ML20154Q382
| ML20154Q382 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 09/19/1988 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML20154Q335 | List: |
| References | |
| NUDOCS 8810030385 | |
| Download: ML20154Q382 (12) | |
Text
{{#Wiki_filter:- . - - , .. . _ - _ - . - . - -- _- - .._ . - - - t h l Attaciunent 1 Proposed Technical Specificatio>i Changes ! t I s f
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g TABLE 2.2-1 (Continued) , REACTOR TRIP SYSTEM INSTtRNENTATION 1 RIP SETPOINTS TOTAL SENSOR ALLOWANCE ERROR
$ (TA) Z_ (S) . TRIP SETPOINT ALLOWA8LE VALUE.
5 FUNCTIONAL UNIT % Steam Generator Water g 13. e- ~ m level Low-Low 14.2 1.5 >17% of span >15.3% of span from
- a. Unit i 17 Tree UK to 30K DK to aos RTP" RTP" increasire increasing linearly linearly to to >38.3% of span
> 40.0% of span from 30% to 1001 RTP" Trom 30% to 1001 RTP*
N TBD TBD 36.9% 35.1%
't
- b. Unit 2 4 e h >WK- of narrow 95. % of narrow l range span range span 8.57 0 1. 0 >77% of bus >76% (5016 volts)
- 14. Undervoltage - Reactor voltage (5082 Coolant Pumps volts) with a 0.7s response time 4.0 0 1.0 >56.4 Hz with a >55.9 Hz
- 15. Underfrequency - Reactor D.2s response time 4
Coolant Pumps
- 16. Turbine Trip N.A. N.A. M.A. >550 psig >500 psig
- a. Stop Valve Elf Pressure Low N.A. N.A. >1% open >1% open
- b. Turbine Stop Valve N.A.
Closure N.A. N.A. N.A. M.A.
- 17. Safety Injection In'put N.A.
from ESF
- RIP = RAi[D THERNAL POWER TBD = T0 BE DETERMIE D
TABLE 3.3-4 (Continued) . ks- ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUE NTATION TRIP SETPOINTS SENSOR h TOTAL ERROR l , (5) TRIP SETPOINT ALLOWA8tE VALUE i c FUNCTIONAL UNIT ALLOWANCE (TA) Z x Q 4. Steam Line Isolation us N.A. -
- a. Manual Initiation M.A. N.A. N.A. N.A.
g
- Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.
b.
" and Actuation Relays
- c. Containment Pressure-High-High 12.7 0.71 1.5 1 3 psig 1 3.2 psig
- d. Steam Line Pressure - Low 4.6 1.31 1.5 1 725 psig 3 694 psig a
- e. Steam Line Pressure- 8.0 0.5 0 $ 100 psi 1 122.8 psi **
Negative Rate - High R 5. Feedwater Isolation Automatic Actuation Logic N.A. N.A. N.A. M.A. N.A. b a. Actuation Relays 4 so
- b. Steam Generator Water Level-High-High (P-14)
- 1. Unit 1 5.4 2.18 1.5 -< 82.4% of < 84.2% of narrow narrow range range instrument instrament span TBD T1BD YBD %.1% 78.9%
- 2. 83 nit 2 h M fr- 1 M-REr of 1 M of narrow l narrow range range instrument instrument span span
- c. T,,g-Low 4.0(6.0 ) 1.12(0.71 ) 1.2(0.8 ) 1 %4*F 1 562*F(561*F )
1.0 0 0.5 11 inches 12 inches
- d. Doghouse Water Level-High above 577* above 577' floor level floor level
- e. Safety Injection See Ites 1. above for all Safety Injection Setpoints and Allowable Values.
TBD = TO BE DETDWIIED
TABLE 3.3-4 (Continued) y ENGINEERED SAFETY FEATURES ACTUATION SYSTEN INSTRUNENTATION TRIP SETPOINTS
- SENSOR TOTAL ERROR h ALLOWANCE (TA) Z, (5) TRIP SETPOINT ALLOWA8LE VALUE
, FUNCTIONAL UNIT .
I 38. ' Auxiliary Feedwater (Continued) 3 c. Steam Generator Water Level - Low-Low e- 1) Unit 1 17 14.2 1.5 3 17% of span 1 15.3% of m from 0% to span from 0% to 30% RTP 30% RTP increasing increasing linearly to linearly to 1 38.3% of span
> 40.0% of from 30% to 100%
span from 30% RTP to 1001 RTP TED TBD TED E SE 35.11 m 2) Unit 2 tP -t*-t -t-tr- 1-*ft of 1-6 -34 of narrow l
} narrow range range instrument 5 Pan span w
h d. Safety Injection See Item 1. above for all Safety Injection Setpoints and Allowable Values.
- e. Loss-of-Offsite Power M.A. N.A. M.A. 1 3500 V 1 3200 V
- f. Trip of All Main feedwater Pumps N.A. N.A. M.A. N.A. N.A.
- g. Auxiliary feedwater Suction Pressure-tow
- 1) CAPS 5220, 5221, 5222 M.A. N.A. M.A. 1 10.5 psig 3 9.5 psig
- 2) CAPS f230, 5231, 5232 4.A. N.A. M.A. 3 6.2 psig 3 5.2 psig
- a. Unit 1 N.A. N.A. N.A. t 6.2 psig 3 5.2 psig
- b. Unit 2 N.A. M.A. N.A. 3 6.0 psig 3 5.0 psig
- 9. Containment Sump Recirculation
- a. Automatic Actuation Logic M.A. N.A. M.A. N.A. M.A.
and Actuation Relays
- b. Refueling Water Storage N.A. M.A. N.A. 3 177.15 inches 1 162.4 inches Tank level-Low Coincident With Safety ,
Injection See Item 1. above for all Safety Injection Setpoints and Allowable Values. TBD = TO BE DETERMINED
._= .. - - -- - _ _ _ . - _ - . . - . . _ _ - - _-
.- -Add i
, LIMITfNG SAFETY SYSTEM SETTINGS 8ASES
- Steam Generator Water Level ya hm- se p g
- . crew.
i , The Steam Generator Water Level Low-Low trip protects the reactor from j j loss of heat sink in the event of a sustained steam /feedwater flow mismatch resulting from41oss of normal feedwatert The specified Setpoint provides allowances for starting dela ys of the Auxiliary Feedwater Systee, as *.16 as
- creers M h twA Lw W en, i j Undervoltace and Underfrecuency - Reactor Coolant Puno Susses !
! The Undervoltage and Underfrequency Reactor Coolant Pump Bus trips pro-vide core protection against DNS as a result of complete loss of forced coolant i flow. The specified Setpoints assure a Reactor trip signal is generated before l
- the Low Flow Trip Setpoint is reached. Time delays are incorporated in the
Underfrequency and Undervoltage trips to prevent spurious Reactor trips from momentary electrical power transients. For undervoltage, the delay is set ; so that the time required for a signal to reach the Reactor trip breakers j e I following the simultaneous trip of two or more reactor coolant puso bus circuit ! breakers shall not exceed 1.2 seconds. For underfrequency, the delay is set l
) so that the time required for a signal to reach the Reactor trip breakers j
j after the Underfrequency Trip Setpoint is reached shall not exceed 0.3 second. i On decreasing powr the Undervoltage and Underfrequency Reactor Coolant Pump , t Bus trips are automatically blocked by P-7 (a power level of approximately 10% i i of RATED THERMAL POWER; with a turbine impulse chamber prescure at approxi-J mately 10% of full power equivalent); and on increasing power, reinstated i automatically by P-7. { j Turbine Trip t j A Turbine trip initiates a Reactor trip. On decreasing power the Reactor i trip from the Turbine trip is automatically blocked by P-9 (a power level of j i I approximately 69% of RATED THERMAL POWER); and on increasing power, reinstated ! I automatically by P-9. ! l Safety injection input from ESF If a Reactor trip has not already been generated by the Reactor Trip ] System instrumentation, the ESF automatic actuation legic channels will ! i initiate a Reactor trip upon any signal which initiates a Safety Injection, i l The ESF instrumeritation channels which initiate a Safety Injection signal ! i are shown in Table 3.3-3. l i '
! I l
l i 1
- CATAW8A - UNITS 1 & 2 8 2-7 l
1
.. - - -. - - = _ . . _ - - _ . _ . . . i l
( Att.achment. 2 l l IJiscussion, tio Significant llazards Analysis and Environmental Impact Statement i I : l i f i
'f.
1 I h i e l I r e f i t l l l i : l l l
DISCUSSION. NO SIGNIFICANT HAZARDS ANA13 SIS AND ENVIRONMENTAL IMPACT STATTMFNT i f The proposed an>endment would: ; (1) Change the Total Allowance, Z, Sensor Error, Trip Setpoint and Allowable l Value for Item 13.b., Table 2.2-1. (2) Change the Tot al Allowance, Z. Sensor Error, Trip Setpoint ; I and Alloewable Value for Items 5.6.2, and 8.c.2), Table 3.3-4. The changes to the tables are needed for Unit 2 only. Catawba Units 1 and 2 are equipped with Westinghouse Model D3 and Hodal D5 etean ; generators, respectively. A major difference in these nodels is the design of the moisturu separator section. Two aspects of this design difference are of j significance with respect. to the proposed modifications the D5 has a higher ; recirculation rato than the D3 and the elevation of the lower deck plate in the D5 is higher than in the D3. See Figuro 1. Due to these difforences the lower , instrument tap for the narrow range level instrumentation was located above the ! transition cone and lower deck plate on the D5 as opposed to below the transition cono in the downcomer region on the D3. This has resulted in significantly different operat.ing characteristics. The proposed modification will rolocate the ; D5 lower instrument tap to the same location as the D3. The old level j instrumentment taps will be capped. ! r Steam generators exhibit a "shrink" or "swell" characteristic at.Iow power Invels ; (when feedwater temperature is low) when feedwater flow is init.ially increased or l decreased, respectively. This nakes plant control difficult and more susceptible to t rips. The "shrink" and "swell" in the Model D3 is much less pronounced due ; to the location of the lower tap in the downcomer region. At the time that the l level "swells", the recirculation flow rate increases thus creating a lower ' pressure at the lower instrument tap and significantly reducing the magnitude of the indicated level increase. , In order t o dutermine t he potential gain in operational control characterist.ics [ of the D5 steam generator if the lower instrument tap were relocated to the l equivalent. location as the D3, Duke and Westinghouse installed instrumentation on [ the Catawba 2C generator. One narrow range level channel was installed using the : new lower tap location. A pitot tube was installed in another available location ! at the same elevation. The pitot tube static pressure connection was used to ( create another narrow range level channel which would be unaffected by the flow ; velocity in the downcomer. Transient data from these instrunents and the i currently installed narrow range icvel instrument. clearly showed (1) that the l shrink an:i swell is significantly reduced by the flow velocity feedback in the l dowcomer and (2) the indicated level when sensed in the downcomer and velocity , compensated correlates extremely well with the current instrument configuration. ! Fee Figure 2. l Due to the higher elevation of the lower instrument tap on the DS, a more restrictive instrumentation range is available to the operator. Steam generators I experience a significant shrink following a reactor trip from power. Emergency procedures require the operator to establish indicated level in the safety grade l l l I 1 ;
DISCUSSION. NO SIGNIFICANT llAZARDS ANAL.YSIS AND FNVIRONMENTAL IMPACT STATFNDIT narrow range instrumentation prior to throttling auxiliary feedwater flow below design values. Due to the wider span of narrow range level available on the model D3, if indicated level does fall off scale it. is generally recovered very quickly. On the model D$ it typically takes 10 minutes before indicated narrow range level is recovered. This creates a high potential for post-trip over cooling and depressurizat. ion. Transient. data has shown that the modified D5 level instrumentation will perform similarly to the D3 in terms of post-trip response. Seu Figure 3. The present span between the high-high level and low-low level trips on the D5 is physically bounded by the elevation of the top of the moisture separator swirl vanes and the elevation of the lower instrument tap, respectively. By relocating the lower tap, the le.-Irv letel trip setyi'.t can be reduced. The high-high level trip setpoint will be reduced by the equivalent error assigned to the maximum velocity in the downcomer. The operating level setpoint. will be reduced to maintain the currently existing margin (36") between this setpoint and the high-high level trip. The low-low level trip setpoint will be set. at the elevation of the lower deck plate. With this arrangement, the margin between operating level setpoint and low-low level trip set. point will be increased from a current 42" to 5B", a 38% increase. This will make the unit more tolerant to feedwater system malfunctions at power, thus reducing unnecessary trips and corresponding challenges to safety systems. Relocating the narrow range instrumentation lower sensing tap on the Westinghouse rodel D5 steam generator to the same elevation as the mode D3 steam generator will provide the following safety enhancements:
- 1. The effects of shrink and swell at low power levels will be greatly reduced, thus reducing the potential for reactor trips.
- 2. The time necessary to recover indicated 1cvel following a reactor trip will be greatly reduced, thus reducing the potential for an an over cooling event due to excessive auxiliary feedwater.
- 3. The margin to low-low level trip will be increased thus reducing the the potent ial for reactor trips at power.
Relocation of the level sensing tap to the downcomer region requires that. the velocity induced error be accoanted for in determination of trip and operating level setpoints. This can be acco:nplished without reducing any current margin to trip. 10 CFR 50.92 states that a proposed anendment involves no significant hazards considerations if operation in accordance with the proposed amendment would nott (1) Involve a significant increase in the probability or consequences { of an accident previously evaluated; or t I (2) Create the possibility of a new or different kind of accident from ! any accident previously evaluated; or l (3) Involve a significant reduction in a margin of safety, i l
DISCUSSION, NO SIGNIFICANT HAZARDS ANALYSIS AND ENVIRONMFNTAL IMPACT STATDENT
- There will not be a significant increase in the probability or consequences of en !
accident previously evaluated. The following FSAR Chapter 15 events have been preliminarily evaluated assuming the new set points shown in Attachment 1:
. loss of External Load / Turbine Trip (FSAR Sections 15.2.2 4 15.2.3)
. Steam System Piping Failure (FSAR Section 15.1.5)
. Loss of Non-Faergency AC Power to the Station Auxiliaries !
(FSAR Section 15.2.6) .
)
. Loss of Normal Feedwater Flow (FSAR Section 15.2.7)
. Feedwater System Pipe Break (FSAR Section 15.2.7) !
. Feodwater Systen Halfunction Causing an increase in Feedwater i 4 Flow (FSAR Seetion 15.1.2) !
. Mass and Energy Release Analysis for Fostulated Secondary Syste2n r Pipe Ruptures Inside containment (FSAR Section 6.2.1.4) !
The following events were re-analyzed because they are the events sust sensitive , - to a reduction in the low-low level trip setpoints: }
. Loss of Non-Faergency AC Power to the Station Auxiliaries j (FSAR Section 15.2,6) '
. Loss of Normal Feedwater Flow (FSAx Sect Sn 15.2.7)
. Feedwater System Pipe Break (FSAR Sectin 15.2.?) :
i a Information concerning Reactor Trips as a result of S'ean Cencrator water level can be found in FSAR sections 7.2.1.1.2 and 7.2.2.3.5. The evaluat. ion of the ' ] above events concluder. there is no significant incre.se in the probability or ; j consequences of an accident previously evaluated. r The proposed amendment will not create the possib',lity of a new or dif f erent i of accident from any accident previously evaluated. Relocating the level tap ! the D-5 stean generators snculd improve the ope'ation of the steam generators. j , Since the steam generators will be operated the same as before the lovel tap t j modification, no new or dif ferent kind of acclients are created.
- i The movement of the steam generator level taps will not involve a significant
] ( 4 reduction in any margins of safety. Furthermore, the modification will improve i I safety since the modified st.eata generators will be less susceptibis to f eedwater ; transients, thus reducing the potential for reactor and turbine trips and ! avoiding unnecessary transients on the primary and secondary systems. j i Based on t he above analysis, this amendment request does not involve Significant [ llazards Considerations. , i t Fnvironmental Impact l I The proposed Technical Specification change ans been revieve.1 an,ainst the f criteria of 10CFR51.22 for the environmental consider >tions. As shown above, the ! l proposed change does not involve a significant hazards consideration, nor increase the types and amounts of ef fluents that may be released of fsite, nor l increase individual or cumulative occupational radiation exposures. Based en the l foregoing, the proposed Technical Specification change meets the criteria given , in 10CFR51.22(c)(9) for a categorical exclusion from the requirement for an *
- Environmental Impact Statement. !
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