ML20245D403
| ML20245D403 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 04/21/1989 |
| From: | DUQUESNE LIGHT CO. |
| To: | |
| Shared Package | |
| ML20245D398 | List: |
| References | |
| NUDOCS 8904280288 | |
| Download: ML20245D403 (22) | |
Text
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y ATTACHMENT A-1
' Revise the Beaver -Valley Unit No.
1 Technical Specifications as follows:-
i Remove Paces Insert Paaes l:
3/4-4-23 3/4 4-23 l
3/4 4-26 B 3/4 4-10 B 3/4 4-10 j
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'l 8904280288 890421 PDR ADOCK 05000334 P
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. REACTOR COdLANT SYSTEM
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SURVEILLANCE REQUIREMENTS 4.4.9.1 a.
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.
b.
The Reactor Coolant System temperature and pressure conditions shall be determined to be to the right of the criticality ' limit line within 15 minutes' prior to achieving reactor criticality.
BEAVER VALLEY - UNIT 1 3/4 4-23 PROPOSED WORDING
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BASES bessel insiae raaius are essentially laentical, tne measurea transition shift for a sample can be applied with confidence to the adjacent section of the reactor vessel.
The heatup and cooldown curves must be recalculated when the o RT determined from the NDT surveillance capsule is different from the calculated ART f0E NDT the equivalent capsule radiation exposure.
The pressure-ter.perature limit lines shown on Figure 3.4-2 for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature
)
requirements of Appendix G to 10 CFR 50 for reactor criticality and i
for inservice leak and hydrostatic testing.
The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in UFSAR Table 4.5-3 to assure compliance with the requirements of l
Appendix H to 10 CFR 50.
The limitations imposed on the pressurizer heatup and cooldown rates and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements.
The OPERABILITY of two PORV's or an RCS vent opening of greater than 3.14 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are $ 275'F.
Either PORV has adequate relieving capability to protect the RCS from over-pressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of the steam generator 5 25'F above the RCS cold leg temperature or (2) the start of a charging pump and its injection into a water solid RCS.
3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2,
and 3
components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant.
These programs are in 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 Commission pursuant to 10 CFR Part 50.55a(g) (6) (i).
3/4.4.11 RELIEF VALVES The relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoperable.
The electrical power for both the relief valves and the block valves is capable of being supplied from an emergency power source to ensure the ability to seal this possible RCS leakage path.
BEAVER VALLEY - UNIT 1 B 3/4 4-10 PROPOSED WORDING
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ATTACHMENT A-2 Revise the Beaver Valley Unit No.
2 Technical Specifications as
-follows:
Remove _Pagen Insert Paaes
-3/4 4-30 3/4.4-30 3/4 4-33 3/4 4-33 B 3/4 4-14 B 3/4 4-14 l'
B 3/4 4-15 B 3/4 4-15
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. REACTOR'C00LANT SYSTEM l
3/4.4.9 PRESSURE / TEMPERATURE LIMITS I
REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION i
3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and j
pressure shall be limited in accordance with the limit lines shown on i
Figures 3.4-2 and 3.4-3 during heatup, cooldown, criticality, and inservice 1
leak and hydrostatic testing with:
{
i a.
A maximum heatup of 60 F in any 1-hour period,
.b.
A maximum cooldown of 100 F in any 1-hour period, and l
c.
A maximum temperature change of s 5 F in any 1-hour period during i
hydrostatic testing operations a5ove system design pressure.
APPLICABILITY:
MODES 1, 2, 3, 4, and 5.
ACTION:
With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an analysis 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 acceptable for continued operations or be in at least HOT STANDBY within the i
next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the RCS T,yg and pressure to less than 200 F and 500 psig, respectively, within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.4.9.1 a.
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.
b.
The Reactor Coolant System temperature and pressure conditions shall be determined to be to-the right of the criticality limit line within 15 minutes prior to achieving reactor criticality.
t --T" raattor vessel material irradiation surveillance s as--s h a t T ~
be removed and Hamine&g-to_.delstrmi age material properties, at the interval ts_nL 1hese examina-
.__.. t4on e used to update Figures 3.4-2 and 3.4-3.
I BEAVER VALLEY - UNIT 2 3/4 4-30 PICOPOSED WOEbh06-
6 t
, TABLE 4.4-5 REACTOR VESSEL MATERIAlf IRRADIATION SUld/EILLANCE SCHEDULE.
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2
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Estimat d Vessel Lead Withdrawal
-Capsule fluenc 2
Capsule Location factor Time (EFPY)
(n/cm )
U 343 3.5
1st Refueling-0.8 x 102 V
07 3.5 3
2.13 x 028 X
28 3.5 6
4.2 x 1029(a)
W 110 2.9 11
.48 x 1029(D)
Y 290 2.9 20 11.77 x 1029 2
340 2.9 Standby (a) Approximate fluence at 1/4 vessel wall thickness a end of-life.
(b) Approximate fluence at vessel inner wall at end-of ife.
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BEAVER VALLEY - UNIT 2 3/4 4-33 bELET& 7$/5 h 5
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BASES NN 3/4.4.9 PRESSURE / TEMPERATURE LIMITS (Continued)
The pressure-temperature limit lines shown on Figure 3.4-2 for reactor criticality and for inservice leak and hydrostatic testing have been provided i
to assure compliance with the minimum temperature requirements of Appendix G j
to 10 CFR 50 for reactor criticality and for inservice leak end hydrostatic testing.
The number of reactor vessel irradiation surveillance specimens and the
/
f requencies for removing and testing these specimens are provided in-Tabic 4.4 1 3
to assure compliance with the requirements of Appendix H to 10 CFR Part 50.
j The limitations imposed on the pressurizer heatup and cooldown rates and j
auxiliary spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements.
The OPERABILITY of two PORVs or an RCS vent opening of greater than 3.14 square inches ensures that the RCS will be protected from pressure tran-sients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are < 350 F.
Either PORV has adequate relieving capability to protect the RCS froiii overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water tempera-ture of the steam generator < 50 F above the RCS cold leg temperature or (2) the start of a charging Jiump and its injection into a water solid RCS.
OVERPRESSURE PROTECTION SYSTEMS ThepaximumAllowedPORVSetpointfortheOverpressureProtectionSystems (OPPS) is derived by analysis which models the performance of the OPPS assuming various mass input and heat input transients.
Operation with a PORV setpoint less than or equal to the maximum setpoint ensures that nominal 10 EFPY Appen-dix G limits will not be violated with consideration for:
(1) a maximum pressure overshoot beyond the PORV setpoint which can occur as a result of time delays in signal processing and valve opening; (2) a 50 F heat transport effect made possible by the geometrical relationship of the RHR suction line and the RCS wide range temperature indicator used for OPPS; (3) instrument uncertainties; and (4) single failure.
To ensure mass and heat input transients more severe than those assumed cannot occur, Technical Specifications require lockout of all but one centrifugal charging pump while in H0 DES 4, 5, and 6 with the reactor vessel head installed and disallow start of an RCP if secondary coolant temperature is more than 50 F above reactor coolant temperature.
Exceptions to these requirements are acceptable as described below.
Operation above 350 F but less than 375 F with only one centrifugal charg-ing pump OPERABLE is allowed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
As shown by analysis LOCAs occurring at low temperature, low pressure conditions can be successfully miti-gated by the operation of a single centrifugal charging pump and a single LHSI pump with no credit for accumulator injection.
Given the short time duration i
BEAVER VALLEY - UNIT 2 B 3/4 4-14 l
(WOf05Eb WORD /NG
t REACT 0'R' COOLANT SYSTEM
'l BASES OVERPRESSUREPROTECTI0ISYSTEMS(Continued) that the condition of having only one centrifugal charging pump OPERABLE is
' allowed and the probability of a LOCA occurring during this time, the failure of the single centrifugal charging pump is not assumed.
Operation below 350'F but greater than 325*F with all centrifugal charging pumps OPERABLE is allowed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
During low pressure, low tempera-ture operation all automatic Safety Injection actuation signals are blocked.
In normal conditions a singlie failure of the ESF actuation circuitry will result in the. starting of at most one train of Safety Injection (one centrifugal charging pump, and one LHSI pump).
For temperatures above 325'F, an overpress-ure event occurring as a result of starting these two pumps can be successfully mitigated by operation of both PORVs without exceeding Appendix G limit. Given.
the short time duration that this condition is allowed and the low probability of a single failure causing an overpressure event during this time, the single failure of a PORV is not assumed.
Initiation of both trains of Safety Injection during this 4-hour time frame due to operator error or a single failure occurring during testing of a reaundant channel are not considered to be credible accidents.
The maximum allowed PORV setpoint for the Overpressure Protection System will be updated based on the results of examinations of reactor vessel material irradiation surveillance specimens performed as required by 10 CFR Part 50, Appendix H and in accordance with the schedule in Table 4.4 5.
l 3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2
and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant.
These programs are in 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 Commission pursuant to 10 CFR Part 50.55a (g)(6)(i).
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BEAVER VALLEY - UNIT 2 B 3/4 4-15 fgoPOSEb t40NDl&
1 b
s ATTACHMENT B
Safety Analysis Beaver Valley Power Station Proposed Technical Specification Change Unit No. 1 Change No. 164 Unit No. 2 Chance No. 23 l
Description of amendment request:
The proposed amendment would relocate Table 4.4-5 and the requirements of surveillance requirement 4.4.9.1.c from the technical specifications to the Updated Final Safety Analysis Report (UFSAR).
Surveillance Requirement 4.4.9.1.c provides for the removal and examination of the reactor vessel surveillance capsules to l
update the plant heatup and cooldown curves.
A change to the BV-1 l
UFSAR Section 4.5.1.2 is provided by Insert 3 (attached) and the same change to the BV-2 UFSAR Section 5.3.1.6 is identified on page 5.3-7 to incorporate these requirements.
Table 4.4-5
" Reactor Vessel Material Irradiation Surveillance Schedule "provides the capsule removal schedule including vessel
- location, lead factors and estimated capsule fluence.
A change to BV-1 UFSAR Table 4.5-3 (Insert 1
attached) incorporates this information which has been updated to provide the latest capsule evaluation data from WCAP-12005
" Analysis of Capsule W from the Duquesne Light Company Beaver Valley Unit 1
Reactor Vessel Radiation Surveillance Program."
A change to the BV-2 UFSAR incorporates a
new Table 5.3-6 (attached) which provides the information from technical specification Table 4.4-5.
The technical specification bases for both units have been revised to replace reference to technical specification Table 4.4-5 with the appropriate UFSAR table.
BV-1 UFSAR Section 4.5.1.2 has also been revised to include Insert 2
which replaces the discussion applicable to Capsule V with reference to the capsule analysis reports identified in the references for Section 4.5.
Therefore, the information provided in UFSAR Tables 4.5-4 and 4.5-5 is no longer needed and has been deleted.
WCAP-12005 for Capsule W
has also been added to the references for UFSAR Section 4.5.
Similar changes to the BV-2 UFSAR may be incorporated following capsule removal and analysis.
Relocating the reactor vessel irradiation surveillance schedule to the UFSAR will not affect the reactor vessel surveillance program.
The program for surveillance of reactor vessel material will continue to be governed by 10 CFR 50 Appendix H.
This change will not alter any plant configuration or mode of operation.
Compliance with existing regulations will ensure continued confidence in the reactor vessel material properties.
The table being removed from the technical specifications will be retained in the UFSAR and any future changes will require a safety analysis in accordance with 10 CFR 50.59.
Therefore, the proposed change is considered to be administrative in nature and will not reduce the safety of the plant.
s ATTACHMENT C
No Significant Hazards Evaluation Beaver Valley Power Station Proposed Technical Specification Change Unit No. 1 Change No. 164 Unit No. 2 Chance No. 23 Basis for proposed no significant hazards consideration determination:
The Commission has provided standards for determining whether a significant hazurds consideration exists in accordance with 10 CFR 50.92(c).
A proposed amendment to an operating license for a facility involved no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or.(3) involve a significant reduction in a margin of safety.
The proposed changes do not involve a
significant hazard consideration because:
1.
The proposed amendment will remove surveillance requirement 4.4.9.1.c and Table 4.4-5 from the technical specifications and incorporate them into BV-1 UFSAR Section 4.5 and BV-2 UFSAR Section 5.3.1.6.
The BV-1 capsule removal schedule has been incorporated into UFSAR Table 4.5-3 which has been revised to provide the latest capsule evaluation data from WCAP-12005
" Analysis of Capsule W
from the Duquesne Light Company Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program."
A change to the BV-2 UFSAR incorporate a new Table 5.3-6 which provides the information from technical specification Table 4.4-5.
The proposed change will not affect the reactor vessel material surveillance program.
Implementation of the proposed change will remove a
technical specification requirement that is redundant to the Code of Federal Regulations.
Therefore, these changes are considered to be administrative and do not significantly increase the probability or consequences of an accident previously evaluated.
2.
These changes will not alter any plant configuration or mode of operation.
Compliance with existing regulations I
will ensure continued confidence in the reactor vessel material properties.
Therefore, the proposed change will 1
not create the possibility of a new or different kind of accident from any previously evaluated.
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g 1Atta^channt C Page 2 3.
Evaluation of the Jeactor vessel material, with respect to radiation embrittlement will not be altered by these
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changes.-
Surveillance Requirement 4.4.9.1.c and Table 4.4-5 do not aid the reactor.operatorfas the primary user of
'the technical specifications.
The surveillance requirement and capsule removal schedule will be retained in the UFSAR and.any future changes will require a safety
')
evaluation in accordance ~ with 10 CFR 50.59.
The 4
technical specification:
bases have been revised to1 identify the applicable UFSAR Table for reference to the y
capsule removal schedule.
Therefore, the proposed change 1
will not involve a significant reduction in the margin of safety.
q l
Based on the above considerations, these changes are j
considered to be administrative in nature and do not involve a i
significant hazards consideration.
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ATTACHMENT D-1
.q Beaver Valley Power Station, Unit No.:1 Proposed. Technical Specification Change 164 i'
i Information' Copies of (TSAR Chanaes l
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BVPS-1-UPDATED FSAR-Rev. 6-(1/8'8)~
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U288 (Cadmium shielded)
.Np2" (Cadmium shielded) t Thermal Monitors 97.5% Pb, 2.5% ng (579 F Melting Point) 97.5% Pb, 1.72% Ag, 0.75% Sn (590 F Melting Point)
The ' chemistry and heat treatment of the surveillance material are provided in Table 4.5-2.
The fast neutron exposure of the specimens occurs at a faster rate than that experienced by the vessel wall with the specimens being located between the core and the vessel and with the sequenced removal and reinsertion of capsules as n'.'ted in the tentative schedule.
Since these specimens experience accelerated exposure and are actual samples from the materials'used in the vessel, the NDTT measurements are representative of the vessel at a later time in life.
Data from fracture toughness samples (WOL) are expected to provide additional information for use in determining allowable stresses for irradiated material.
ated lead factors for all of the Beaver Valley Powe yr Stion
/Th Unit No.
illance
- capsules, as taken fro eference 1 are provided in Table 4.5 The revised heat nd limit curves eration /
developed
-ro illance
- data, are provided in
" nical/
S ect 1 cation.
' ~fffL/f(,l~ WIT /f^W$M7 3~
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Correlations between the calculations and the measurements on the irradiated samples in the
- capsules, assuming the same neutron spectrum at the samples and the vessel inner wall, are described in Appendix 4A.
-~.- -
f, lux agreement between calculation and measurement of fast neu f
a luence levels, as taken from Reference 1, are ed in '
Table 4.5-4.
more recent measured and calcula alues see the most recent capsule sis report ident ith the references for Section 4.5.
Using the data derived fo on along wit ead factor determined for capsule V,
t rrent and end of life fast on fluence for capsule Va
-actor vessel are summarized in Table 4.5 or more /
recen a
see the most recent capsule analysis report identi
/
the references for Section 4.5.
The anticipated degree to which the specimens will perturb the fast neutron flux and enerE~
distribution will be considered in the evaluation of the surveillance specimen data.
Verification and possible readjustment of the calculated wall exposure will be made by
/
use of data on all capsules withdrawn.
A E fl. ) C E hil T}/ 3 1/S M T 2 4.5-6
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INSERT 1.
REACTOR VESSEL MATERIAL IRRADIATION SURVEILLANCE SCHEDULE Estimated-
{
Vessel Capsule Location Lead Removal Fluence Capsule (deg)
Factor Time (a)
(n/cm/cm)
V 165 1.65 1.16 (removed) 2.91 x E18 (b)
U 65 1.10 3.59 (removed) 6.54 x E18 (b)
W 245 1.10 5.89 (removed) 9.49 x E18 (b)
Y 295 1.10 15 2.10 x E19 (r.)
X 285 1.65 32 6.72 x E19 T
55 0.70 Standby Z
305 0.70 Standby l
S 45 0.54 Standby (a) Effective full power years from plant startup (b) Actual fluence (c) Approximate fluence at vessel 1/4 thickness at end of design life (32 EFPY)
NOTE : This information was obtained from Section 4.5 Reference 6
~
INSERT 2 For a detailed evaluation of the capsule results see the capsule analysis reports identified in the references for Section 4.5.
INSERT 3 The reactor vessel material' irradiation surveillance' capsules are removed and examined, to determine changes ^ in material properties, at the intervals shown in Table 4.5-3.
The he~atup and-coo.ldown limit curve's for normal-operation are developed-from.these, examinations and are..pr6videdfin.ithe.
tech 6ical speelfications.
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Rev. 6 -(1/88 )
BVPS-1-UPDATED FSAR l
N the.new. capsule to vessel inner: wall lead fac ftors'ved' i
a
'from?'the. cap T--V-Jan ' sis 9 data, provided' irLRe.ference 1,..' andfthe
(-(?
new withdraw schedule ide 82, a revised'dapsule.
withdraw sche n
blishe The revised sche'dule is provided in the chnical_
4.5.1.3 Electroslag Weld Quality Assurance The 90 degree elbows used in the reactor coolant loop piping'are electroslag welded.
The following efforts were performed for quality assurance of these components.
1.
The electroslag welding procedure employing one wire technique was qualified in accordance with the requirements of ASME Boiler and Pressure Vessel Code Section IX and Code Case 1355 plus supplementary evaluations as requested by WNES-PWRSD.
The following test specimens were removed from a 5 inch thick weldment and successfully tested.
They are:
a.
6 Transverse Tensile Bars - as welded b.
6 Transverse Tensile Bars - 2050 F, H O Quench 2
F, H O Quench +
2050 c.
6 Transverse Tensile Bars 2
750 F stress relief heat treatment I
d.
6 Transverse Tensile Bars H O Quench, 2050 F,
2 tested at 650 F e.
12 Guided Side Bend Test Bars.
2.
The casting segment s were surface conditioned for 100 percent radiographic and penetrant inspections.
The acceptance standards were ASTM E-186(2) severity level 2,
except no category D
or E
?sfectiveness was permitted and USAS Code Case N-10, respectively.
3.
The edges of the electroslag weld preparations were machined.
These surfaces were penetrant inspected prior to welding.
The acceptance standards were USAS Code case N-10.
4.
The completed electroslag weld surfaces were ground flush with the casting surface.
Then, the electroslag weld and-adjacent base material were 100 percent radiographer in accordance with ASME Code Case 1355.
ilso, the electroslag weld surfaces and adjacent base material were penetrant inspected in accordance with USAS Code Case N-10, 5.
Weld metal and base metal chemical and physical analysis were determined and certified.
(
6.
Heat treatment furnace charts were recorded and certified.
j 4.5-7 N --- -___ - _- -_ _ - _-
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BVPS-1-UPDATED FSAR Rev. 6 (1/88)
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TABLE 4.5-3
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/ CALCULATED FAST NEUTRON FLUX (E >1.0 Mev) ANDLEADFACTO[
/
FOR BEAVER VALLEY UNIT 1 SURVEILLANCE CAPSULES
/
/
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Capsule Azimuthal 4(E > 1.0 Mev)
Lead 2
Factor l
(n/cm -sec)
Identification Location f
V 15' 8.00 x 10 1.37
\\
X 15 8.00
.101' 1.37 j
U 25*
5 2 x 101' O.89 W
25' 5.22 x 102' O.89 Y
25' 5.
x 102' O.89 I
T 35' 3.42 x l'
O.58 i
Z 5'
3.42 x 102' O.58
\\
j S
45*
2.51 x 1010 0.43
/
The values given were calculated for capsule V.
Fo ore r
ent data see the most recent capsule analysis rep t
dentified with the references for Section 4.5.
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2 2
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8 e
e 8
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t(
1 1
t 1
1 0
u d
0 0
0 0
0 e
e 1
1 1
1 1
N$
r
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x x
x x
x e
s R
a 5
0 8
7 5
e 5
6 1
1 1
M 2
2 3
3 3
V EL 8
8 U
d 1
1 1
1 l
S e
0 0
0 0
P t
1 1
1 1
A a
C l
x x
x x
x u
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R xvc c
0 0
0 O
0 O
uee l
0 0
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F lMs a
F C
8 8
8 8
R Y
n.-
0 A
R S
T o1 F
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m
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4 M
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1 1
1 1
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S e( n d
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1 1
1 1
T A
4 D
r D
u x
x x
x P
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s U
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a 8
6 9
B R
e 9
8 8
7 1
A T
M T
U 7
8 9
9 9
S E
P N
V y
B T
t S
i d
7 5
7 A
v e
0 0
0 0
0 F
i t
1 1
1 1
t a
F c
l x
x x
x x
uN5 O
A c
5 5
6 5
1 2
1 4
7 S
d T
e)
L tm 4
6 3
3 4
U ag S
r/
E us R
tP ad 5
7 5
7 S
d 0
0 0
0 0
e 1
1 1
1 1
d r
e u
x x
x x
x t
s s
a 0
1 6
4 1
u e
2 2
2 1
8 j
M 4
6 4
4 5
A 7
3 7
e 1
8 s
1 5
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s C
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C C
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p a,
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3 s
2 s
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..R:v. 0 (1/82)
TABLE 4'.5'-5 CURRENT AND END OF LIFE FAST NEUTRON FLUENCE FOR CAPSULE V AND REACTOR VESSEL
/
Current Fluence EOL Fluence 4
(E > 1.0 v)
N t(E > 1.0 Mev)
(n/cm/
t 2
(n/cm )
Location Mea ured Calculated Measured Calculated 18 Capsule V 2.55 x One 2.58 x 10 1
8 2e 18 5.90 x 10 '
Vessel IR*
1.86 x 10 1.88 x 10 5.84 x 10 Vessel 1
18 18
.58 x 10**
3.61 x 10 '
1/4T**
1.14 x 10
.15 x 10 Vessel 17 1
18 se 3/4T***
2.54 x 10 2.5 x 10 7.97 x 10 8.06 x lo
(
NOTE:
EOL Fluences are based on eration at 2766 MWT For 32 Effective Full Power Yeafs
/
- Approximate EOL Fluence It Reacto Vessel Inner Wall
- Approximate EOL Fluence /at Reactor Vessel 1/4 Thickness Location j
- Approximate EOL Fluence at Reactor V ssel 3/4 Thickness Location
/
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/
/
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BVPS-1 UPDATE 6 FSAR'
, RavJfL6 (1/88).
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e.
s
- .s -,..,
- ]h.' T.' l efefences fo'r Sect!ioh'"4.5' i-
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7 1,,
'l..., ', / ~ '
e,.
,e.v
~
O 1.
S.
E.
Yanichko,-
S..
L.
Anderson; W.
' T.
Kaiser, "An'alyses of Capsule V
fr.om 'the, Duquesne;1 Light l Company'BeaveriValley Poser Radiation ~ -Surveillance
- Station ' Unit 'N.on l
~l' Reactor Vessel (Electri'c' Corp' ration 1(January, Wes'iinghouse Program",- WCAP-9860 t
o 1981).
~
2.
" ASTM Heavy-Walled 2
to 4-1/2 inch (Si to 114 mm)
Steel
-Castings",
ASTM E-186, The -American Society for Tests and-Materials.
3.
" ASTM Standard' Recommended Practice for Radiographic Testing",
ASTM E-94, The American Soc'iety for Tests and Materials.
4.
" ASTM Heavy-Walled (4-1/2 to 12 inch (114 to 305 mm) Steel Castings",
ASTM E-280, The American Society for Tests and Materials.
5.
R.S.
- Boggs, S.
L. Anderson, W.T. Kaiser, " Analysis of Capsule U from the Duquesne Light Company Beaver Valley Power Station Unit No.
1 Reactor Vessel Radiation Surveillance Program", WCAP -
10867, Westinghouse Electric Corporation (September 1985).
6.
S.
E. Yanichho.
S.
L. Anderson.
L.
Al be r t.i n. " Analyses of Capsule W from the Duquesne Light Company Beaver Valley Power Station Uriit No. 1 Reactor Vensel Radiat. ion Surveillance Program". WCAP - 12005, Westinghouse Electric Corporation (November 1988) s t
4.5-14
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ATTACHMENT D-2 I
' Beaver Valley Power Station,. Unit No. 2 I
Proposed Technical Specification Change'23 Information Copies'of UFSAR Changes 1-I i
i 1
1
o 1~,-
BVPS-2 UFSAR
,0 Section 5.3.1.6.1.
They have ' indicated good a'greement.
The calculations of the integrated flux at the vessel wall are conservative. The anticipated degree to which the specimens will rerturb the fast neutron flux and energy distribution will be considered in the evaluation of the surveillance specimen data.
Verification and possible readjustment of the calculated wall exposure will be made by use of data on all capsules withdrawn.
he schedule for removal of the capsules for postirradiation testing conforms with ASTM E-185-73 and Appendix H of 10 CFR 50, 5.3.1.6.1 Measurement of Integrated Fast Neutron (E > 1.0MeV) Flux at the Irradiation Samples In order to effect a correlation between fast neutron (E > 1.0MeV) exposure and the radiation induced property changes observed in the test specimens, a number of fast neutron flux monitors are included as an integral part of the Reactor Vessel Surveillance Program.
In particular, the st.rveillance capsules contain detectors employing the following reactions:
{
g 7
g, gg 1.
Fe-54 (n,P) Mn-54, SWdYllaace Cafs u/es aR t% uw c'eql a ncl eYaml&, YE delWmi+1e Cbo[s os in gggj fjg, gg f,,je,yg/g 2.
Ni-58 (n,P) Co-58,
- 'd " ' #1 8 b 7' 6
- [
3.
Cu-63 (n,=) Co-60, osd Ccolclcuht li m// C w y e s S co^ floowa/
we c/et e/ogd hom f/gg 4.
Np-237 (n,f) Cs-137, and age,Ac[hn
/
(
GX am lefo/ib erS aordan ft%bdIN Ya i
5.
jeg4 gjf, g In addition, thermal neutron flux monitors, in the form of bare and cadmium shielded Co-Al wire, are included within the capsules to enable an assessment of the effects of isotopic burnup on the response of the fast neutron detectors.
The use of activation detectors such as those listed previously does not yiuld a direct measure of the energy dependent neutron flux level at the point of interest.
- Rather, the activation process is a measure of the integrated effect that the time and energy dependent neutron flux has on the urget material. An accurate estimate of the average neutron flux level incident on the various detectors may be derived from the activation measurements only if the parameters of the irradiation are well known.
In particular, the following variables are of interest:
1.
The operating history of the reactor, 2.
The energy response of the given detector, and 3.
The neutron energy spectrum at the detector location.
5.3-7 i
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___m_ _ _ _. _ - _ _ _ _. _ _ _ _ _ _ _ _
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BVPS-2 UFSAR j.._*:
~ TABLE 5'.3-6 REACTD'R' VESSEL MATERI Alf IRRADIATION SURVEILLANCE SCHE'DULE '
'E
?
Estimated Vessel-Lead Withdrawal Capsule Fluence 2
. Capsule-Location Factor Time (EFPY)
(n/cm )-
U 343
- 3. 5 1st Refueling-0.8 x 1019 2.13 x 1019 V
107 3.5 3-
'X 287 3.5 6
4.26 x'1019(a)-
W 110
' 2. 9 11 6.48 x 1019(b)
Y 290 2.9 20 11.77 x 1029 Z
340 2.9 Standby (a) Approximate fluence at 1/4 T vessel wall thickness at end-of-life.
(b) Approximate fluence at vessel inner wall at end-of-life.
1 of 1
- +
.