL,E AR,,P,0WER, P,L Ati,T,,, y,NI,T, flp. 2
.AF,E!;,DF,Ep,T,,Tp, [p,CJ,LJ,T,Y OP E R AT I NG L I C E N S E Amendment flo.138 License No. OPR-69 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Baltimore Gas and Electric Company (the licensee) dated May 24, 1991, 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 in 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 Commission; 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 common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CER Part 51 of the Commission's regulations and all applicable requirements d
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 2.C.2 of Facility Operating Licensc No.
bPR-69 is hereby amended to read as follows:
4 2
(2) LechnicalSpecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.138, 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 COMMISSION 8
Robert A. Capra, Director Project Directorate 1-1 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: July 30,1991
ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT NO. 158 FACILITY OPERATING LICENSE NO. DPR-53 AMENDMENT NO.138 FACILITY OPERATING LICENSE NO. OPR-69 DOCKET NOS. 50-317 AND 50-318
-Revise Appendix A as follows:
Remove Pages Insert Pages 3/4 4-23 (DPR-53 only) 3/4 4-23 (DPR-53 only) 3/4 4-25 (DPR-53 only) 3/4 4-25 (DPR-53 only) 83/4 4-5 through 4-9 B3/4 4-5 through 4-9 3/4 4-24 (DPR-69 only) 3/4 4-24 (DPR-69 only) 3/4 4-26 (DPR-69 only) 3/4 4-26 (DPR-69 only)
3/4.4.9 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITlRG CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figure 3.4-2 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:
a.
A maximua heatup of:
Maximum Allowable Heatuo Rate RCS Temperature 0
40 F in any one hour period 70PF to 313 F 0
0 10 F in any one hour period 314 F to 327 F 0
0 60 F in any one hour period
> 327 F b.
A maximum cooldown of:
Maximum Allowable Cooldown Rate RCS Temoerature 0
100 F in any one hour period
> 250UF 0
0 0
20 7 in any one hour period 250 F to 170 F l
10 F in any one hour period
< 170 F 0
0 0
A maximum temperature change of 5 F in any one hour period, c.
during hydrostatic testing operations above system design pressure.
APPLICABILITY: At all times.
1 l
ACTION:
With any of the above limits exceeded, restore the temperature and/or i
pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the fracture toughness properties of the Reactor Coolant System; determine that the Reactor Coolant System remains acreptable for continued operations or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and l
reduce the RCS T and pressure to less than 200 F and 300 psia, 0
l respectively,wiNinthefollowing30 hours.
l l
SVRVEILLANCE RE0VIREMENTS l
4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during i
I system heatup, cooldown, and inservice leak and hydrostatic testing operations.
4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material l
properties, as required by 10 CFR Part 50, Appendix H.
The results of these examinations shall be used to update Figure 3.4-2.
CALVERT CLIFFS - UNIT 1 3/4 4-23 Amendment No. JfMJfE,158
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CALVERT CLIFFS - UNIT 1 3/4 4-25 Amendment No. JM,158 1
. =
-8 BASES steam generator tube rupture accident in conjunction with'an assumed steady state primary to-secondary steam generator leakage rate of 1.0 gpm and a concurrent loss of offsite electrical power.
The values for the limits on specific activity represent interim limits based upon a parametric evaluation by the NRC of typical site locaticns. These values are conservative in that specific site parameters of the Calvert Cliffs site, such as site boundary location and meteorological conditions, were not considered in this evaluation. The NRC is finalizing site specific criteria which will be used as the basis for the reevaluation of the specific activity limits of this site. This reevaluation may result in higher limits.
The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity >1.0 uC1/ gram DOSE EQUIVALENT-I-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER. Operation with specific i
activity levels exceeding 1.0 uCi/ gram DOSE EQUIVALENT I-131 but within the limits shown on figure 3.4-1 must be restricted to no more than 10 percent of the unit's yearly operating time since the activity levels allowed by Figure L 4-1 increase the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid dose at the site boundary by a factor of up to 20 following-a postulated steam _ generator l
tube rupture.
0 Reducing T to < 500 F prevents the release of activity should a steamgenerator,dberupturesincethesaturationpressureoftheprimary i
l coolant is below the lift prassure of the atmospheric steam relief I
valves.
The surveillance requirements provide adequate assurance that exces::ive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action.
Information obtained on iodine spiking will be used to assess the-parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses following pcwer changes may be permissible if justified by the data obtained.
3/4.4.9 PRESSURE / TEMPERATURE 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 operation. The various categories of load cycles used for design purposes are provided
- in Section 4.1.1 of the UFSAR. During startup and 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.
CALVERT CLIFFS - UNIT 1 8 3/4 4-5 Amendment No. JAB,158
BASES l
Operation within the appropriate heatup and cooldown curves assures l
l the integrity of the reactor vessel against fracture induced by combinative thermal and pressure stresses. As the vessel is subjected to
(
increasing fluerte, the toughness of the limiting material continues to decline, and ever more restrictive Pressure / Temperature limits must be i
observed. The current limits, Figures 3.4-2a and 3.4-2b, are for up to and including 12 Effective Full Power Years (EFPY) of operation.
l i
The reactor vessel materials have been tested to determine their initial RT the UFSAR.ND1; the results of these tests are shown in Section 4.1.5 of Reactor operation and resultant fast neutron (E>l Nev) irradiation will cause an increase in the RT nT.
The actual shift in u
RTNDT of the vessel material will be establisned periodically during operation by removing and evaluating reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor i
vessel in the core area. The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these i
l specimens are provided in UFSAR Tab'e 4-13 and are approved by the NRC prior to implementation in compliance with the requirements of Appendix H i
l to 10 CFR Part 50.
I 1
The shift in the material fracture toughness, as represented by l
RT nT, t the 1/4 T position, the adjusted reference temperature (A N
is calculated using Regulatory Guide 1.99, Revision 2.
For 12 i
EFPY, a value is 222 F.
At the 3/4 T position the ART value is 162.5 F.
These 0
values are used with procedures developed in the ASME Boiler and Pressure Vessel Code,Section III, Appendix G to calculate heatup and cooldown limits in accordance with the iequirements of 10 CFR Part 50, Appendix G.
To develop compositI pressure-temperature limits for the heatup transient, the isothermal,1/4 T heatup, and 3/4 T heatup pressure-temperature limits are compared for a given thermal rate. Then the most l
restrictive pressure-temperature limits are combined over the complete temperature interval resulting in a composite limit curve for the reactor vessel beltline for the heatup event.
To develop a composite pressure-temperature limit for the cooldown event, the isothermal pressure-temperature limit must be calculated.
The isothermal pressure-temperature limit is then compared to the pressure-temperature limit associated with a cooling rate and the more restrictive allowable pressure-temperature limit is chosen resulting in a composite limit curve for the reactor vessel beltline.
Both 10 CFR Part 50 Appendix G and ASME, Code Appendix G require the development of pressure-temperature limits which are applicable to inservice hydrostatic tests. The minimum temperature for the inservice hydrostatic test pressure can be detemined by entering the curve at the test pressure (1.1 times normal operating pressure) and locating the corresponding temperature. This curve is shown for 12 EFPY on Figures 3.4-2a and 3.4-2b.
CALVERT CLIFFS - UNIT 1 B 3/4 4-6 Amendment No. JM, 158
{
m REACTOR COOLANT SYSTEM ILASES Similarly, 10 CFR Part 50 specifies that core critical limits be established based on material considerations.
This limit is shown on the heatup curve, figure 3.4-2a.
Note that this limit does not consider the core reactivity safety analyses that actually control the temperature at which the core can be brought critical.
The Lowest Service Temperature is the minimum allowable temperature at pressures above 20% of the pre-operational system hydrostatic test pressure (625 psia).
This temperature is defined as equal to the most limiting RT for the balance of the Reactor Coolant System components plus 100 F,NDTper Article NB 2332 of Section III of the ASME Boiler and 0
Pressure Vessel Code.
The horizontal line between the minimum boltup temperature and the Lowest Service Temperature is defined by the ASME Boiler and Pressure Vessel Code as 20% of the pre operational hydrostatic test pressure.
The 0
change in the line at 150 F on the cooldown curve is due to a cessation cf RCP flow induced pressure deviation, since no RCPs are permitted to operate during a cooldown below 150 F.
The minimum boltur temperature is the minimum allowable temperature at pressures below 2J% of the pre-operational system hydrostatic te.,
pressure. The minimum is defined as the initial RT of the higher stressed region of the reactor vessel DT f or the naterial Nplus any effects for irradiation per Article G-2222 of Section III of the ASME Boiler and Pressure Vessel Code.
The initial reference temperature of the reactor vessel and closure head flanges was determined using the certified material test reports and Branch Technical Position MTEB 5-2.
The maximum initial RT nT associated with the stressed region of the closure y
head flange is -10 F.
The minimum boltup temperature including 0
0 0
temperature instrument uncertainty is -10 F + 10 F - 0 F.
However, 'ar 0
conservatism, a minimum boltup temperature of 70 F is utilized.
The design basis events in the low temperature region assuming a water solid system are:
A RCP start with hot steam generators; and, l
An inadvertent HPSI actuation with concurrent charging.
Any measures which will prevent or mitigate the design basis events are sufficient for any less severe incidents. Therefore, this section will discuss the results of the RCP start and mass addition transient analyses. Also discussed is the effectiveness of a pressurizer steam bubble and a single PORY relative to mitigating the design basis events.
The RCP start transient is a severe LTOP challenge for a water solid RCS.
Therefore, during water solid operations all-4 RCPs are tagged out of service. Analysis indicates the transient is adequately controlled by placing restrictions on three parameters:
initial pressurizer pressure and level, and the secondary-to-primary temperature difference. With these restrictions in place and when decay heat level is low (reactor has CALVERT CLIFFS - UNIT 1 B 3/4 4-7 Amendment No. JfA(JfE,158
I REACTOR COOLANT SYSTEM BASES been shutdown 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or longer), the transient is adequately controlled without the assistance of the PORVs. Operating procedures require that during normal cooldowns, entry into MPT enable (327 F and below) will not 0
occur until 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after reactor shutdown. This restriction is not intended to delay cooldown in situations where plant or personnel safety considerations make expeditious cooldown prudent.
If RCPs are restored in response to a loss of decay heat removal when decay heat loads are high and operator actions were either not taken or ineffective, a single PORV will protect the Appendix G limits.
The inadvertent actuation of one HPSI pump in conjunction with one charging pump is the most severe mass addition overpressurization event.
Analyses were performed for a single HPSI pump and one charging pump assuming one PORV available with the existing orifice area of 1.29 in,
2 For the limiting case, only a single PORV is considered available due to single failure criteria. A figure was developed which shows the calculated RCS pressures versus time that will occur assuming HPSI and charging pump mass inputs, and the expansion of the RCS following loss of l
decay heat removal. Sufficient overpressure protection results when the equilibrium pressure does not exceed the limiting Appendix G curve pressure. Because the equilibrium pressure exceeds the minimum l
Appendix G limit for full HPSI flow, HPSI flow is throttled to no more than 210 gpm indicated when the HPSI pump is used for mass addition.
The HPSI flow limit includes allowances for instrumentation uncertainty, charging pump flow addition and RCS expansion following loss of decay heat removal. The HPSI flow is injected through only one HPSI loop MOV to limit instrumentation uncertainty. No more than one charging pump (44 gpm) is allowed to operate during the HPSI mass addition.
Comparison of the PORV discharge curve with the critical pressurizer pressure of 46a.1 psia indicates that adequate protection is provided by a sirigle PORV for RCS temperatures of 70 F or above when all mass input 0
i is limited to 380 gpm. HPSI discharge is limited to 210 gpm to allow for one charging pump and system expansion due to loss of decay heat removal.
To provide single failure protection against a HPSI pump mass addition transient, the HPSI loop MOV handswitches must be placed in pull-to-override so the valves do not automatically actuate upon receipt of a SIAS signal. Alternative actions, described in the ACTION STATEMENT, are to disable the affected MOV (by racking out its motor i
circuit breaker or equivalent), or to isolate the affected HPSI header.
Examples of HPSI header isolation actions include; (1) de-energizing and tagging shut the HPSI header isolation valves; (2) locking shut and tagging all three HPSI pcmp discharge MOVs; and (3) disabling all three HPSI pumps.
CALVERT CLIFFS - UNIT 1 B 3/4 4-8 Amendment No. J/ MJfE, De
~
/
i REACTOR COOLANT SYSTEM BASES l
Three 100% capacity HPSI pumps are installed at Calvert Cliffs.
l Procedures will require that two of the three HPSI pumps be disabled 0
(breakers racked out) at RCS temperatures less than or equal to 327 F and that the remaining HPSI pump handswitch be placed in pull-to-lock.
Additionally, the HPSI pump normally in pull-to-lock shall be throttled to less than or equal to 210 gpm when used to add mass to the RCS.
I Exceptions are provided for ECCS testing and for response to LOCAs.
A pressurizer steam volume and a single PORV will provide satisfactory control of all mass addition transients with the exception of a spurious actuation of full flow from a HPSI pump. Overpressurization due to this transient will be precluded for temperatures 327 F and less by disabling two HPSI pumps, placing the third in pull-to-lock, and by throttling the third pump to less than or equal to 210 gpm ficw when it is used to add mass to the RCS.
Note that only the design bases events are discussed in detail since the less severe transients are bounded by the RCP start and inadvertent HPSI actuation analysis.
RCS temperature, as used in the applicability statement, is determined as follows: (1) with the RCPs running, the RCS cold leg temperature is the appropriate indication, (2) with the shutdown cooling system in operation, the shutdown cooling temperature indication is appropriate, (3) if neither the RCPs or shutdown cooling is in operation, the core exit thermocouples are the appropriate indicators of RCS temperature.
CALVERT CLIFFS - UNIT 1 8 3/4 4-9 Amendment No. ME,158
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En REACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS 4.4.9.1.1 -The Reactor Coolant System temperature and pressure shall be determined-to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations.
4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined,- to determine changes in material properties, as required by 10 CFR Part 50, Appendix H.
The results of these examinations shall be used to update Figure 3.4-2.
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l l-CALVERT CLIFFS - UNIT 2 3/4 4-24 Amendment No.138
g.
DELETED
-4 i
CALVERT CLIFFS - UNIT 2 3/4 4-26 Amendment No.138
1 REACTOR COOLANT SYSTEM BASES steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 gpm and concurrent loss of offsite electrical power.
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 of the Calvert Cliffs site, such as site boundary location and meteorological conditions, were not considered in this evaluation. The NRC is finalizing site specific criteria which will be used as the basis for the reevaluation of the specific activity limits of this site.
This reevaluation may result in higher limits.
The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity
>l.0 uCi/ gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER. Operation with specific activity levels exceeding 1.0 uti/ gram DOSE EQUIVALENT I-131 but within the limits shown on Figure 3.4-1 must be restricted to no more than 10 percent of the unit's yearly operating time since the activity levels dllowed by Figure 3.4-1 increase the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid dose at the site boundary by a factor of up to 20 following a postulated steam generator tube rupture.
0 Reducing T to < 500 F prevents the release of activity should a steam generator N be rupture since the saturation pressure of the primary a
coolant is below the lift pressure of the atmospheric steam relief valves.
The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action.
Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.
3/4.4.9 PRESSURE / TEMPERATURE 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 operation. The various categories of load cycles used for design purposes are provided in Section 4.1.1 of the UFSAR. During startup and 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.
CALVERT CLIFFS - UNIT 2 8 3/4 4-5 Amendment No. J3J,138
4 REACTOR COOLANT SYSTEM BASES Operation within the appropriate heatup and cooldown curves assures the integrity of the reactor vessel against fracture ir>Nced by combinative thermal and pressure stresses. As the vessel is subjected to increasing fluence, the toughness of the limiting material continues to decline, and even morc restrictive Pressure / Temperature limits must be observed.
The current limits Figures 3.4 2b and 3.4 2c, are for up to and including 12 Effective Full Tower Years (EFPY) of operation.
The reactor vessel materials have been tested to determine their initial RT the UFSAR NDT; the results of these tests are shown in Section 4.1.5 of Reactor operation and resultant fast neutron (E>l Mev) irradiation will cause an increase in the RT nT.
The actual shift in N
RTNOT of the vessel material will be establisned periodically during operation by removing and evaluating reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area. The number of reactor essel irradiation surveillance specimens and the frequencies for r amoving and testing these specimens are provided in UFSAR Table 4-13 and are approved by the NRC prior to implementation in compliance with the requirements of Appendix H to 10 CFR Part 50.
The shift in the material fracture toughness, as represented by RT s
is calculated using Regulatory Guide 1.99, Revision 2.
For 12 ND, t the 1/4 T position, the adjusted reference temperature (ART)
EFPY a
0 value is 171 F, At the 3/4 T position the ART value is 125 F.
These 0
values are used with procedures developed in the ASME Boiler and Pressure
('
Vessel Code, Section Ill, Appendix G to calculate heatup and cooldown limits in accordance with the requirements of 10 CFR Part 50, Appendix G.
To develop composite pressure-temperature limits for the heatup transient, the isothermal,1/4 T heatup, and 3/4 T heatup pressure-temperature limits are compared for a given thermal rate. Then the most restrictive pressure-temperature limits are combined over the complete temperature interval resulting in a composite limit curve for the reactor vessel beltline for the heatup event.
To develop a composite pressure-temperature limit for the cooldown event, the isothermal pressure-te:nperature limit must be calculated. The isothermal pressure-temperature limit is then compared to the pressure-temperature limit associated with a cooling rate and the more restrictive allowable pressure-temperature limit is chosen resulting in a composite limit curve for the' reactor vessel beltline.
'(
Both 10 CFR Part 50 Appendix G and ASME, Code Appendix G require the development of pressure-temperature limits which are applicable to ins rvice hydrostatic tests. The minimum temperature for the inservice hydrostatic test pressure can be determined by entering the curve at the test pressure 'l.1 times normal operating pressure) and locating the corresponding temperature. This curve is shown for 12 EFPY on Figures 3.4-2b and 3.4-2c.
CALVERT CLIFFS - UNIT 2 B 3/4 4-6 Amendment No. JJJ,138
m REACTOR COOLANT SYSTEM BASES Similarly,10 CFR Part 50 specifies that core critical limits be established based on material considerations.
This limit is shown on the heatup curve, Figure 3.4-2b.
Note that this limit does not consider the core reactivity safety analyses that actually control the temperature at which the core con be brought critical.
The Lowest Service Temperature is the minimum allowable temperature at pressures above 20% of the pre-operational system hydrostatic test pressure (625 psia).
This temperature is defined as equal to the most limiting RT for the balance of the Reactor Coolant System components plus 100 F,NDTper Article NB 2332 of Section III of the ASME Boiler and 0
Pressure Vessel Code.
The horizontal line between the minimum boltup temperature and the Lowest Service Temperature is defined by the ASME Boiler and Pressure Vessel Code as 20% of the pre-operational hydrostatic test pressure.
The 0
change in the line at 150 F on the cooldown curve is due to a cessation of RCP flow induced pressure deviation, since no RCPs are permitted to 0
operate during a cooldown below 150 F.
The minimum boltup temperature is the minimum allowable temperature at pressures below 20% of the pre operational system hydrostatic test pressure.
The minimum is defined as the initial RTNDT f r the material of the higher stressed region of the reactor vessel plus any effects for irradiation per Article G 2222 of Section III of the ASME Boiler and Pressure Vessel Code. The initial reference temperature of the reactor vessel and closure head flanges was determined using the certified material test reports and Branch Technical Position MIEB 5-2, The maximum initial RT head flange is 30 F.T associated with the stressed region of the closure ND 0
The minimum boltup temperature including 0
0 temperature instrument uncertainty is 30 F + 10 F = 40 F.
However, for 0
conservatism, a minimum boltup temperature of 70 F is utilized in the analysis to establish the. low temperature PORV lift setpoint.
The design basis events in the low temperature region assuming a water solid system are:
A RCP start with hot steam generators; and, An inadvertent HPSI actuation with concurrent charging.
Any measures which will prevent or mitigate the design basis events are sufficient for any less severe incidents.
Therefore, this section will discuss the results of the RCP start and mass addition transient analyses.
Also discussed is the effectiveness of a pressurizer steam bubble and a single PORV relative to mitigating the design basis events.
CALVERT CLIFFS - UNIT 2 8 3/4 4-7 Amendment No. JJJ, 138
REACTOR COOLANT SYSTEM fLASES The RCP start transient is a severe LTOP challenge for a water solid RCS.
Therefore, during water solid operations all four RCPs are tagged out of service. Analysis indicates the transient is adequately controlled by placing restrictions on three parameters:
initial pressurizer pressure and level, and the secondary-to-primary temperature difference.
With these restrictions in place, the transient is adequately controlled without the assistance of the PORVs.
The inadvertent actuation of one HPSI pump in conjunction with one charging pump is the most severe mass addition overpressurization event.
Analyses were performed for a single HPSI pump and one charging pump assuming one PORV available with the existing orifice area of 1.29 in,
2 For the limiting case, only a single PORY is considered available due to single failure criteria. A figure was developed which shows the calculated RCS pressures versus time that will occur assuming HPSI and charging pump mass inputs, and the expansion of the RCS following loss of decay heat removal. Sufficient overpressure protection results when the equilibrium pressure does not exceed the limiting Appendix G curre pressure. Because the equilibrium pressure exceeds the minimum Appendix G limit f r full HPSI flow, HPSI flow is throttled to no more than 210 gpm indicated when the HPSI pump is used for mass addition. The HPSI flow limit includes allowances fo? instrumentation uncertainty, charging pump flow addition and RCS expansion following loss of decay heat removal. The HPSI flow is injected through only one HPSI loop H0V to limit instrumentation uncertainty, No more than one charging pump (44 gpm) is allowed to operate during the HPSI mass addition.
Comparison of the PORV discharge curve with the critical pressurizer i
pressure of 471.2 psia indicates that adequate protection is provided by a single PORV for RCS temperatures of 70 F or above when all mass input 0
is limited to 380 gpm. HPSI discharge is limited to 210 gpm to allow for one charging pump and system expansion due to loss of decay heat removal.
The low temperature PORY pressure lift setpoint is set to protect the most restrictive Appendix G pressure limit (471.2 psia). A PORV setpoint of 430 psia, which includes instrumentation uncertainties and sufficient margins for PORV response time requirements necessary for the protection of 471.2 psia, was selected.
To provide single failure protection against a HPSI pump mass addition transient, the HPSI loop M0V handswitches must be placed in pull-to-override so the valves do not automatically actuate upon receipt of a SIAS signal. Alternative actions, described in the ACTION statement, are to disable the affected M0V (by racking out its motor circuit breaker or equivalent), or to isolate the affected HPSI header.
Examples of HPSI header icolation actions include; (1) de-energizing and tagging shut the HPSI header isolation valves; (2) locking shut and tagging all three HPSI pump discharge MOVs; and (3) disabling all three HPSI pumps.
CALVERT CLIFFS - UNIT 2 8 3/4 4-8 Amendment No. JJJ,138
L REACTOR COOLANT' SYSTEM BASES Three 100% capacity HPSI pumps are installed at Calvert Cliffs.
Procedures will require that two of the three HPSI pumps be disabled-(breakers racked out) at RCS temperatures less than or equal to 305 F and that the remaining HPSI pump handswitch be placed in pull-to-lock.
Additionally, the HPSI pump normally in pull-to-lock shall be throttled to less than or equal to 210 gpm when used to add mass to the RCS.
Exceptions are provided for ECCS testing and for response to LOCAs.
A pressurizer steam volume and a single PORV will provide satisfactory control of-all mass addition-transients with the exception of a spurious actuation of full flow from a HPSI pump. Overpressurization due to this transient will be precluded for temperatures 305 F and less by disabling 0
two HPSI pumps, placing the third in pull-to-lock, and by throttling the third pump to less than or equal to 210 gpm flow when it is used to add mass to the RCS.
Note that only the design bases events are discussed in detail since the less severe transients are bounded by the RCP start and inadvertent HPSI actuation analysis.
RCS temperature, as used in the applicability statement, is determined as follows: (1) with the RCPs running, the RCS cold leg temperature is the appropriate indication, (2) with the shutdown cooling system in operatiot, the shutdown cooling temperature indication is appropriate, (3) if neither the RCPs or shutdown cooling is in operation, the core exit thermocouples are the appropriate indicators of RCS tempe,ature.
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I CALVERT CLIFFS - UNIT 2
-B 3/4 4-9 Amendment No. JJJ,138 i
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