ML20137R805
| ML20137R805 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 04/07/1997 |
| From: | Milano P NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20137R810 | List: |
| References | |
| NUDOCS 9704140254 | |
| Download: ML20137R805 (12) | |
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t UNITED STATES j
j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. - "1 s.+..../
BOSTON EDISON COMPANY i
DOCKET NO. 50-293 PILGRIM NUCLEAR POWER STATION l
AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 171 I
License No. DPR-35 1.
The Nuclear Regulatory Comission (the Comission or the NRC) has found 1
that:
.A.
The application for amendment filed by the Boston Edison Company (the licensee) dated January 24, 1997, as supplemented March 27, 1997, complies with the standards and requirements of the Atcaic Energy Act i
of 1954, as amended (the Act), and the Comission's rules and regulations; 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 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 Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment, and paragraph 2.C(2) of Facility Operating License No. DPR-35 is hereby amended to read as follovs:
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9704140254 970407 PDR ADOCK 05000293 P
PDR w
I 3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION 3% _
c Patrick D. Milano, Acting Director Project Directorate I-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation 4
Attachment:
Changes to the Technical
' Specifications Date of Issuance: April 7, 1997 l
P i
A ATTACHMENT TO LICENSE AMENDMENT NO. 171 EACILITY OPERATING LICENSE NO. DPR-35 DOCKET NO. 50-293 Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
Remove Insert 2.1 2.1 B2-1 B2-1 B2-2 82-2 B2-3 B2-3 B2-4 B2-4 3/4.2-22 3/4.2-22 l
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l 2.0 3AFETY LIMITS 2.1 Safety Limits 2.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10% of rated core flow:
THERMAL POWER shall be < 25% of RATED THERMAL POWER.
2.1.2 With the reactor steam dome prem> ire t 785 psig and core flow 110% of rated core flow:
MINIMUM CRITICAL POWER fu a shall be 21.08.
2.1.3 Whenever the reactor is.in the cold shutdown condition with irradiated fuel in the reactor vessel, the water level shall not be less than 12 inches above tne top of the nomial active fuel zone.
2.1.4 Reactor steam dome pressure shall be 51325 psig at any time when irradiated fuel is present in the reactor vessel.
2.2 Safety Limit Violation With any Safety Limit not met the following actions shall be met:
2.2 1 Within one hour notify the NRC Operations Center in accordance with 10CFR50.72.
2.2.2 Within two hours:
A.
Restore compliance with all Safety Lim' its, and B.
Insert allinsertable control rods.
2.2.3 The Station Director and Senior Vice President - Nuclear and the Nuclear i
Safety Review and Audit Committee (NSRAC) shall be notified within 24 bours.
2.2.4 A Licensee Event Report shall be prepared pursuant to 10CFR50.73. The Licensee Event Report shall be submitted to the Commission, the Operations Review Committee (ORC), the NSRAC and the Station Director and Senior Vice President - Nuclear within 30 days of the violation.
2.2.5 Critical operation of the unit shat! not be resumed until authorized by the Commission.
4 Amendment No. 45, 27,12, 72,133,146, 171 2-1
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2.0 SAFETY LIMITS INTRODUCTION The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers to the releasa of radioactive materials to the environs. Safety Limits are established to protect l
the integrity of these barriers during normal plant operations and antic %ated transients. The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a stepback approach is used to establish a Safety Limit such that the Minimum Critical Power Ratio (MCPR) is not less than the limit specified in Specification 2.1.2. MCPR greater than the j
specified limit represents a conservative margin relative to the conditions required to maintain fuel cladding integrity.
The fuel cladding is one of the physical barriers which separate j
the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking. Although some corrosion or use-related cracking may occur during the life of the cladding, fission product
. migration from this source is incrementally cumulative and i
continuously measurable. Fuel cladding perforations, however, o
can result from thennal stresses which occur from reactor
]
operation significantly above design conditions.
While fission product migration from cladding perforation is just i
as measurable as thet from use-related cracking, the thermally caused c! add;ng perforations signal a thresho!d beyond which still j
greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding Safety Umit is defined with a margin to the conditions which would produce onset of transition boiling (i.e., MCPR of 1.0).
y.
These conditions represent a significant departure from the condition intended by design for planned operation. The MCPR fuel cladding integlity Safety Limit assures that during normal operation and during anticipated operational occurrences, at least 99.9% of the fuel rods in the core do not experience transition boiling.
Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperature because of the onset of transition boiling and the resultant sharp reduction in i
heat transfer coefficient. Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled rolease of activity to the reactor coolant.
. GE critical power correlations are applicable for all critical power l
INTEGRITY (2.1.1) calculations at pressures at or above 785 psig or core flows at or
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above 10% of rated flow. For operation at low pressures and low J
flows another basis is used as follows:
(Cont)I Amendment No. 15, 42, 72,1M,129,133, 171 82-1 i
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2.0 SAFETYjJMITS (Cont)
FUEL CLADDING Since the pressure drop in the bypass region is essentially all INTEGRITY (2.1.1) e!avation haad, the core pressure drop at low power and flows (Cont) will always be greater than 4.5 psi. Analyses show that with a 3
bundle flow of 28 x 10 lbs/hr, bundle pressure drop is nearly independent of bundle pcw::r and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be greater than 28 3
x 10 lbs/hr. Full scale ATLAS tett data taken at pressures from t
14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the des'gn peaking factors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POiNER. Thus, e THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.
MINIMUM The Safety Limit MCPR is determined using the General Electric CRITICAL Thermal Analysis Basis, GETAB (2), which is a statistical model POWER RATIO that combines all of the uncertainties in operating parameters and (2.1.2) the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the General Electric Critical Quality (X) - Boiling Length (L), GEXL, correlation.
The GEXL correlation is valid over the range of conditions used in the tests of the dcta used to develop the correlation. These conditions are:
Pressure:
800 to 1300 psia Mass Flux:
0.1 to 1.5 Mib/hr-ft2 Inlet Subcooling:
0 to 70 Btu /lb Axial Profile:
1.5 chopped cosine 1.7 intet peaked 1.7 outlet peaked R Factor 0.95 to 1.20 Rod Array 9X9 GE 11 array The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated. Since the parameters which result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur. Although it is recognized that a departure from nucleate boiling would not result in damage to BWR fuel rods, the critical power at (Cont)
Amendment No. 46;-42,72,105,139,133,165, 171 82-2
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SASES:
~2.0 SAFETY LIMITS (Cont)
MINIMUM
. which boiling transition is calculated to occur has been adopted
. CRITICAL as a convenient limit. However, the uncertainties in monitoring POWER RATIC
- he cora operating state and in the procedures used to calculate (2.1.2) (Cont) the critical power result in an uncertainty in the value of the entcal power. Therefore, the fuel cladding integrity Safety Limit is defined as the CPR in the limiting fuel assembly for which more
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than 99.9% of the fuel rods in the core are expected to avoid boilinC transition considering the power distribution within the core and all uncertainties.
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1 The Safety Limit MCPR is determined using a statist %I model 4
that combines all of the uncertainties in operating parameters and
. the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined ushg the i
approved General Electric Critical Power correlations. Details of i
the fuel cladding integrity Safety Limit calculation are given in i
Reference 1. Reference 1 includes a tabulation cf the uncertainties used in the determination of the Safety Limit MCPR and of the nominal values of ihe parameters used in the Safety Limit MCPR statistical analysis.
1 REACTOR With fuel in the reactor vessel during periods when the reactor is WATER shutdown, consideration must be given to water level LEVEL (Shutdown requirements due to the effect of decay heat. If reactor water Condition) level should drop below the top of the active fuel during this time, (2.1.3) the ability to cool the core is reduced. This reduction in core cooling capability could lead to elevated cladding temperatures and clad perforation. The core can be cooled sufficiently should
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the water level be reduced to two thirds the core height.
Establishment of the safety limit at 12 inches above the top of the fuel provides adequate margin. This level will be continuously monitored.
I (Cont)
)
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I
. Amendment No. 44, 42, 72,133. 171 82-3
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2.0 SAFETY 8 MTS INTRODUCTION The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers to the release of rad:oactive materials to the environs. Safety Umits are established to protect the integrity of these baniers during normal p' ant operations and
^
anticipated transients. The fuel cladding integrity Safety Limit is
(
set such :t no fuel damage is calculated to occur if the limit is not violacco. Because fuel damage is not directly observable, a
{
stepback approach is used to establish a Ssfety Limit such that
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the Minimum Cdtical Power Ratio (MCPR) is not less then the limit specified in Specification 2.1.2. MCPR greater than the specified I;mit represents a conservative margin relative to the conditions required to maintain fuel dadding integnty.
j The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freadem from perforations t
i or cracking. Although some corresion or use-related cracking may occur during the life of the cladding, fission product migration from this source is inctsmentally cumulative and 1
continuously measurable. Fuel cladding perforations, however, 3
can result from thermal stresses which occur from reactor j
operation significantly above design conditions.
1 While fission product migration from cladding perforation is just as mea;urable as that from use-related cracking, the thermally i
caused claddin; perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. Tnerefore, the fuel cladding Safety Umrt is defined with a margin to the conditions which would produce onset of transition boiling (i.e., MCPR of 1.0).
These conditions represent a significant departure from the condition intended by design for planned operation. The MCPR fuel dadding integrity Safety Limit assures that during normal t
operation and during anticipated operational occurrences, at least 99.9% of the fuel rods in the core do not experience transition boiling.
Operation above the boundary of the nucleate boiling regime could result in excessive cladding teniperature because of the onset of transition boiling and the resultant sharp reduction in heat transfer coefficient. Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally waaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.
FUEL CLADDING GE critical power correlations are applicable for all critical power 4
INTEGRITY (2.1.1) calculations at preswres at or above 785 psig or core flows at or above 10% of rated flow. For operation at low pressures and low j
flows another basis is used as follows:
4 i
. (Cunt)l
- Amendment No.15, c. 72,1^",129, ???, 171 82-1 1
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2.0 SAFETY LIMITS (Cont)
FUEL CLAD 0 LNG Since the pressure drop in the bypass region is essentially all INTEGRITY (2.1.1) elevation head, the core pressure drop at low power and flows (Cont) will always be greater plbs/hr, bundle pressure drop is nearly an 4.5 psi Analyses show that with a bundle flow of 28 x 10 independent of bundle pc./:r and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be greater than 28 3
x 10 lbs/hr. Full scale ATI.AS test data taken at pressures from 14.7 psia to 800,;sia indicatc that the fuel assembly critical power at this flowis approximately 3.35 MWt. With the design i
peaking factors, :his corresponds to a THERMAL POWER of 1
more than 50% of RATED THERMAL POWER. Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.
i MINIMUM The Safety Limit MCPR is determined using the General Electric CRITICAL Thermal Analysis Basis, GETAB (2), which is a statistical model i
POWER RATIO that combines all of the uncertainties in operating parametsrs and 1
(2.1.2) the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the General Electric Critical Quality (X) - Boiling Length (L), GEXL, correlation.
i i
The GEXL correlation is valid over the range of conditions used l
in the tests of the data used to devrlon the correlation. These conditions are:
Pressure:
800 to 1300 psia i
Mass Flux:
0.1 to 1.5 Mib/hr ft2 Inlet Subcooling:
0 to 70 Stu/lb
{
Axial Profile:
1.5 chopped cosine 1.7 inlet peaked 1.7 outlet peaked R Factor 0.95 to 1.20 Rod Array 9X9 GE 11 array l
The fuel cladding integrity Safety Limit is set such that no fuel damage is calcuieted to occur if the limit is not violated. Since l
the parameters which result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic l
conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur. Although it is recognized that a departure t
from nucleate boiling would not result in damage to BWR fuel tods, the critical power at (Cont) 4 Amendment No. 44,-42r7h-405.129,133.195, 171 E2-2
l E8EEI' 2.0 SAFETY LIMITS (Cont) i
- MINIMUM which boiling transit lon is calculated to occur has been adopted CRITICAL as a convenient limit. However, the uncertainties in monitoring POWER RATIO the core operating state and in the procedures used to calculate (2.1.2) (Cont) the critical power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity Safety Limit is defined as the CPR in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertainties.
The Safety Limit MCPR is determined using a statistical model that combines all of the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the approved General Electric Critical Power correlations. Details of the fuel cladding integrity Safety Limit calculation are given in j
Reference 1. Reference 1 includes a tabulation of the uncertainties used in the determination of the Safety Lir tit MCPR and of the nominal values of the parameters used in the Safety Limit MCPR statistical analysis.
REACTOR Wth fuel in the reactor vessel during periods when the reactor is WATER shutdown, consideration must be given to water level LEVEL (Shutdown requirements due to the effect of decay heat. If reactor water Condition) level shouid drop below the top of the active fuel during this time, (2.1.3) the ability to cool the core is reduced. This reduction in core cooling capability could lead to elevated cladding temperatures and clad perforation. The core can be cooled sufficiently should l
the water level be reduced to two thirds the core height.
i Establishment of the safety limit at 12 inches above the top of the i
i fuel provides adequate margin. This level will be continuously monitored.
(Cont) e a
I L
Amendment No.
171 B2-3
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BAEEE:
2.0 SAFETY LIMITS (Cont)
REACTOR STEAM The Safety Limit for the reactor steam dome pressure has been DOME PRESSURE selected such that it is at a pressure below which it can be shown (2.1.4) that the integrity of the system is not endangered. The reactor pressure vesselis designed to Section illof the ASME Boiler and Pressure Vessel Code (1965 Edition, including the January 1966 Addendum), which permits a maximum pressure transient of 110%,1375 psig, of design pressure 1250 psig. The Safety Limit of 1325 psig, as measured by the reactor steam dome pressure indicator, is equivalent to 1375 psig at the lowest elevation of the reactor coolant system. The reactor coolant system is designed to the USAS Nuclear Power Piping Code, Section B31.1.0 for the
. reactor recirculation piping, which permits a maximum pressure transient of 120% of design pressures of 1148 psig at 562*F for suction piping and 1241 psig at 562'F for discharge piping. The pressure Safety Limit is selected to be the lowest transient overpressure allowed by the applicable codes.
REFERENCES 1)
" General Electric Standard Application for Reactor Fuel,"
NEDE-24011 P-A (Applicable Amendment specified in the CORE OPERATING LIMITS REPORT).
2)
General Electric Thermal Analysis Basis (GETAB): Data, Correlation and Design Application, General Electric Co, BWR Systems Department, January 1977, NEDE-10958-PA and NEDO-10958-A.
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s Revision 177 Amendment No.15,132, MS, 171 B2-4
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NOTES FOR TABLE 3.2.C-1
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. Wdh the number of operable channels:
a.
One bss than required by the minimum operable channels per trip function -
P requirement, restore an inoperable channel to operable status within 7 days or place an inoperable channel in the tripped condition within the next hour.
l b.
Two or more less than required by the minimum operable channels per trip function requirement, place at least one inoperable channel in the tripped condition within one hour.
2.
's.
Wdh one RBM Channelinoperable:
(1) restore the inoperable RBM channel to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; 1
otherwise place one rod block monitor channe! in the tripped condition within the next hour, and; j.
(2) prior to control rod withdrawal, perform an instrument function test of the operable RBM channel.
t b.
Wdh both RBM channels inoperable, place at least one inoperable rod block monitor channel in the tripped condition within one hour.
3.
If the number of operable channels is less than required by the minimum operable channels
-1 per trip function requirement, place the inoperable channel in the tripped condition within one
- hour, e
4.
SRM operability requirements during core alterations are given in Technical Specifmation 4
3.10.
5.
RBM operability is required in the run mode in the presence of a limiting rod pattem with reactor power greater than the RBM low power setpoint (LPSP). A limiting rod pattern exists when:
MCPR < 1.41 for reactor power 190%
MCDR < 1.72 for reactor power < 90%
The allowable value for the LPSP is < 29% vi rated core thermal power.
6.
When the reactor mode switch is in the Refuel pos, tion, the reactor vessel head is removed, and control rods are inserted in all core cells containing one or more fuel assemblies, these j.
Rod Block functions are nct required.
7.
With one or more Reactor Mode Switch - Shutdown Position channels inoperable, suspend control rod withdrawal and initiate action to fully insert all insertable control rods in core cells containing one or more fuel assemblies immediately.
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Amendment No. 15, 27, 42. SS, "'7,' 1^^,1 ?S 199'. 171 3/4.2 22 l
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