ML20212C888
| ML20212C888 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 10/22/1997 |
| From: | DUQUESNE LIGHT CO. |
| To: | |
| Shared Package | |
| ML20212C877 | List:
|
| References | |
| GL-95-05, GL-95-5, NUDOCS 9710300208 | |
| Download: ML20212C888 (23) | |
Text
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ATTACIDiENT A l
l Beaver Valley Power Station, Unit No. 2 l
Proposed Technical Specification Change No. 115 I
The following is a list of the affected pages:
l Affected Pages:
1-4 3/4 4-27 3/4 4-28 3/4 4-29 B 3/4 4-3b B 3/4 4-4g B 3//; 4-4h B 3/4 4-5 B 3/4 4-6 B 3/4 7-2 i
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9710300208 971022 i
PDR ADOCK 05000412 P
PDR 1
l NPF-73 DEFINITIONS l
DOSE EQUIVALENT I-131 IM E M E d'f 8 4 7 A 1.19 DOSE EQUIVALM'fT I-131 shall be that concentration of I
(#Ci/ gram) which alone would produce the same thyroid dose as the l
quantity and isotopic mixture 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 or ID 14844.
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 cosponent at the beginning of each subinterval.
FREOUENCY NOTATION 1.21 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspcnd to the intervals defined in Table 1.2.
HEAGTOR TRIP SYSTEM RESPONSE TIME 1.22 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until loss of stationary gripper coil voltege.
1 1
l EEINEERED SAFETY FEATURE RESPONSE TIME 1.23 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF squipment is capable of performing its safety funct!an (i.e.,
the valves travel to their required positions, pump discharge pressurec reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.
AXIAL FLUX DIFFERENCE 1.24 AXIAL FLUX DIFFEPF4CE shall be the difference in normalized flux signals between the t-and bottom halves of a two-section excore neutron detector.
BEAVER VALLEY - UNIT 2 1-4 Amendment No *-
YN/03YlOroYo 0
INSERT A QQE.. EQUIVALENT I-131-1.19 DOSE - EQUIVALENT I-131 shall be that concentration cf I-131 (microcuries/ gram) that alone would produce the same thyroid dose as
- the quantity and isotopic rixture 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:
Cg_g33
= C g.3 3 3 +
- +
- +
- +
D8'-
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.
1 dB
1 NPF REACTOR COOLANT SYSTEM 3/4.4.8 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION
.3.4.8 The specific activity of the reactor coolant shall be limited to:
a.
I pC1/ gram DOSE EQUIVALENT I-131, and l
'b.
i 100 /E pCi/ gram APPLICABILITY:
MODES 1, 2, 3, 4, and S.
ACTION:
MODES 1, 2 and 3*:
0.'5f 4.
With the specific accivity of the primary coolant > ++pCi/ gram i
DOSE EQUIVALENT I-131 for more than 48 hotirs csuring one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in HOT STANDBY with T,yg < 500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b.
With the specific activity of the primary coolant > 100 /E pCi/ gram, be in HOT STANOBY with T,yg < 500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
-MODES 1, 2, 3, 4, and 5 rimary coolant >
pCi/ gram l
With the specific activity of the hCi/ gram, perform the sampling a.
OOSE EQUIVALENT I-131 or > 100 /E analysis requirement of item 4a of Table 4.4 U until the specific activity of the primary coolant is restored to within its limits.
SURVEILLANCE REOUIREMENTS 4.4.8 The specific activity of the primary coolant shall be determined to be within the performance limits of the sampling and analysis progras of Table 4 4-12.
- With T,yg > 500*F.
BEAVER VALLEY - UNIT 2
( %
r/4 4-27alWedy) 3
NPF-73 TABLE 4.4-12 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALY5IS PROGRAM TYPE OF MEASUREMENT MINIMUM MODES IN WHICH AND ANALYSIS FREQUENCY SURVEILLANCE REQUIRED 1.
Gross Activity 3 times per 7 days 1, 2, 3, 4 Determination with a maximum time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> between samples.
2.
Isotopic Analysis for 1 per 14 days 1,
DOSE EQUIVALENT I-131 Concentrat.on 3.
Radiochemical for E 1 per 6 months 1,
Determination 4
Isotopic Analysis for 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#, 5#
Iodine including I-131 whenever the I-133, and I-135 specific ac i 1/ygram O.Tf l
exceeds DOSE EQUIVALENT I-131 or 100 /E pCi/ gram, and b) One sample betweer, 1, 2, 3 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> follow-ing a THERMAL POWER change exceeding 15 percent of the RATED THERMAL POWER within a 1-hour period.
9
- Until the specific activity of the p.imary coolant system is restored to i
within its limits..
(Frepes/4 4-28e/ Weehy, )
3 BEAVER VALLEY - UNIT 2
NPF k PLAC6 wtrH rysMV )
i E
\\
-[
a
- \\
b 250 b
i g
UNACCEPTABLE OPERATION
\\
b
\\
< 200 N
t O
$e 150 O
b
{
100 E
I m
e-ACCEPTABLE OPERATION i
8 50
=s I
8
\\u 8
20 30 40 50 60 70 80 90 100 PERCENT OF RATED THERMAL POWER FIGURE 3.4-1 DOSE EQUIVALENT l-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific Activity > 1.0 ocl/ gram unna Eaulvalent 1-131 BEAVER VALLEY - UNIT 2 @rops/4 4-29tdNH0s) 3
INSERT 1 NPF-73 EE
\\
S250
-t-2U UNACCEPTABLE OPERATION b200
\\
U E
v>
150 3
a 100 e
s ACCEPTABLE OPERATION 50 b5 8
=
b0 20 30 40 50 60 70 80 90 100 PERCENT OF RATED THERMAL POWER FIGURE 3.4-1 DOSE EQUIVALENT l-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 l
BEAVER VALLEY - UNIT 2 3/4 4-29 Amendment No.
(Proposed V!ording)
-REACTOR COOLANT SYSTEM BASES 1
') / 4. 4. 5 STEAM GENERATORS (Continued)
~ where Vor represents the allowance' for degradation growth between inspections and Vuog represents the allowance for potential sources of error in the measurement of the bobbin coil voltage.
Further discussion of the assumptions necessery to determine the voltage repair limit are discussed in GL 95-05.
$WW The mid-cycle equation in SR 4.4.5.4.a.10.d should only be used during unplanned inspections in which eddy current data is acquired for indications at the tube support plates.
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 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.a.3 reporting criteria, then the results of the projected EOC voltage distribution chould be provided per the GL section 6.b (c) criteria.
Whenever the results of any steam generator tubing inservice inspection f all into Category C-3, these results will be reported to the Comnission 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.
BEAVER VALLEY - UNIT 2 B 3/4 4-3b Amendment No.
(Proposed Wording)
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 the maximum. MSLB-induced leak rate and a summary of the analyses are provided in Section 15~.1.5 of the UFSAR.
'T
NPF-73 REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)
LCO (Continued) through a
component which can be isolated from the balance of the Reactor Coolant System, plant operation may continue provided the leaking componem. is promptly isolated from the Reactor Coolant System c.;rca isolation removes the source of potential failure.
b.
Unidentified LEAKAGE 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 period.
Violation of this LCO could result in continued degradation of the RCPB, if the LEAKAGE is from the pressure boundary, c.
Primary-to-Secondary LEAKAGE through Any One SG f
crpd M f aintaining an operating LEAKAGE limit of 150 f steam generator will minimize the potential for a large LEAKAGE event during a main steamline break.
Based on M48Cf the non-destructive examination uncertainties, bobbia gg coil voltage distribution, and crack growth " ate from the previous inspection, the expected leak rate following a J#$fdT 3 steamline rupture is limited to less' than or equal to 2.95 gpm in the faulted loop.
Maintaining LEAKAGE less than or equal to the 2.95 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.
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 primary-to-secondary LEAKAGE greater than 2.95 gpm in the faulted loop during a postulated sr.eamline break event, additional tubes must be renoved from service or repaired in order to reduce the postulated steamline break LEAKAGE to less than or equal to 2.95 gpm.
BEAVER VALLEY - UNIT 2 B 3/4 4-4g Amendment No.
(Proposed Wording)
I.
NPF-73' L
BASES
- 3 3 / 4. 4.~ 6. 2 '- OPERATIONAL LEAKAGE-(Continuagu LCO (Continued)
P e
[Also, the'150-gallons per day leakage limit incorporated
_ N M 8 T-into this specification is more restrictive than' the i
gg
~i standard operating; leakage limit and is intended to provide an additional margin to accommodate'a crack which 3"#$btfJ 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 combined with an' effective leak rate monitoring program, provides additional - assurance- - that should a-significant -
leak be experienced in service, it will be detected, andJ the plant shut down in a timely manner.
d.
Identified LEAKAGE Up to 10 gpm of identified LEAKAGE is considered allowable because LEAKAGE is from known sources that do r
not interfere with detection of identified LEAKAGE and is well within the capability of ' the RCS Makeup System.
Identified LEAKAGE - includes LEAKAGE.to the containment from specifically known and located sources, but does not include pressure boundary LEAKAGE or controlled reactor coolant. pump (RCP) seal leakoff (a normal function not considered LEAKAGE).
Vio1Lation of this LCO 'could result in continued degradation of a component or system.
APPLICABILITY In MODES 1,
2, 3,
and 4,
the potential for RCPB LEAKAGE is greatest when the RCS is pressurized.
L In MODES 5 and 6, LEAKAGE limits are not required because the reactor coolant -pressure is far lower,. resulting in lower stresses
. and reduced' potentials for LEAKAGE.
LCO 3.4.6.2, "RCS Pressure Isolation Valve (PIV), " measures leakage through each individual PIV-andx can impact this LCo.
of-the - two PIVs in: series in each isolated line, leakage measured ^
through one'PIV does not result in RCS LEAKAGE when the other is l leak tight.
If both valves-leak and result in a loss of mass from
~
the RCS, the. loss must ~ be. : included in : the allowable identified m LEAKAGE.
(
-f
' BEAVER VALLEY - UNIT 2-B 3/4 4-4h Amendment No.
(Proposed Wording) s
-=, -,
~ ~,
..c-#,---.
INSERT 3 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 additional margin to accommodate a tube flaw which. might grow at-a greater than expected rate or unexpectedly extend outside the thickness of the tube support plate.
This reduced LEAKAGE limit, in conjunction with a leak rate monitoring program, provides additional assurance that this precursor LEAKAGE will be detected and the plant shut down in a timely manner.
- ~...
~. - - -
n NPF-73 REACTOR COOLANT' SYSTEM BASES 3/4.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that I
corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure dun to stress corrosion. Maintaining the chemistry within the Steady State' Limits provides adequate corrosion protection to ensure the structural integrity of the. Reactor Coolant System ovey 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 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 thra steady State Limits.
The surveillance requirements provide adequate assurance that concentrations in. excess of the limits will be detected in sufficient time to take corrective action.
I 3/4.4.8 SPECIFIC ACTIVITY pggy upp7Af ggMf 4,g The limitations on the specific activity of the primary coolant naure 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 1
exceed an appropriately small fraction of 10 CFR Part 100 limits following a steam generator tube rupture accident in conjunction with q rate of 1.0 GPM.anassumedsteadystateprimary-to-secondarysteamgeneratorleaka The ACTION statement permitting POWER OPERATION to conti*nue for limited time periods with the primary coolant's specific activity >
pCi/ gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accommodates possibla iodine spiking phenomenon which may' occur following changes in THERMAL POWER. Operation with i
specific activity levels exceeding
- Ci/ gram DOSE EQUIVALENT I-131 for more than_48 hours-during one ontinuous time interval
.2.Cr
' BEAVER VALLEY - UNIT 2 B 3/4.4-5 Amendment No.
- i W epsH ule 4 Q
>w,'
m+nn,
.e,.
m
.er-n,-
v.
INSERT 4 The primary coolant. specific activity--is limited in order to maintain offsite: and control room operator doses associated-with postulated accidents-within applicable requirements.
Specifically, the
- 0.35 pCi/gm DOSE EQUIVALENT I-131 limit ensures that the offsite-dose-
- doss o not exceed ' a small fraction of 10 CFR Part 100 guidelines and that control room operator thyroid dose does not exceed'GDC-19 in the I
event of primary-to-secondary ~ leakage induced by a main steam line break.
j i
1 d
Tho's settim also Neluees tha fnsssas sloWerentitl Iter-73
- ,9** h*MM N N84 *Mm /Ime,
,e, o
.p oe.,
b' REACTOR COOLANT SYSTEMOriekeco 4 on,olue.e} paary-re.swary 69 BASES
& CA tutt. N essu mfAos.t AMh t*O 3/4.4.8 SPECIFIC ACTIVITY (Continued) k #SKeecMe/ada/yses met pr[j u$ ra tghof.
or excsading the limits shown on Figure 3.4-1 :,ust ne restricted d x f tM x-
!b!
535 b b 5 N b !555 $ k N'5. ~5535.v - m$ b95 N 5f b 55 9 5 bb; N
- -.. r i m n v i
=____ 7 -, -i - - -. -
y
~.
..r.....
Reducing T,yg to<500*FI.U-Y/thereleaseofa:tivityshouldasteam generator tube rupture since the saturation pressure offthe primas coolant below the lift pressure of the atmospheric steam relief valves.#The surveil-lance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take cor-rective 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 3.9 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 will produce thermal stresses which vary frem 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 state conditiors (i.e., no thermal stresses) represents a lower bound of all similar curves for finii.e heatup rates when the inner wall of the vessel is treated as the governing location.
The heatup analysis also covers the determination of pressure-temperature licitations for the case in which the outer wall of the vessel becomes the con-trolling location.
The thermal gradients established during haatup produce ten-sile stresses at the outer wall of the vessel.
These stresses are additive to the pressure induced tensile stresses which are already present.
The thermal induced stresses at the outer wall of the vessel are tensile and are dependent on both the rate of heatup and the time along the heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannt be defined.
Subsequently, for the cases in which the outer wall of the vessel l
becomes the stress controlling location, each heatup rate of interest must be l
analyzed on an individual basis.
I-BEAVER VALLEY - UNIT 2 B 3/4 4-6 (Propsed Werd, )
NPF-73 3/4.7 PLANT SYSTEMS j
i BASES 3/4.7.1.1 SAFETY VALVES (Continued)
X Total relieving capacity of a stfety valves per steam _line
=
in 1bs/ hour (4,242,375)
Y
=
Maximum relieving capacity of one safety valve in 1bs/ hour (848,475)
/.
3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The OPERABILITY of the Auxiliary Feedwater System ensures that the Reactor Coolant System can be cooled down to less than 350*F from normal operating conditions in the event of a total loss of offsitc power.
Each electric driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 350 gpm.at a pressure of 1133 psig to the entrance of the steam generators.
The steam driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 700 gpm at a pressura of 1133 psig to the entrance of the steam generators.
This capacity is sufficient to ensure that adequate feedwater '?ow is available to remove decay heat and reduce the Reactor Coolant System temperature to less tnan 350*F when the Residual Heat Removal System may be placed into operation.
3/4.7.1.3 PRIMARY PLANT DEMINERALIZED WATER (PPDW)
The OPERABILITY of the PPDW storage tank with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> with steam discharge to atmosphere.
3/4.7.1.4 ACTIVITY ggp g g falf Sd7".f The limitations on secondary system specific activity ensure that the resultant offsita radiation dose will be limited to t small fraction of 10 CFR Part 100 limits in the event of a steam line rupture.
This dose aisc includes the effects of a coincident 0.35 gpm prin.ary-to-secondary tube leak in the steam l
generator of the affected steam line.
These values are consistent with the imptions used in the accident analyses.
BEAVER VALLEY - UNIT 2 B 3/4 7-2 0Myt*St$ 0
INSERT 5 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.
5 L
i
ATTACHMENT B Beaver Valley Power Station, Unit No. 2 Proposed Technical Specification Change No. 115 RCS SPECIFIC ACTIVITY r
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.
Spacification 3.4.8, Specific
- Activity, has been revised by reducing the DOSE EQUIVALENT I-131 limit from 1.0 pCi/ gram to 0.35 pCi/ gram.
Item 4.a) in Table 4.4-12, Primary Coolant Specific Activity Sample and Analysis Program, Figure 3.4-1 and Bases 3/4.4.8 have been modif'ied to reflect the reduced DOSE EQUIVALENT I-131 limit.
B.
BACKGROUND Proposed Technical Specification Change No. 109 was submitted to the NRC by letter dated June 18, 1996, to implement the voltage-based steam generator (SG) repair criteria for th6 tube support plate elevations in accordance with generic letter (GL) 95-05,
" Voltage-Based Repair Criteria for West!nghouse Steam Genera *,.or Tubes by. Outside Diameter Stress Corrosion Cracking."
The GL provides for reduced RCS specific activity limits as an cption to ensure that the total leak rate from the affected SG during a main stent.l!.ne break (MSLB) would be less than a rate that could lead to radiological releases in excess of that licensed for the plant.
Proposed Change No. 109 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 several analyses, with one of those analyses previously provided in Unit 1 proposed Technical Specification Change No.
- 240, submitted to the NRC by letter dated March 10, 1997.
That analysis; RG 1.145 Short-Term Accident 4/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.
A second analysis is provided in Attachment C; U2 RCS and Staam Generator Isotopic Concentrations, Pre-incident Spike j
B-1 j
1
ATTACHMENT-B,.continu d.
4 Proposed Technical Specification Change =No.
115-Page-2 4
. Concentrations, and' Iodine ' Spike : Appearance ' Rates Corresponding 1*:o'O.35_-and 0.5 pCi/gm RCS Specific Activity;; determines:RCSLand SG-isotopic
. concentrations that correspond:
to
- technical-specification (TS)1 RCS~ specific activity limits of both 0.35.- and.
~0.5 pCi/ga.-
Even though both of these limits have;been evaluated and found acceptable for inclusion in the TS, the.0.35 pCi/gm value was salseted as tha propcsed TS limit.
A' third --analysis is also provided in Attachment C;
Safety Analysis of the EAB, LPZ_and Common Control Room Doses from a-1
. Main Steam Line Break outside ' of -. CNMT - at' U2 with Increased Primary-to-Secondary Leakage '(SG APC); documents an analysis of
'the' postulated dose in the common control room following a_ main steamline. break (MSLB) outside containment (CNMT) ~ with.the 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 acceptable means for accepting higher projected laakage rates and still meeting the applicable limits of Title 10 of the Code - of Federal Regulations (CFR) Part 100 and general design criteria (GDC) 19 utilizing licensing basis assumptions.
i D.
SAFETY-ANALYSIS l.
The-DOSE EQUIVALENT I-131 definition has been revised to reflect the intent of the ISTS where in the first sentence (pci/graa):has
' been replaced with - (microcuries/ gram) and the word awhich" has been replaced with "that."
The. rest of the new definition wtrding 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 i
o secondary side equilibrium activity is a function of the RCS activity;-however, GL 95-05 does-not address changing this limit, i
Since: the initial activity. in the SGs is a relatively minor L 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 Analysis 1--that was provided with Unit ~1
. Technical Specification Change No. 240.
L 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 - (9.6 gpa) will1be B-2 i.
A-,
...,. ~,
2, A
~,
_____-_,...'_--_..u_
ATTACHMENT B,a continued'
-Proposed Technical ~ Specification Change No.-115 lPageL3:
i describad in. thei UFSAR H and ;will - be-used sin the SG inspection-
? program ; to conply-with the ; SG tube repair = criteria._ _ ' Based. on i
these: analyses, _ the control roon and _ offsite doses - have been -
i limits;-
analyzed-and >shown to comply: with the regulatory therefore,l these changes do not _ reduce the margin of safety of the plant.
-The-Specific - Activity Bases for Specification 3'. 4. 8 h a v e been
' modified to retlect the.more restrictiva limits and new analyses.
Ino addition,. the-Bases for other specifications _have-been i
- modified -so. that the wording. is consistent-with the Specific i
The -SG Baces for Activity LCo and.the UFSAR changes.
~
incorporating: an Specification 3.4.5 'have been modified-by additional paragraph discussing. the dose calculation-and provid.ing -reference to the applicable. UFSAR' section.
The
' operational Leakage Bases for specification - 3. 4. 6. 2 have been revised by replacing the primary-to-secondary leakage discussion with a new paragraph'which provides clarification of the-150 gpd 4
SG ;1aakage limit.
The Activity Bases for the Specification 3.7.1.4 SG' secondary side activity limits have been raplaced with a.. n e w ' p a r a g r a p h to provide consistency with the Specificatlon 3.4.8 discussion.
These Bases changes are consistent with the analysis providing a basis for the changes to Specification 3.4.8 i
and the_ plant design; therefo'Ja, they have been determined to be safe and do not reduce the safety of the plant.
E.
NO SIGNIFICANT HAZARDS-EVALUATION The no significant hazard considerations involved with the proposed amendment have been evaluated, focusing on the three standards set forth in 10 CFR 50.92(c) as quoted below:
The Commission may make a final determination, pursuant to
-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-f acility - involves no significant hazards consideration, if o p e r a t i o n -' o f the facility in accordance with the proposed amendment would nott (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.
E The.following evaluation is provided for the no significant
-hazards consideration standards.
I,'
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IATTACHMENT1B, continu d-
- ProposedLTechnical. Specification Change No. 115:
Page 4 4
1.;
fDoes--the-change. involve a
significant increase 'in the-probability; or consequences l of-lan-accident previously-avaluated?
The proposed change reduces;the reactor coolant-system-(RCS)-
1 specific activity _ l i m i t s o f '. Specification 3.4.8-from 1.0 pci/ gram to 0.35 pCi/ gram and lowers the graph in Figure
. 3.4 by 39 - pCi/ gram following-the guidance provided 'ln Generic Letter - (GL) 95-05.-
This - reduces the RCS activity l
allowed to -leak to the secondary side when the - plant is operating so that additional margin is-available to support a i
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.
These changes -implement the more restrictive RCS activity limits in accordance with applicable analyses and GL 95-05 to
- ensure the regulations are satistied.
Therefore, these changes do - not-involve a
significant increase in the probability-or consequences of an accident previously evaluated.
2.
Does the. change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change does not alter the configuration of the plant or affect the operation with the reduced specific activity limit.
By reducing the specific act2.vity limit, the limit would be reached sooner to -initiate evaluation of the
.out of limit condition.
The proposed changes will not result
- in any additional challenges to the main steam system or the reactor' coolant system pressure boundary.
Consequently, no
-new failure modes-are-introduced.as a result of the proposed changes.
As a result, the main steam line break, steam generator tube rupture and loss of coolant accident ~ analyses remain bounding.
Therefore, the proposed change - will not e
create the possibility of a new or different kind of accident from any accident previously evaluated; o
- 3.
. Does the1 change involve a significant reduction -in a margin
- of safety?
The proposed change reduces the RCS specific activity limit to 0.35 pCifgram along with lowering the-Figure 3.4-1-ilimits by 39 pCi/ gram.'
Reduction of the - RCS specific activity limits allows an increase in the limit for the projected SG
!1eakage following SG tube inspection and repair in accordance with the - voltage-based.- SG tube alternate ' repair criteria (ARC)..
This follows the guidance provided in GL 95-05 and 4
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ATTACHMENT B, continu;d Proposed Technical Specification Change No. 115 i
Page 5 effectively-takes margin available in the specific activity-limits and applies it to the projected SG 1eakage for the ARC.
This has been determined to be an acceptable means for accepting higher projected leakage rates while still meeting the applicable limits of 10 CFR 100 and GDC 19 with respect to offsite and control room doses.
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 CFR 50.92(c)
- and, accordingly, a
no significant hazard.
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.
B-5
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ATTACHMENT C Beaver Valley Power Station, Unit No. 2 Proposed Technical Specification Change No. 115 SUPPORTING ACCIDENT ANALYSES U2 RCS and Steam Generator Isotopic Concentrations, Pre -incident Spike Concentrations, and Iodine Spike Appearance Rates Corresponding to 0.35 and 0.5 pCi/ gram RCS Specific Activity Safety analysic of the EAB, LPZ and Common Control Room Doses from a Main Steam Line Break Ou', side of CNMT at U2 with Increased Primary-to-Secondary Leakage (SG APC) i 4
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