ML20136G265
| ML20136G265 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 03/10/1997 |
| From: | DUQUESNE LIGHT CO. |
| To: | |
| Shared Package | |
| ML20136G263 | List:
|
| References | |
| GL-95-05, GL-95-5, NUDOCS 9703170254 | |
| Download: ML20136G265 (21) | |
Text
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ATTACHMENT A l
a Beaver Valley Power Station, Unit No. 1 Proposed Technical Specification Change No. 240 1
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The following is a list of the affected pages:
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Affected Pages:
1-4 j
3/4 4-18 j
3/4 4-20 i
3/4 4-21 l
B 3/4 4-4
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B 3/4 4-5 l
B 3/4 4-2b j
B 3/4 4-3g B 3/4 7-3 m
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9703170254 970310 PDR ADOCK 05000334 P
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.DPR-66 DEFINITIONS 1
e c.
Pressure Boundary LEAKAGE shall be LEAKAGE (except steam generator tube LEAFAGE) through a nonisolable fault in a Reactor Coolant System component body, pipe wall or vessel j
wall.
1.15 THROUGH 1.17 (DELETED) l OUADRANT POWER TILT RATIO (OPTR) l i
1 l
1.18 QPTR shall be the ratio of the maximum upper excore detector i
calibrated output to the average of the upper excore detector I
calibrated outputs, or the ratio of the maximum lower excore detector j
calibrated output to the average of the lower excore detector i
calibrated outputs, whichever is greater.
4 A f PL M stryyy r g y,/
1 osE EOUIVALFMT I-131 j
1.19 DOSE EQUIVALENT I-171'shall be that concentration of I-131 4
(pci/ gram) which alone would produce the same thyroid dose as the quantity and isotopic mixtu're of I-131, I-132, I-133, I-134, and I-135 actually present.
The thyroid dose conversion factors used for this calculation shall be those listed in Regulatory Guide 1.109, 1977.
1 STAGGERED TEST BASIS 1.20 A STAGGERED TEST BASIS shall consist of:
a.
A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into n equal subintervals; b.
The testing of one (1) system, subsystem, train or other designated component at the beginning of each subinterval.
FREOUENCY NOTATION 1.21 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.
BEAVER VALLEY - UNIT 1 1-4 Amendment No. 499-(A pese/W,d,j)
l-4 INSERT A 4
DOSE EQUIVALENT I-131 1.19 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries/ gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.
The DOSE EQUIVALENT I-131 is calculated with the following equation:
C -132 C -133 C -134 C -135 I
I I
I
_. C -131 +
+
+
+
C -131p *g * -
I I
170 6
1000 34 Where "C"
is the concentration, in microcuries/ gram of the iodine isotopes.
This equation is based on dose conversion factors derived from ICRP-30.
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1
1 REACTOR COOLANT SYSTEM SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION i
3.4.8 The specific activity of the primary coolant shall be limited to:
O.3f a.
1 4w4uci/ gram DOSE EQUIVALENT I-131, and l
l b.
1 100/I uCi/ gram.
APPLICABILITY:
MODES 1, 2, 3, 4 and 5 ACTION MODES 1, 2, and 3*
o.3r a.
With the specific activity of the primary coolant
>4,4-l uC1/ gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in HOT STANDBY with T
<500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
avg b.
With the specific activity of the primary coolant > 100/I uC1/ gram, be in HOT STANDBY with T vg < 500*F within 6 a
hours.
MODES 1, 2, 3, 4 and 5 0,3r a.
With the specific activity of the primary coolant > +re-l uCi/ gram DOSE EQUIVALENT I-131 or > 100/Y uC1/ gram, perform the sampling and analysis requirement of item 4a of Table 4.4-12 until the specific activity of the primary coolant is restored te within its limits.
SURVEILLANCE REQUIREMENTS 4.4.8 The specific activity of the primary coolant shall be determined to be within the limits performance of the sampling and analysis program of Table 4.4-12.
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" With Tavg 2. 500*F BEAVER VALLEY - UNIT 1 3/4 4-18 (next page is 3/4 4-20) h4/#fM M[ah Amendment No. 0#,12:
IABl.E 4.4-12 l'Rif1ARY C00LArti SPtCil IC ACilVITY SAtlPLE cu9
' ~ ~ A.HI.F.Ah.Al'Ys..i s ~ ~P.liDIGR. At1~~~~~
lYPE OF MEASURElittil MINIMUM MODES IN WillCil s
Atl0 AtlALYSIS FREQUENCY SllRVEILLANCE REQUIRED 1.
Gross Activity Detentiination 3 times per 7 days with a 1, 2, 3, 4 maximum time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> a
g between samples.
[
2.
Isotopic Analysis for IJ0SE EQulVA-i per 14 days 1,
LEt4I l-131 Concentratioin 3.
Ha:lioclicitiical f or i Dett:rsitination I per 6 itioinths 1,
4.
Isotopic Analysis for lodine a) Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, 1,2,3,4,S locluding 1-131, 1-133, and 1-135 whenever the specific 4,,,
activity exceeds M O,3 r I
g 7, nC1/ gram DOSE EQUIVALErlT I-131 ay N ',
or 100/E pCi/ gram, and O
E b) One sample between 1, 2, 3 4
2 A 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following
(
a TilERMAL POWER change exceeding N',
15 percent of the RATED TilERMAL POWER within a one hour period.
thitil Llie specit ic activity of the priinary coolarit system is restored within its limi..
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ACCEPTABLE Wj OPERATION 50 4
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20 30 40 50 60 70 80 90 100 l
PERCENT OF RATED THERMAL POWER FIGURE 3.4-1 l
DOSE EQUIVALENT l-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific i
Qtivity > 1.0pCi/ gram Dose Equivalent 1131 BEAVER VALLEY - UNIT 1 3/4 4-21
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INSERT 1 DPR-66 t
7
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5 UNACCEPTABLE OPERATION i
5 b
- 200
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150 3
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1 E 100 E
5 h
ACCEPTABLE OPERATION \\
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8 Wb0 20 30 40 50 60 70 80 90 100 PERCENT OF RATED THERMAL POWER FIGURE 3.4-1 4
DOSE EQUIVALENT I-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific Activity > 0.35 pCl/ gram Dose Equivalent 1-131 I
BEAVER VALLEY - UNIT 1 3/4 4-21 Amendment No.
(Proposed Wording)
DPR-66 REACTOR COOLANT SYSTEM j
BASES
]
3/4.4.5 STEAM GENERATORS (Continued)
J correlation.and then the subsequent derivation of the voltage repair _ limit from the structural limit (which is then implemented by this surveillance).
The voltage structural limit is the voltage from the burst pressure / bobbin. voltage correlation, at the 95-percent prediction interval. curve reduced to account for the lower 95/95-percent tolerance bound for tubing. material properties at 650*F (i.e.,
the i
95-percent LTL curve).
The voltage structural limit must be adjusted downward to account for potential degradation growth during an operating interval and to account for NDE uncertainty.
The upper voltage repair limit; Vun,
is determined from the structural voltaga limit by applying the following equation:
Vug = VsL - Vor - VNDE where var represents the allowance for degradation growth between inspections and Vgog represents the allowance for potential sources of error in the measurement of the bobbin coil voltage.
Further discussion of the assumptions necessary to determine the voltage repair limit are discussed in GL 95-05.
'X M h-9 The mid-cycle equation in SR 4. 4. 5. 4.10.d -should only be used during unplanned inspections in which eddy current data is. acquired for indications at the tube support platss.
SR 4.4.5.5 implements several reporting requirements recommended by GL 95-05 for situations which the NRC wants to be notified prior to returning.the SGs to service.
For the purposes of this reporting requirement, leakage and conditional burst probability can be calculated based on the as-found voltage.
distribution rather than the projected end-of-cycle (EOC) voltage distribution (refer to GL 95-05 for more information) when it is not practical to complete these calculations using the projected i
EOC voltage distributions prior to returning the SGs to service.
Note that if leakage and conditional burst probability were calculated using the measured EOC voltage distribution for the purposes of addressing the GL section 6.a.1 and 6.h.3 reporting
- criteria, then the results.
of the projected EOC voltage distribution should be provided per the GL section 6.b (c) criteria.
Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be reported to the Commission pursuant to Specification 6.6 prior to resumption of plant operation.
Such cases will be considered by the Commission on a
case-by-case basis and may result in a requirement for
- analysis, laboratory examinations,
- tests, additional eddy-current inspection, and revision ' of the Technical Specifications, if necessary.
.v BEAVER VALLEY - UNIT 1 B 3/4 4-2b Amendment No.
@gsed ad4)
INSERT 2 Safety analyses were performed pursuant to Generic Letter 95-05 to determine the maximum MSLB-induced primary-to-secondary leak rate that could occur without offsite doses exceeding a small fraction of 10 CFR 100 (concurrent iodine spike), 10 CFR 100 (pre-accident iodine spike), and without control room doses exceeding GDC-19.
The current value of this allowable leak rate and a summary of the analyses are provided in Section 14.2.5 of the UFSAR.
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DPR-66 1
BASES 4
3/4.4.d.2 OPERATIONAL LEAKAGE (Continued)
I LCO (Continued) b.
s One gallon per minute (gpm) of unidentified LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air monitoring and containment sump level monitoring' equipment can detect within a reasonable time l
period.
Violation of this LCO could result in continued j
degradation of the RCPB, if the LEAKAGE is from the j
pressure boundary.
7 c.
Primarv-to-Secondary LEAKAGE throuah Any One SG l
aintaining 'an operating LEAKAGE limit of 150 gpd per steam generator will minimize the potential for a large l
i LEAKAGE event during a main steamline break.
Based on the non-destructive examination uncertainties, bobbin ggy coil voltage distribution, and crack growth rate from the i
previous inspection, the expected leak rate following a WN steamline rupture is limited to below 4.5 gpm in the ggg7 J faulted loop.
Maintaining LEAKAGE within the 4.5 gpm limit will ensure that postulated offsite doses will remain within the 10 CFR 100 requirements and that control. room habitability continues to meet GDC-19.{
f LEAKAGE in the intact loops will be limited to the operating limit of 150 gpd.
If the projected end-of-cycle distribution of crack indications results in l
primary-to-secondary LEAKAGE greater than 4.5 gpm in the l
j faulted loop during a postulated steamline break event, j
additional tubes must be removed from service or repaired in order to reduce the postulated steamline break LEAKAGE to below 4.5 gpm.
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Also, the 150 gallons per day leakage limit incorporated into.this specification is more restrictive than the j
standard operating leakage limit and is intended to provide an additional margin to accommodate a crack which j
might grow at a
greater than expected rate or unexpectedly extend outside the thickness of the tube support plate.
Hence, the reduced leakage limit, when 4
combined with an effective leak rate monitoring program, provides additional assurance that should a significant leak be experienced in service, it will be detected, and j
the plant shut down in a timely manner.
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BEAVER VALLEY - UNIT 1 B 3/4 4-3 Amendment No.'"
YMjkY I4l0t &)
r --
W
+--e F
"w P'
INSERT 3 c.
Primarv-to-Secondary Leakaae throuah Any One SG operating experience at PWR plants has shown that sudden increases in leak rate are often precursors to larger tube failures.
Maintaining an operating LEAKAGE limit of 150 gpd per steam generator will minimize the potential for a large LEAKAGE event at power.
This operating LEAKAGE limit is more restrictive than the operating LEAKAGE limit in standardized technical specifications.
This provides i
additional margin to accommodate a tube flaw which might grow at a greater than expected rate or unexpectedly extend 3
outside the thickness of the tube support plate.
This reduced LEAKAGE limit, in conjunction with a leak rate j
monitoring program, provides additional assurance that this precursor LEAKAGE will be detected and the plant shut down in a timely manner.
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DPR-66 i
REACTIVITY CONTROL SYSTEMS j
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BASES
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3/4.4.7 CHEMISTRY i
The limitations on Reactor Coolant System chemistry ensure 'that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress i
corrosion.
Maintaining the chemistry within the Steady State Limits l
provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant system over the life of the plant.
The associated effects of exceeding the oxygen, chloride and fluoride limits are time and temperature dependent.
Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant a
i effect on the structural integrity of the Reactor Coolant System. The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady State Limits.
i The surveillance requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient i
time to take corrective action.
4 3/4,4.8 SPECIFIC ACTIVITY i
he limitations on the specific activity of the primary coolant ensu that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will not exceed an appropriately small fraction of Part 100 limits following a steam generator tube rupture accident in conjunction with an assumed steady tate primary-to-secondary steam generator leakage rate of 1.0 GPM.
AWN Ml/&
f tr he primary coolant specific activity is limited in order to maintain offsite and control room ope doses associated with postulated accidents within applicable requirements. Specifically, the 0.35 pCi/gm DOES EQUlVALENT l-131 timit ensures that the offsite dose does not exceed a small fraction of 10 CFR Part 100 guidelines and that control room operator thyroid dose does not I
d GDC-19 in the event of primary-to-secondary leakage induced by a main steam line break.
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BEAVER VALLEY - UNIT 1 B 3/4 4-4 Amendment No.*::
YYY
^
( M 3 M ddC/M8M Ov9W/yJAs 04 M Mdid!fM REACTOR COOLANT SYSTEM f
BASES 3/4.4.8 SPECIFIC ACTIVITY Gbontinued) i O.
The ACTION p
itting POWER O
TION to continue for lim' me periods w' h the primary coolan s specific activity >
N uCi/ gramFigure [EQ DOSF VALENT I-131, but withi the allowable limit shown on 3.
-1, accommodates poss le iodine spiking phenomenon which ma occur following changes 1
THERMAL POWER.
Operation with spe fic activity levels exceeding uCi/ gram DOSE EQUIVALENT I-131 r
more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exc ing the limits shown on Figure 3.4-1 must be restricted
- i..;;
th;
- tivity levels all.ed by rigur; 2.4 1 5
- 55......__b...,5 5 y E5
..m m
5 5
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3 m...
m w.
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Reducing Tavg to
<500*F $
the release of activity should a steam generator tube rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves.j The surveillance requirements provide adequate assurance unan 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 operations.
The various categories of load cycles used for design purposes are provided in Section 4.1.4 of the FSAR.
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.
During
- heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner wall to tensile at the outer wall.
These thermal-induced compressive stresses tend to alleviate the tensile stresses induced by the internal pressure.
Therefore, a pressure-temperature curve based on steady stats conditions (i.e.,
no thermal stresses) represents a
lower bc ad of all similar curves for finite heatup rates when the inner wall of the vessel is treated as the governing location.
O s's acj> 5e 0 lS o M d etes $lg. f M sg d,at d, $ t e f f) g g y gg &
s k o m a k ta es & n,y A p a i,L/rh e J
- M.19,ea.
p
< s}s. W Ln.k acede# Aesa ya.aen -k-see.$a l
- v lee kan.
e BEAVER VALLEY - UNIT 1 B 3/4 4-5 Amendment No. 102 b)Deteg[ J44P#d
DPR-66 PLANT SYSTEMS BASES 3I4.7.1.4 ACTIVTTY k & f i k C E (Al / W X A I S & )t r y
[The limitations on secondary system specific activity ensure that
~
Ithe resultant off-site radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture.
This dose also includes the effects of a coincident 1.0 GPM primary to secondary tube leak in the steam generator of the affected steam line.
These values are consistent with the (assumptions used in the accident analyses.
3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blowdown in the event of a steam line rupture.
This restriction is required to 1) minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and 2) limit the pressure rise within containment in the event the steam line rupture occurs within containment.
The OPERABILITY of the main steam isolation valves within the closure times of the surveillance requirements are consistant with the assumptions used in the accident analyses.
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i BEAVER VALLEY - UNIT 1 B 3/4 7-3 Amendment N o. 44-Y lOOY
4
, INSERT 4 I
The limitations on secondary system specific activity ensure that steam releases to the environment will not be significant contributors to radioactivity releases resulting from analyzed accidents.
Many of the analyzed accidents assume that a loss of auxiliary AC power occurs, making the main condenser unavailable for plant
- cooldown, and making it necessary to dump steam to the environment via SG atmospheric dump valves.
Maintaining secondary system specific activity within the limits ensures that these
- releases, in conjunction with other releases associated with the accident, will be within applicable dose criteria.
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i ATTACHMENT B Beaver Valley Power Station, Unit No. 1 Proposed Technical Specification Change No. 240 RCS SPECIFIC ACTIVITY i
A.
DESCRIPTION OF AMENDMENT REQUEST The proposed amendment would modify the technical specifications to reduce the RCS specific activity limits.
The definition of DOSE EQUIVALENT I-131 would be replaced with the ISTS definition wording in the first sentence and an equation added based on dose conversion factors derived from International Commission on Radiation Protection (ICRP)
ICRP-30.
Specification 3.4.6, specific
- Activity, has been revised by reducing the DOSE EQUIVALENT I-131 limit from 1.0 pCi/ gram to 0.35 p.Ci/ gram.
Item 4.a) in Table 4.4-12, Primary Coolant Specific Activity i
Sample and Analysis Program, Figure 3.4-1 and Bases 3/4.4.8 have been modified to reflect the reduced DOSE EQUIVALENT I-131 limit.
B.
BACKGROUND Technical Specification Amendment No.
198 (April 1,
1996) implemented the voltage-based steam generator (SG) repair criteria for the tube support plate elevations in accordance with generic letter (GL) 95-05,
" Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes by Outside Diameter Stress l
Corrosion Cracking."
The GL provides for reduced RCS specific activity limits as an option to ensure that the total leak rate from the affected SG during a main steamline break (MSLB) would be less than a rate that could lead to radiological releases in excess of that licensed for the plant.
Amendment No. 198 did not use this option since a new dose calculation was required.
The new dose calculation has since been performed (Attachment C) and provides the basis for the reduced RCS specific activity limits.
UFSAR changes to reflect the analyses results will be incorporated in the update following approval of this amendment.
C.
JUSTIFICATION The changes made to reflect the reduced RCS specific activity limits have been justified based on the analyses provided in Attachment C.
The first analysis; RG 1.145 Short-Term Accident X/Q Values for EAB and LPZ, Unit 1 and Unit 2, based on 1986-1995 Observations; documents updated Unit 1 and Unit 2 exclusion area boundary (EAB) and low population zone (LPZ) values.
The X/Q values have been updated based on the 1986 to 1995 data.
The Unit 2 information is not applicable for this proposed change but is provided since it is available and will be referenced in the future for a similar Unit 2 change.
The second analysis; U1 RCS and Steam Generator Isotopic Concentrations, Pre-incident Spike Concentrations, and Iodine Spike Appearance Rates Corresponding to 0.35 and 0.5 pCi/gm RCS Specific Activity; determines RCS and SG isotopic concentrations that correspond to technical B-1
5 ATTACHMENT B, continued Proposad Tschnical Spacification Change No. 240 Page 2 i
specification (TS) RCS specific activity limits of both 0.35 and
)
0.5 pCi/gm.
Even though both of these limits have been evaluated l
and found acceptable for inclusion in the TS, the 0.35 pCi/gm value was selected as the proposed TS limit.
The third analysis; Safety Analysis of the Common Control Room, EAB and LPZ Doses I
from a Main Steam.Line Break Outside of CNMT at U1 with Increased Primary-to-Secondary Leakage; documents an analysis of the i
postulated dose in the common control room following a main steamline break (MSLB) outside containment (CNMT) with the i
objective of determining the maximum allowable primary-to-secondary leakage in the faulted SG.
In GL 95-05 the NRC states that a reduction in the reactor coolant iodine activity is an i
acceptable means for accepting higher projected leakage rates and e',
meeting the applicable limits of Title 10 of the Code of acGeLal Regulations (CFR) Part 100 and general design criteria (GDC) 19 utilizing licensing basis assumptions.
D.
SAFETY ANALYSIS
}
The DOSE EQUIVALENT I-131 definition has been revised to reflect the intent of the ISTS where in the first sentence ( Ci/ gram) has been replaced with (microcuries/ gram) and the word "which" has i
been replaced with "that."
The rest of the new definition wording provides an equation showing how the DOSE EQUIVALENT I-131 is determined along with reference to dose conversion factors derived from ICRP-30.
This is consistent with the ISTS which provides for using various conversion factors including
)
those from ICRP-30.
The use of ICRP-30 is recognized as an accepted source of conversion factors; therefore, this change
]
will not reduce the safety of the plant.
The Specification 3.4.8 limiting condition for operation (LCO),
Action Statements, Table 4.4-12 and Figure 3.4-1 DOSE EQUIVALENT I-131 limit has been reduced from 1.0 pCi/ gram to 0.35 pCi/ gram.
This is as provided for in GL 95-05 as a means'of increasing the allowable MSLB-induced primary-to-secondary leakage in implementing the alternate SG tube repair criteria.
The-secondary side equilibrium activity is a function of the RCS activity; however, GL 95-05 does not address changing this limit.
Since the initial activity in the SGs is a relatively minor contributor to dose, the TS 3.7.1.4 secondary side specific activity limit will not be changed.
The justification for these changes is provided in Attachment C Analysis 1.
Attachment C Analysis 2 provides the new primary-to-secondary SG leakage limit in the faulted SG based on the reduced RCS specific activity.
.The new faulted SG leakage limit (11.75 gpm) will be described in the UFSAR and will be used in the SG inspection program to comply with the SG tube repair criteria.
Based on these analyses, the control room and offsite doses have been analyzed and shown to comply with the regulatory limits; therefore, these changes do not reduce the margin of safety of the plant.
B-2
ATTACHMENT B, continutd-
+
Proposed Technical Specification Change No. 240 3-Page 3 i
The Specific Activity Bases for Specification 3.4.8 has been modified to reflect the more restrictive limits and new analyses.
In
- addition, the Bases for other specifications have been modified so that the wording is consistent with the Specific Activity.LCO and the UFSAR changes.
The SG Bases for J
Specification 3.4.5 has been modified by incorporating an additional paragraph discussing the dose calculation and providing reference to the applicable UFSAR.section.
The Operational Leakage Bases for Specification 3.4.6.2 has been revised by replacing the primary-to-secondary leakage discussion with a new paragraph which provides clarification of the 150 gpd l
SG leakage limit.
The Activity Bases for the Specification
]
3.7.1.4 SG secondary side activity limits has been replaced with j
a new paragraph to provide consistency with the Specification 3.4.8 discussion.
These Bases changes are consistent with the 1
analysis providing a basis for the changes to specification 3.4.8 and the plant design; therefore, they have been determined to be j
. safe and do not reduce the safety of the plant.
I E.
NO SIGNIFICANT HAZARDS EVALUATION i
l The no significant hazard considerations involved with the proposed amendment have been evaluated, focusing on the three j
standards set forth in 10 CFR 50.92(c) as quoted below:
The Commission may make a final determination, pursuant to i;
the procedures in paragraph 50.91, that a proposed amendment to an operating license for a
facility licensed under paragraph 50.21(b) or paragraph 50.22 or for a testing facility involves 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; or (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 following evaluation is provided for the no significant hazards consideration standards.
1.
Does the change involve a
significant increase in the probability or consequences of an accident previously evaluated?
The proposed change reduces the reactor coolant system (RCS) specific activity limits of Specification 3.4.8 from 1.0 pCi/ gram to 0.35 pCi/ gram and lowers the graph in Figure 3.4-1 by 39 pCi/ gram following the guidance provided in Generic Letter (GL) 95-05.
This reduces the RCS activity B-3
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-ATTACHMENT B, continund Proposed Technical Specification Change No. 240 Page 4 i
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allowed to leak to the secondary side when t':e plant is operating so that additional margin is available to support a higher allowable accident-induced leakage.value as justified by analysis.
]
The.
proposed changes to Specification 3.4.8 and the definition of DOSE EQUIVALENT I-131 ensure these requirements are consistent with the latest analyses.
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These changes implement the more restrictive RCS activity l
limits in accordance with applicable analyses and GL 95-05 to j
ensure the regulations are satisfied.
Therefore, these 1
changes do not involve a
significant increase in the i
probability or consequences of an accident previously l
evaluated.
i 2.
Does the change create the possibility of a new or different i
kind of accident from any accident previously evaluated?
1;-
The proposed change does not alter the configuration of the l
plant or affect the operation with the reduced specific activity limit.
By reducing the specific activity limit, the limit would be reached sooner to initiate evaluation of the out of limit condition.
The proposed changes will not result 1
in any additional challenges to the main steam system or the j
reactor coolant system pressure boundary.
Consequently, no i
new failure modes are introduced as a result of the proposed j
changes.
As a result, the main steam line break, steam generator tube rupture and loss of coolant accident analyses i
i remain bounding.
Therefore, the proposed change will not create the possibility of a new or different kind of accident i
from any accident previously evaluated.
2 3.
Does the change involve a significant reduction in a margin j
of safety?
1 l
The proposed change reduces the RCS specific activity limit to 0.35 pCi/ gram along with lowering the Figure 3.4-1 limits i
by 39 pCi/ gram.
Reduction of the RCS specific activity i
limits allows an increase in the limit for the projected SG leakage following SG tube inspection and repair in accordance 4
with the voltage-based SG tube alternate repair criteria (ARC) incorporated by Amendment Nc. 198, This follows the guidance provided in GL 95-05 and effectively takes margin available in the specific activity i.imits and applies it to 1
the projected SG leakage for the ARC.
This has been determined to be an acceptable means for accepting higher projected leakage rates while still meeting the applicable I
limits of 10 CFR 100 and GDC 19 with respect to offsite and f
control room doses.
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4 ATTACHMENT B, continuad Proposed Technical Specification Change No. 240 Page 5 t
The capability for monitoring the specific activity and complying with the required actions remains unchanged.
In addition, there is no resultant change in dose consequences.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
F.
NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION Based on the considerations expressed above, it is concluded that the activities associated with this license amendment request satisfy the no significant hazards consideration standards of 10 4
CFR 50.92(c)
- and, accordingly, a
no significant hazards consideration finding is justified.
G.
UFSAR CHANGES UFSAR changes to reflect the analyses results will be incorporated in the update following approval of this amendment.
An information copy of the proposed UFSAR change is provided in Attachment D.
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j B-5 4
j ATTACHMENT C Beaver Valley Power Station, Unit No. 1 Proposed Technical Specification Change No. 240 SUPPORTING ACCIDENT ANALYSES i
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RG 1.145 Short-Term Accident X/Q Values for EAB and LPZ, j
Unit 1 and Unit 2, Based on 1986-1995 Observations j
U1 RCS and Steam Generator Isotopic Concentrations, Pre-incident Spike Concentrations, and Iodine Spike Appearance Rates Corresponding j
to 0.35 and 0.5 pCi/ gram RCS Specific Activity i
j Safety Analysis of the Common Control Room, EAB and LPZ Doses from a Main Steam Line Break Outside of CNMT at U1 with Increased Primary-to-Secondary Leakage j
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4 f
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