ML040900237
| ML040900237 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee  |
| Issue date: | 03/29/2004 |
| From: | NRC/NRR/DLPM |
| To: | |
| References | |
| TAC MB8119, TAC MB8379 | |
| Download: ML040900237 (7) | |
Text
VYNPS 3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS OPERATION B. Coolant Chemistry B. Coolant Chemistrv
- 1. a. During reactor power 1 *a. A sample of reactor operation, the coolant shall be radioiodine taken at least every concentration in the 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> and reactor coolant analyzed for shall not exceed radioactive iodines 1.1 microcuries of of 1-131 through I-131 dose I-135 during power equivalent per gram operation. In of water, except as addition, when steam allowed in jet air ejector Specification monitors indicate an 3.6.B. .b. increase in radioactive gaseous effluents of 25 percent or 5000 pCi/sec, whichever is greater, during steady state reactor operation a reactor coolant sample shall be taken and analyzed for radioactive iodines.
Amendment No. 33, 94., G43, 218 116
VYNPS Figure 3.6.1 Reactor Vessel Pressure-Temperature Limitations Hydrostatic Pressure and Leak Tests, Core Not Critical 40F1hr HeatuplCooldown Limit Vaid Through 4.46E8 MWH(t) 11
....I ...i....I...
I
. I-./.i... 0. ... . ....
. . . .I-I i I --- Bottom Head Curve. ....... .... .
Use minimum of bottom ......... e...Re....s.
.... fluid temperature and ......... Use minimum of downcomer ..
1 100 hea suraceregion otto fluid temperature and
.. temperature. .. ... .. f....
lange region outside 1000...... .
........ ... jrete tan110F ue nl the downcomer region. fluid
..... temperature.
- 0. .... .........
4 . 4 .4............. 4
- 0. ..... ... . * .... b ....... .. .... b... ... .. . J.b.. . .. .
0 I-80 0
..800 . ... ... .. 80.... 65...253.
w ....................... .... 85 ... 712 . 253
.. I.....
.. ....... .0........ 7..4 .. 253......
- a. 100 885to 253e 11 03 64
...... 115...
8 1117 8853 200 ....... ........... .. . . . 125121 9532 130 76 1042 tOO --.... 135 88 1153 00145 e5 1253 200 60 80 100 120 140 180 180 200 TEMPERATURE (7)
Amendment No. 3, 462, .84~, .g-, 1-2O, 24", 21815 135
VYNPS FIGURE 3.6.2 Reactor Vessel Pressure-Temperature Llmitatlons Normal Operation, Core Not Critical Upper Reaions, 100'Fhr HeatuplCooldown Limit Use minimum of Valid Through 4.46E8 MWH(t) downcomer region fluid temperature and Rance It * . region outside surface
- I : : : I :: .1 ::I. I it" temperature, except when 12001. l - - - - - -
... :...: .' . .. :... ... :......... ..:.-. --...-... - : Rlange temperature Is
. -. greater than 140-F, use
-vr- .. ' i'... ...... ....... ........
- t. only the downcomer
......... . ....... ....... _.. . i....... ............. i......... _ region flu d temperature.
1100 tonom Head Curve:
Use minimum of bottom fluid
- .1...I
- ...I....
I.....
- ... ' ../.
. /.
f7tzz-jt-..nzX4 Botm UUU .: j: :
- : emperature and
... I...i... ... bottom head surface !
- Ji I.,--
Tenp Hwd IrIorni temperature. 1- <.... ... I.. .1.-....... . M (Eo rnald
., . ~~ I.I.l...... 83 0 0 S
Is oo 1 I j I' .sj 439 253
-...-- ^--;...-t-;--;-:r 85 474 253
... .... . . so 513 253 A0 n: :: : :* : 195 555 253
... .. si.~
----'-z'l-i--
, .............. _.... 63 253 IL-105 685 253 E 700 110 713 253 2 . . .. .......'.... .. . _ 115 m 253 120 848 253 0: 55 4 1-..- ...... .... 125 253 t 600 5j..4 4 ~ . . ... .......
1 Im 1013 273 1109 253
- ........... . . ?j ...................
^i ..... .... ,, . . 140 1214 253 t 500 t.. .w . -----.
f.fi-n.U O . .............. 140 1214 833
.. . .. .. -. ;.-... ... /...-............... ^..;.,.:.<.;....:..... . .....
145 1312 163
'400 155
,,j, ,, * ,,, r :.. . T....
............ 160 1103
.... ! _ . I.........
1..... .- ..... 165 1190
- l-.--
300 . . . . . . 1 125.
- --'1*-*i*-...........
.... ... ... ... ...... .. . .. I:.. ... .... L ..... ...
..-;.i-1--'
. . ........ During tensioning and detensboning operations with the vessel vented and the vessel 200 - - .luid level beiow the flange region, the flange temperature may be monitored with test
- - - -- instrumentation In lleu of process Instrumentation for the downcomer region fluid
.i--. itemperature and permanent flange region outside surface temperature. The test
.Ii-..4...... ..*-i-- instrumentation uncertainty must be less than 4)- 2-F. The flange region temperatures 100 - ust be maintained greater than or equal to 72 F when monitored with test i... 4.- .. j instrumentaton during tensioning. detensloning, and when tensioned.
60 80 100 120 140 160 180 200 TEMPERATURE (F)
Amendment No. -33, 9a, 2, 218 136
VYN PS FIGURE 3.6.3 Reactor Vessel Pressure-Temperature Limitations Normal Operation, Core Critical 100Flhr HeatuplCooldown Limit If Pressure < 253 psig, Water Level must be within Normal Range for Power Operation Valid Through 4.46EB MWH(:)
1200-4... . .. ;..,_ _:I .. . . I.
. Ii...:: I:_ ~... ... ... i... . : .
Temp HI-ead te2oglrS I...:...:....:....
(F) (psig) (pg) Bottom Head .. _......
Renton-E0 0 _
Use minimum of 1100- E0 114 114 253 253 402 bottom fluid temperature and bottom head
......,...y..r-:... T--..K...
r:... ...... I..-
- 1~14 . 1 .j T 2....
1000 - 115 407 _ surtace 116 120 413 139 253 0 temperature. .. ... .. -1j
.. .4-;t-;-----
j
. .,..,/....
J;.i-
. -' -';I'-..--_ ----- l-------. i--T- ____
125 474 253 130 C513 253 :- -- -- - - -:. -- , - - - 'r- ........:
- a. ...... ... : ...... :.. ....... ....... :.............
135 555 253 .......
T .. .................
A 140 1:03 253
= 800 i i - i i i.
- 0. 145 1:555 253 150 1:713 253 155 77 253 ........ .......... ..... . .. 5....b.
V 700*
1c0 B48 253 . .4 . . . . , . 4 ..
165 926 253 4T t I t -- ' I-@-1.2e- t 4-eb*- --w 170 013 253 600 175 180 108 214 253 253 4.~........
,, ... '.i...:.... .... :...:...:............:..... .:.:.:.. .-..
180 214 830 : : ~~~~~...:::...-. ..... -:--- _.... -
U a: 500 185 312 889 953 190 -
195 - 1024
=C, 400 200 - 1103 I . . . I*. . .
UJ 205 _ 1190 .-.
4 '1 ., I L ... .........-,Upper
- Remionu
- 0. .
210 - 1258 _ ,. . ... . .....
. . ..... . . ^.-Use n-Tnimum of
...- r .e....f........~t....... . : . ...... I I downcomer region
- w*S*.l : w:.
vuu * < nn.
fuid temperature i
4 ... i....
- i . ... 4.... .......I-i-i.g..
'. . . .L I.. ... E .... j.;. ....... i...*--,--
4...4.
- ... i-----
and flange region
. . j
. . . . . .I . :-" ' i...l ...i..i:i. outside surface t I:
. 1:::- temperature. except 2uo ^
nnn . .
. .j. .
l r-:- -:--:) ::: r . ... ..... when flange
... ;...;...._... >...:...:.--:--4 I.....:....... I .. :-- -- v-temperature Ls
... ...I...N3; .... ;...:...:....
...>.......4..;. I...: ...
- l. -! -.i. --- -
.F...I...I.j..-i.l 9111-.
i;4 .. I..-. greater than 180'F, use only the 100 I.. .. :... ..
I.. :.. . . ... ....... ........ downcomer region
... :...:.... . ... .... :.. .:...:.I. :...:...:.... ........ :.......
I....... .... ....
I :I
_. I...',...i:....'....
...... ose ... fluid temperature.
r ::.. ;-.l- .w::12
... I...I....
I. . ... I.. ... :..I...~wo I...seseo
. . .I : : : i
- i 'I I i i- ' ' ' +i .. X . - . . . I -
60.00 80.00 100.00 120.00 140.00 160.00 180.00 200.00 TEMPERATURE (7)
Amendment No. 3, 4a, 2-3, 218 137
VYNPS BASES:
3.6 and 4.6 REACTOR COOLANT SYSTEM A. Pressure and Temperature Limitations 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 4.2 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.
The Pressure/Temperature (P/T) curves included as Figures 3.6.1, 3.6.2, and 3.6.3 were developed using IOCFR50 Appendix G, 1995 ASME Code,Section XI, Appendix G (including the Summer 1996 Addenda), and ASME Code Case N-640. These three curves provide P/T limit requirements for Pressure Test, Core Not Critical, and Core Critical. The P/T curves are not derived from Design Basis Accident analysis. They are prescribed to avoid encountering pressure, temperature or temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the reactor pressure boundary, a condition that is unanalyzed.
During heating events, the thermal gradients in the reactor vessel wall produce thermal stresses that vary from compressive at the inner wall to tensile at the outer wall. During cooling events the thermal stresses vary from tensile at the inner wall to compressive at the outer wall. The thermally induced tensile stresses are additive to the pressure induced tensile stresses. In the flange region, bolt preload has a significant affect on stress in the flange and adjacent plates.
Therefore heating/cooling events and bolt preload are used in the determination of the pressure-temperature limitations for the vessel.
The guidance of Branch Technical Position - MTEB 5-2, material drop weight, and Charpy impact test results were used to determine a reference nil-ductility temperature (RTN=) for all pressure boundary components. For the plates and welds adjacent to the core, fast neutron (E > 1 Mev) irradiation will cause an increase in the RTnDT.
For these plates and welds an adjusted RTN= (ARTIDT) of 890 F and 73eF (1 and ~h thickness locations) was conservatively used in development of these curves for core region components. Based upon plate and weld chemistry, initial RTNw values, predicted peak fast neutron fluence (2.99 x 1017 n/cm2 at the reactor vessel inside surface) for a gross power generation of 4.46 x 108 MWH(t), these core region ARTND values conservatively bound the guidance of Regulatory Guide 1.99, Revision 2.
There were five regions of the reactor pressure vessel (RPV) that were evaluated in the development of the P/T Limit curves: (1) the reactor vessel beltline region, (2) the bottom head region, (3) the feedwater nozzle, (4) the recirculation inlet nozzle, and (5) the upper vessel flange region. These regions will bound all other regions in the vessel with respect to considerations for brittle fracture.
Two lines are shown on each P/T limit figure. The dashed line is the Bottom Head Curve. This is applicable to the bottom head area only and includes the bottom head knuckle plates and dollar plates. Based on bottom head fluid temperature and bottom head surface temperature, the reactor pressure shall be maintained below the dashed line at all times.
l.mpndlmpnt Nn. 3 G2. a., .94, a-24, 44-4, .3 , 218 138
VYNPS BASES: 3.6 and 4.6 (Cont'd)
Due to convection cooling, stratification, and cool CRD flow, the bottom head area is subject to lower temperatures than the balance of the pressure vessel. The RTv= of the lower head is lower than the ARTNDT used for the beltline. The lower head area is also not subject to the same high level of stress as the flange and feedwater nozzle regions. The dashed Bottom Head Curve is less restrictive than the enveloping curve used for the upper regions of the vessel and provides Operator's with a conservative, but less restrictive P/T limit for the cooler bottom head region.
The solid line is the Upper Region Curve. This line conservatively bounds all regions of the vessel including the most limiting beltline and flange areas. At temperatures below the 10CFR50 Appendix G minimum temperature requirement (vertical line) based on the downcomer temperature and flange temperature, the reactor pressure shall be maintained below the solid line. At temperatures in excess of the 10CFR50 Appendix G minimum temperature requirement, the allowable pressure based on the flange is much higher than the beltline limit.
Therefore, when the flange temperature exceeds the 10CFR50 Appendix G minimum temperature requirement, the reactor pressure shall be maintained below the solid line based on downcomer temperature.
The Pressure Test curve (3.6.1) is applicable for heatup/cooldown rates up to 40'F/hr. The Core Not Critical curve (3.6.2) and the Core Critical curve (3.6.3) are applicable for heatup/cooldown rates up to 100'F/hr. In addition to heatup and cooldown events, the more limiting anticipated operational occurrences (AOOs) were evaluated (Structural Integrity Report, SIR-00-155). For the feedwater nozzles, a sudden injection of 501F cold water into the nozzle was postulated in the development of all three curves. The bottom head region was independently evaluated for AQOs in addition to 40aF/hr and 100'F/hr heatup/cooldown rates. This evaluation demonstrated that P/T requirements of the bottom head would be maintained for transients that would bound rapid cooling as well as step increases in temperature.
The rapid cooling event would bound scrams and other upset condition (level B) cold water injection events. The bottom head was also evaluated for a series of step heatup transients. This would depict hot sweep transients typically associated with reinitiation of recirculation flow with stratified conditions in the lower plenum.
This demonstrated that there was significant margin to P/T limits with GE SIL 251 recommendations for reinitiating recirculation flow in stratified conditions.
Adjustments for temperature and pressure instrument uncertainty have been included in the P/T curves (Figures 3.6.1, 3.6.2 and 3.6.3). The minimum temperature requirements were all increased by 100 F to compensate for temperature loop uncertainty error. The maximum pressure values were all decreased by 30psi to account for pressure loop uncertainty error. In addition, the maximum pressure was reduced further to account for static elevation head assuming the level was at the top of the reactor and at 70'F.
Specification 3.6.A.3 requires that the temperature of the vessel head flange and the head be greater than 70'F before tensioning. The 70'F is an analytical limit and does not include instrumentation uncertainty, which must be procedurally included depending upon which temperature monitoring instrumentation is being used. The temperature values shown on Figures 3.6.1, 3.6.2 and 3.6.3 include a 10'F instrumentation uncertainty.
t"C"Aomamf No I-. 21R 139
VY1JPS BASES: 3.6 and 4.6 (Cont'd)
A Note is included in Figure 3.6.2 that specifies test instrumentation uncertainty must be +/- 20F and the flange region temperatures must be maintained greater than or equal to 720F when using such instrumentation in lieu of permanently installed instrumentation.
Qualified test instrumentation may only be used for the purpose of maintaining the temperature limit when the vessel is vented and the fluid level is below the flange region. If permanently installed instrumentation (with a 10F uncertainty) is used during head tensioning and detensioning operations, the 80F limit must be met.
In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures will be maintained within 500 F of each other prior to startup of an idle loop.
Vermont Yankee is a participant in the Boiling Water Reactor Vessel and Internals Project Integrated Surveillance Program (ISP) for monitoring changes in the fracture toughness properties of ferritic materials in the reactor pressure vessel (RPV) beltline region. (See UFSAR Section 4.2 for additional ISP details.) As ISP capsule test reports become available for RPV materials representative of VYNPS, the actual shift in the reference temperature for nil-ductility transition (RTI=) of the vessel material may be re-established. In accordance with Appendix H to IOCFR50, VY is required to review relevant test reports and make a determination of whether or not a change in Technical Specifications is required as a result of the surveillance data.
B. Coolant Chemistry A steady-state radioiodine concentration limit of 1.1 gCi of I-131 dose equivalent per gram of water in the Reactor Coolant System can be reached if the gross radioactivity in the gaseous effluents is near the limit, as set forth in the Offsite Dose Calculation Manual, or if there is a failure or prolonged shutdown of the cleanup demineralizer. In the event of a steam line rupture outside the drywell, the NRC staff calculations show the resultant radiological dose at the site boundary to be less than 30 Rem to the thyroid. This dose was calculated on the basis of the radioiodine concentration limit of 1.1 pCi of I-131 dose equivalent per gram of water, atmospheric diffusion from an equivalent elevated release of 10 meters at. the nearest site boundary (190 m) for a X/Q = 3.9 x 10-3 sec/m 3 (Pasquill D and 0.33 m/sec equivalent), and a steam line isolation valve closure time of five seconds with a steam/water mass release of 30,000 pounds.
The iodine spike limit of four (4) microcuries of I-131 dose equivalent per gram of water provides an iodine peak or spike limit for the reactor coolant concentration to assure that the radiological consequences of a postulated LOCA are within 10CFR Part 100 dose guidelines.
The reactor coolant sample will be used to assure that the limit of Specification 3.6.B.2 is not exceeded. The radioiodine concentration would not be expected to change rapidly during steady-state operation over a period of 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />. In addition, the trend of the radioactive gaseous effluents, which is continuously monitored, is a good indicator of the trend of the radioiodine concentration in the reactor coolant.
When a significant increase in radioactive gaseous effluents is indicated, as specified, an additional reactor coolant sample shall be taken and analyzed for radioactive iodine.
Amendment No. A-3, 64, 91, 04, -64, 193,0-a 4 , 218 140