ML19347C073
| ML19347C073 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 10/01/1980 |
| From: | PUBLIC SERVICE CO. OF COLORADO |
| To: | |
| Shared Package | |
| ML19347C066 | List: |
| References | |
| TAC-47416, NUDOCS 8010160543 | |
| Download: ML19347C073 (52) | |
Text
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i ATTACHMENT 3 a
- PROPOSED TEGNICAL SPECIFICATION REVISIONS i
ir i
Section 2.0 Definitions 1 Page LSSS 3.3, Table 3.3.1 3 Pages LCO 4.4.1, Tables 4.4-1 4.4-2 4.4-3 4.4-4 8 Pages Notes for Tables 4.4-1 through 4.4-4 1 Page 1
Basis for LCO 4.4.1 4 Pages 80101606%.
o Fort S t. Vrain #1 Technical Specific:' c ions Amendment Page 2-2 2.0 DEFINITIONS 2.3 Instrumentation Surveillance Add new definitions d) and e) as follows:
d) Instrument Trip Setting A specified range in which the instrument is set to trip at the time of calibration.
The instrument trip settings listed in the Tables or LSSS 3.3 and LCO 4.4.1 are nominal values. Trip settings may be changed to provide more conservative trip points without prior License Amendment to Table LSSS 3.3.1 or Tables 4.4-1, 4.4-2, 4.4-3, or 4.4-4, provided a revision to the affected table is included with a subsequent Li-conse Amendment request.
e) Absolute Value The trip limit at which the consequences of accidents have been analyzed.
The margin between the absolute value and the instrument trip setting is adequate to assure the instrument chan-nel will trip and the protective action will occur before the absolute value is encountered, considering accumulated instrument channel inaccuracies.
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Speel f icat ion I.SSS 'I.3 - I.loiting Safety System Settings .
4 TABLE 3.1.1 Absolute i Parameter Function Value Instrument Trip Setting l
- 1. Reactor Core I. limiting Safety Systein Settings 7
l a) liigh Neutron Flux Scram < 140% of rated Progranuned with indicated thermal power power. Program shall be ,
approved by the Nuclear !
Fac ili ty Sa fe ty Convai t tee for em.h cycle.
b) ' lligli Relieat Steata bcram ,< 1075*F 1059 + 4*F Temperature c) I.ow Prinaary Coolant Scr a < 50 psi helow rated Category III Pressure programmed with load '
- 2. Reactor Vessel Pressure Liiniting Safety System Settings a) liigh Primary coolant Scram and Preselected < 51 psi above rated, Categw y III Pressi,re I.oop Shutdown and programmed with load.
Steam / Water Dump 'Ipper programmed 1imit set to produce trip at m .
-< 775 psia $ M*8 yam anon
- n. p h) liigh Holuture in the Scram, l.oop Shutdown < 67*F dewpoint tempera- Category III F N $ n" Primary Coolant and Steam / Water Dump ture (corresponds to YS$'
_< 500 ppmv l1.90 at 700 psia ym N, pressure) ar
- t o
3 E
a (a
Specification I.SSS 'i.3 - I.imit ing Safety System Set t ings (cont inucil)
TAlti.E 't . 3.1 (cont inneil)
Absolute Parameter Function Vaine Instrument Trip Settings
- 2. Reactor Vessel Pressure '
l.imiting Safety System I Settings (continued) f c) PCRV Pressure Pressure Rellef Rupture I)ise (l.ow <820 psig 812 psig i 1% ;
Set Safety Valve) {
i I.ow Set Safety $804 psig 796 psig i 1%
Valve Rupture Disc (Iligh 1840 psig 832 psig i 1% l Set Safety Valve) i liigh Set Safety $820 psig 812 psig i 1% !
Valve ,.
I d) llelium Circulator Pressure Relief -
Penetrat tim Inter-space Pressure Rupture Disc (2 per Penetratlon) 1842 psig 825 psig i 2% ygyy g,gg u 0' O vs Safety Valve (2 1829 psig 805 psig i 2% *g" per Penetratlon)
{nH y
$U c) Steam Generator Pressure Relief *p Penetration Inter- ,
Q space Pressure pe W
Rupture Disc (2 1842 psig 825 psig i 2% h for Each Steam g Generator)
Specification I.SSS 3.3-- 1.imiting Safety System Settings (continued)
(
l TAltl.E 3. 3.1 (continued) ;
Absolute Parameter Function Value Instrument Trip Set t ings, 20 Reactor Vessel Pressure I.imiting Safety System Settings (continued) e) Steam Generator ,
Pressure Heticf Penetration Inter-space Pressure (con- !
tinued)
Safety Valve (2 <489 psig 475 psig 4 2%
for Fach Steam Generator) .'
?*&n m&Rm
- 8 O rt VUk
- t rt H
$ mk e =
mW P3 n
p W
O (A
Specification 1.C0 4.4-1 I
TAHl.E 4.4-1 INSTRUM&fi' OPERATING RE0llIREMENTS FOR Pl. ANT PROTECTIVE SYSTDitSCRAH Hinimum Hinimum Permissible .
Absolute Operable Degree of Bypass Instrument Trip f No. Functional IJnit Value Cliannel s itedundancy Conditions Setting _ l la. Hanual (Control Room) 1 0 None ---
lb. Manual (Emergency Board) ---
2 (f) 1 None -------
- 2. Startup Channel - liigh <105 cps 2 1 Reactor Mode Category III Switch in "RtlN" ,
3a. I.inear Channel - liigh 1 140% power 2 (f) 1 None Programmed with in-Channels 3, 4, 5 (a) dicated power (u).
3h. 1.inear Channel - liigh 1 140% power 2 (f) 1 None Progranuned with in-(a)
Channels 6, 7, 8 dicated power (u).
- 4. Primary Coolant Holsture Category 111 liigh I.evel Monitor 1 67*F Dewpoint I (f,t) 1 (c) None 1.oup Honitor ~< 27*F Dewpoint 2/l.nop 1/ loop (h)
(f,t) E @
'E o{ @Er N
- 5. Reheat Steam Temperature ~< 1075*? (a) 2 (b)(f) 1 Nonc 1059* + 4*F ,o { S vi
- liigh (h) oa8" b" <:
- 6. Primary Coolant Pressure 1 50 psig below 2 (f)(k) 1 1.ess than Category III @N
- 1.ow normal, load 30% Rated $$
prograumied (a) Power $ m, re o
- 7. Primary Coolant Pressure 1 7.5% above 2 (f)(k) 1 None Category III $
- liigh normal rated, E load prograimned E (a)
Specification 1.C0 4.4-i jcontinued)
TAlli.E 4 4-1 (continued)
Hinimum Minimum Permissible f Absolute Operable Degree of Bypass Instrument Trip i No. Functional linit Value Channels Redundancy Conditions Setting 6' !
- 8. Ilot Reheat lleader 1 10 psig 2 (f) 1 1.ess than 35 i 5 psig ,
Pressure - Iow 30% Rated I Power f
- 9. Main Steam Pressure 1 1500 psig 2 (f) 1 1.ess than 1635 1 25 psig
- I.ow 30% Rated Power 10 Plant Electrical (d .t 2 (e)(f) 1 None System - I.oss
- 11. Two I.oop Trouble ---- 2 1 Reactor Mode -- --
Switch in
" Fuel 1.oading"
- 12. liigh Meactor Building < 325'F 2 (f) 1 None 175 i 10*F Temperature (Pipe Cavity) m s ,-n m N tb O '
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Speciffeation I.00 4.4-1 (continued)
TAlli.E 4.4-2 (cont inued) l Hinimum Minimum Permissible Absolute Operable Degree of Bypass Instrument No. Functional Unit Value Channels Itedundang Conditions Trip Setting i 3h. loop 2 Shutdown I.ogic -- -
2 1 None - - - - -
4a. Circulator IA and 111 Circtlator 2 1 None -
Sluitdown - Inop Shut- IA ar.d IB !
down Logic Shutdown 4h. Circulator IC and ID Circulator 2 1 None ---
Shutdown - loop Shut- IC and ID i down 'ogic Shutdown Sa. Steam Generator Pene- < 810 psig 2 (f) 1 None 757 1 5 psig ,
tration overpressure, !
Imop L Sh. Steam Generator Pene- < 810 psig 2 (f) 1 None 757 1 5 psig !
tration Overpressure, !
loop 2 6a. liigh iteheat Ileader < 5 mr/hr 2 (f) 1 None Category III Activity, loop 1 Above llack-ground ,
m 6h. liigh Reheat Ileader < 5 mr/hr 2 (f) 1 None Category III $ g16-i$m0
- Activity, loop 2 Ibovellack-
- O. $
1 e un ground .d Mnn I f, $ '
7a. Iow Superheat lleader -> 800*F 2 (f) 1 1.ess than Category III un #
Temperatiere, Isop 1 (p) 30% Rated E Power S.
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7h. Iow Superheat llender ,> 800*F 2 (f) 1 1. css than Category III O Temperature, loop 2 (p) 30% Rated E Power O m
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Specification 1.C0 4.4-1 (continued) ,
TA111.E 4.4-2 (continued) '
f i
Minimum Minimum Permissilele '
Absolute Oceralsle Degree of Bypass Instrument j No. Functional linit Value Clia 'nels Redundancy Conditions Trip Setting 7c. liigli Dif ferential. Tem- < 50*F 2 (f) 1 None Category III !
perature Between loop 1 and loop 2 (p)
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M cification I.C0 4.4-1 (continued) ,
TAllI.E 4.4-3 INSTRllHENT OPERATING REQUIREMENTS FOR Pl. ANT PROTECTIVE SYSTEH, CIRCIILATOR TRIP _
Minimiun Minimum Permissible Absolute Operable Degree of Bypass Instrument ,
- No._ Functional Unit Value Channcis Redundancy conditions ,Triy Setting i
- 1. Circulator Speed - low $ 2390 rpm 2 (f) 1 1.ess than 1910 1 75 rpm j (r) Below Normal 30% Rated Below Normal as !
as Programened Power Progransned by by Feedwater Feedwater Flow ,
Flow l s
- 22. l.oop 1 Flxed Feedwater 2 5% of Hated 2 (f) 1 1.ess than 10 1 0.5%
Flow - low (Both Circu- Full Load 30% Rated of Rated Full lators) Power load !'
2b. I.oop 2, Fixed Feedwater 2 5% of Hated 2 (f) 1 1.ess Than 10 1 0.5%
Flow - Low (Both Circu- Full 1.oad 30% Rated of Rated Full ,
lators) Power Load
- 3. loss of Circulator 2 450 psid 2 (f) 1 None 475 +10, -0 psid Bearing Water (r)
- 4. 'treulator Penetration 5 810 psig 2 (f) 1 None 757 1 5 psig Trouble (r)
'o
- 5. Circulator vain Hal- 2 5 paid 2 (f) 1 None 7 + 1 paid $ pW*
s =s
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function (r)
Circulator Speed - High 5 11,000 rpm 2 (f) 1 Nans 10,770 1 60 rpm
- h" 7 S$*
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- 7. Hanua1 --------
1 0 None' O. "
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- i Specification I.00 4.4-1 (continued) ;
TABl.E 4.4-4 i i
INSTRUMENT OPERATING REQllIRi}fENTS FOR REACTOR PROTECTIVE SYSTEH, HOI) WITIIDRAWAI. PRollIBIT (RWP) j 1
Minimum Minimum Permissible !
Absolute Operable Degree of Bypass Instrument :
No. Functional Unit Value Channels Redunilancy Conditions Trip Setting
- 1. Startup Channel - I.ow > 2.5 cps 2 1 Above 103% 5 i 1 cps !'
Count Rate Rated Power 2a. I.inear Channel - Iow N.A. 2 1 (g) 5 1 2% (m) ;
Power RWP (Channels 3 j 4, and 5) 2b. 1.inear Channel - Iow N.A. 2 1 (g) 5 1 2% (m)
Power RWP (Channels 6, 7, and 8) 3a. I.incar Channel - Ilf gh N.A. 2 (f) 1 None 30 t- 2% (n) .
Power RWP (Channels 3 l 4, and 5) 3h. 1.inear Channel - liigh N.A. 2 (f) 1 Hone 30 1 2% (n)
Power RWP (Channels 6, 7, and 8)
- 2{@o' naa
- n. n
. 6 O $
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a" a3 0
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l Fo rt .S,t . Vrain #1 l Technical Specifications
! Amendment
! Page 4.4-8a l
l SPECIFICATION LCO 4.4-1 (continued) l NOTES FOR TABLES 4.4-1 THROUGI 4.4-4 l
Ar*.d new Note (u) to read as follows:
l l (u) The progra:rted instrument trip setting is to be approved by the l
Nuclear Facility Safety Committee for each refueling cycle con-sistent with approval of the control rod sequencing pattern spe-cified in LCO 4.1.3.
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Fort St. Vrain #1
. Technical Specifications Amend =ent Page 4.4-9 Basis for Specification LCO 4.4.1 The plant protection system automatically initiates protective functions to prevent establish *d limits from being exceeded. In addition, other protec-tive instrumentation is provided to initiate action which mitigates the con-sequences of accidents. This specification provides the liniting conditions for operation necessary to preserve the effectiveness of these instrument systems.
If the minimum operable channels or the mini =um degrees of redundancy for each functional unit of a table cannot be =et or cannot be bypassed under the stated permissible bypass conditions, the following action shall be taken:
For Table 4.4-1, the reactor shall be shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
For Table 4.4-2, the affected loop shall be shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
For Table 4.4-3, the affected helium circulator shall be shutdown with-in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
For Table 4.4-4, the reactor shall be shutdown sithin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
j If, within the indicated time limit, the minimum number of operable channels
! and the mini =um degree of redundancy can be reestablished, the system is con-sidered normal and no further action needs to be taken.
The absolute values and instrument trip settings are included in this section of the specification. The bases for the absolute values are briefly dis-cussed below. Additional discussions pertaining to the scram, loop shutdown and circulator trip inputs may be found in Section 7.1.2.3, 7.1.2.4, and 7.1.2.6, respectively, of the Final Safety Analysis Rayort. High moisture idstrumentation is discussed in Section 7.1.2.5.
a) Scram Inputs Manual Scram is provided to give the operator means for emergency shut-down of the reactor independent of the automatic reactor protective sys-tem.
Startup Channel - High Countrate is provided as a scram input during fuel loading and tero power operations.
Linear Channel Flux - High (see Technical Specification LSS3 3.3).
High Reactor Moisture (see Technical Specification LSSS 3.3) .
High Rehene Svstem Temeerature (see Technical Specification LSSS 3.3). l l
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Fort St. Vrain #1 Technical Specifications Amendment Page 4.4-10 Basis for Specification LCO 4.4.1 (continued) a) Scram Inputs (continued)
Low Reactor Pressure is an indication of possible helium leakage from the system. A scram is required because the reactor is in danger of being inadequately cooled which would increase the hatard associated with activity release from the PCR7. The trip is programmed with plant load (similar to the high pressure trip) to reduce the response time when the plant is at high power. The low pressure trip point is 50 psi below normal during operation between 30% and 100% rated power which is lower than the pressures reached on normal transient conditions.
High Primary Coolant Pressure (see Technical Specification LSSS 3.3).
Low Hot Reheat Steam Pressure is an indication of either a cold reheat steam line rupture or a hot reheat steam line rupture and necessitates plant shutdown due to the potential loss of steam turbine circulator i motive power. The absolute value is selected to be below normal op-j erating levels which vary over a wide range.
Low Main Steam Pressure is an indication of main steam line rupture or loss of feedwater flow and necessitates plant shutdown due to potential loss of steam turbine circulator motive power. The absolute value is selected to be below normal operating levels. ,
Plant Electrical System Power Loss requires a scram to prevent any power-co-flow mismatches from occuring. A 30-second delay is provided fol- ;
lowing a power loss before the scram is initiated to alle s the emergency I diesel generator to start. If it does start, the scram is avoided.
Two-Loco Trouble. Operation on one loop at a maximum of about 50: power may continue following the shutdown of the other loop (unless preceded by scram as in the case of high moisture). Onset of trouble in tSe re-saining loop (two-loop trouble) results in a scram. Trouble is defined as a signal'which normally initiates a loop shutdown. SLn11arly, simul-taneous shutdown signals to both loops result in shutdown of one of the two loops only and a reactor scram.
High Temperature in the pipe cavity would indicata the presence of an undetected steam leak or the failure of the steam pipe rupture detection -l system to differentiate in which loop the leak had occurred and to shut I
.the faulty loop down.
The absolute value is selected to be above the temperature that would be expected to occur in the pipe cavity if the steam leak were detected and the faulty loop shutdown for all steam leaks except those of major proportion or'due to an offset rupture of one of the steam lines.
Fort St. Vtain #1 Technical Specifications
- Amendment Page 4.4-11 i
i Basis for Specification LCO 4.4.1 (continued) a) Scram Inputs (continued)
High Temperature (continued)
An undetected steam leak or pipe rupture under the PCRV within the sup-port ring would also be detectable in the pipe cavity, therefore only one set of sensors and logic is required to monitor both areas.
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b) Loop Shutdown Inputs Steam Pipe Rupture in the Reactor Building necessitates shutdown of the leaky loop to terminate the pressure and temperature buildup within the' building. Ultrasonic noise caused by escaping steam in conjunction with a pressure or temperature rise will cause the appropriate loop to shut-down. ,
The absolute value of the ultrasonic detection system is selected to be at a level which corresponds to 9 v.dc output from the ultrasonic ampli-fier. The pressure and temperature trips are set above normal operating building pressurs and temperature levels.
i Shutdown of Both Circulators is a loop shutdown input which is necessary to insure proper action of the reactor protective (scram) system (through the two-loop trouble scram) in the event of the icss of all circulators and low feedwater flow.
The remaining loop shutdown inputs are equipment protection items which are included because their malfunction could prevent a scram due to loss of the two-loop trouble scram input.
c) Circulator Shutdown Inputs All circulator shutdown inputs (except circulator speed high on water turbines) are equipment protection items which are tied to two-loop trouble through the loop shutdown system. These items are included in Table 4.4-3 because a malfunction could prevent a scram due to loss of the two-loop trouble scram input. Circulator speed high on water turbines is included
.co assure continued core cooling capability on loss of steam drive.
d) Rod Withdraw Prohibit Inouts Startup Channel Countiate - Low is provided to' prevent control rod witn-drawal and reactor startup without adequate neutron flux indication. The trip level is selected to be above the background noise level, i
( Linear Channel (57. Power) directs the operator's attention to either a
- downscale failure of a powe; range channel or improper positioning of the I.S.S.
Fort S t. Vrain #1 Technical Specifications Amendment Page 4.4-12 Basis for Soecification LCO 4.4.1 (continued) d) Rod IJithdraw Prohibit Inputs (continued)
Linear Channel (307. Power) is provided to prevent control rod withdrawal if reactor power exceeds the I.S.S. limit for the " Low Power" position of the I.S.S. or if the I.S.S. is improperly positioned.
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t _
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-- - . . ~ , .. . . - - . - - -- ., . . , ,
ATTACW.ENT 4 SUPPORTIVE INFORMATION FOR EVALUATION OF HOT REHEAT HEADER PRESSURE - LOW TRIP SETTING l
Previously Submitted to the NRC in P-77228 - Reference 6 1
f
Attachment 4 Page 1 of 1 The selection of this reactor scram is based on its being an indication of l
either a cold or hot reheat steam line rupture which necessitates a plant shutdown because of the potential of losing circulator steam i trbine drive steam. The current version of Table 4.4-1, item 8 trip settings, only al-lows 35 psig as the trip limit. To allow for expected instrument drif t and l calibration uncertainties of 10 psig, the settings had previously been ad-l justed to 45 psig.
A setting of 45 psig reduces the margin between the normal range of hot re-heat pressures and the scram actuation point such that unnecessary scrams
- would occur during normal operation. To correct this situation, it is pro-posed to return the serpoint to 35 psig and add the designation of " Absolute Trip Value" = 10 psig. The interpratation of this value is that the cali-bration check of "as found" crip setting must not be less than this value, j The nominal trip setting for calibration and functional test would remain The margin between the setpoint of 35 psig and the absolute at 35 psig.
l trip value of 10 psig is an allowance for system inaccuracies and calibra-l tion uncertainties.
l l The only effect should the pressure sensors fail to actuate a trip until a pressure of 10 psig were reached, would be to delay the time of scram actua-tion. If the low pressure was caused by a break of the cold reheat loop common piping or one loop hot reheat pipe, the scram actuation delay would amount to 0.5 seconds and 1.0 second, respectively. Neither of these time delays alter the safety evaluation presented in Final Safety Analysis Re-port Section 14.4.2. The maximum system inaccuracies and calibration un-certainties hLve been reported previously as 10 psig. Therefore, with the trip setting at 35 psig and maximum possible subsequent drift of setpoint to 25 psig, a least margin of 15 psig remains between the actual setpoint and the absolute value.
1 We therefore conclude that the trip setting can be safely returned to a value of 35 psig consistent with adequate calibration uncertainties and in-accuracy margin and that the value of 35 psig should not interfere with the normal range of operating pressures for the hot reheat system.
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%e4e.- h -- + m..h w- 4w e e ATTACHMENT 5 SUPPORTIVE INFORMATION FOR REEVALUATION OF LOW CIRCULATOR SPEED TECHNICAL SPECIFICATION LIMITS 1
i i
. . . . . . , _ -. . _ - - . , - , - . . . - -- . - ~ - - - - - --
Attech=ent 5 Pags 1 of 4 SUPPORTIVE INFORMATI0M FOR REEVALUATION OF LOW CIRCULATOR SPEED TECHNICAL SPECIFICATION LIMITS 4
a Low circulater speed is an indication of a speed control or other equip =ent failure and results in a decrease in loop helium flow. This leads to a mis = acch between heat input and heat re= oval (feedwater flow) in a steam generator. Such a mismatch can result in a low superheat header te=perature and a high differential te=perature between loops. This condition, in turn, causes a Plant Protective System (PPS) loop trip to prevent steam generator sain bundle floodout.
i To minimize the necessity of tripping a loop when only one circulator is =alfune-
- tioning, a circulator PPS trip on Jov circulator speed has been incorporated. Upon circulator trip, a power runback to 50% (when initial power is greater than 307.) is
. initiated and the re=aining circulator in the affected loop is released to exceed 10% of the progra==ed speed (progra==ed with loop feedwater flow). Thus , two-loop i plant operation can be =aintained at up to 50% power with the re=aining circulator j providing full flow for the affected loop.
1 \
This docu=ent provides information supporting a revision in the low circulater l speed PPS limits which are defined in the Fort St. Vrain Technical Specification
! Table 4.4-3 of LCO 4.4.1. The instru=ent trip setting re=ains at the present a Technical Specification value of 1910 rpm (20% of design circulater speed) below nor=al operating speed progra==ed with loop feedwater flow. The absolute value
- which provides allowance for instru=ent calibration error and instru=ent drift is 1 2390 rpm (25% of design circulator speed) below nor=al operating speed progra==ed l with loop feedwater flow. The supporting safety analyses assu=e that the instru-
=ent trip does not occur until the absolute value is reached.
Discussion i
FSAR Section 7.1.2.6 presents a discussion of circulater trip design for the Fort St. Vrain plant. (So=e supple = ental infor=ation is provided in Section 4.3.1.)'
I Note that a 5-second delay in the safety action is included to discri=inate against transient deviations. Trip of both circulators, discussed briefly in .
Section 7.1.2.3, results in a loop shutdown. In the case of speed decay on one or both circulators in a loop, the low superheat header ce=perature PPS setpoint is always encountered with sufficient =argin to prevent flooding (floodcut) of the I superheater section of the steam generator. Consequently, an evaluation of changes in the low circulator speed limic =ust be perfor=ed in the context of the increase (or decrease) in the possibility of tripping a loop.
For the case where both circulators in a loco loose scead relative to the nor=al.
, operating speed progra=med with flow, a loop trip will always result due to either a PPS trip of both circulators or encountering the low superheater header ta=pera-l 2
ture PPS setpoint. Hence only the case of a speed decay of a single circulater !
will be considered in detail.
I i
I
, Attachment 5 l .
Page 2 of 4
' SUPPORTIVE INFORMATION FOR REEVALUATION OF LCW CIRCULATOR oo o - -
SPEED TECHNICAL SPECIFICATION LIMITS D ,
oo v[ - 3
, The ti=e constant for the response of a circulator to a change in driving conditions '
is of the order of three seconds, as noted in Section 4.3.1 of the FSAR. Further, the circulator turbine bypass and throttle valves are hydraulically driven and capable of a full stroke in approximately one second. Conversely, the ci=e constant I
of the steam generator superheater header te=perature responding to a change in i
helium flow is approx 1=ately 30 seconds (an order of =agnitude greater). A i
relatively rapid (less than 30 seconds) reduction in circulator speed of 2390 rp=
l will, therefore, result in a single circulator trip followed by a power runback l transient (if required) and ce=pensation by the remaining circulater without ancountering the icw superheat header te=perature se: point. To confir: this, a
! nu=ber of transient cases were reviewed for various nu=bers of circulators at j various power levels assu=ing the rapid closure of a circulator control valve.
Results, given in Table I, show that the circulator speed trip occurs well before the steam generator superheater header te=perature appreaches the 1cep trip setpoint.
A relatively slow (greater than 30 seconds) decay in circula:or speed of 2390 rps would probably result in a icop trip en low superheat header te=perature before the
! circulator speed trip is encountered. The probability of a loop trip is increased at lower pcuer levels since the nor=al main steam ta=perature is d: coped closer to the =ain steam ta=perature PPS setpoint. The change in probability of encountering a leop trip in utilizing an absolute trip setting of 2390 rp= rather than 1910 rp=
below nor=al operating speed for slew circulator speed decays is considered insignificant.
- The effects of a circulator speed decay during operation with a total of two
- circulators and a single loop is ad-dlar to the above (see Table I) . Again the change in setpoint li:1c has no significant effect en the probability of a loop
, trip. For operatica with one circulator in each lcep, a loop trip due to low circulator speed will occur for rapid speed decay; a loop trip due to low super-heat herder ta=perature =ay occur in the event of a slow circulater speed decay, i
Instru=entation Accuracy The circulator speed-low progra==ed with loop feedwater flow receives processed -
signals from both-para =eters. The circulator speed portica consists of a speed l
clement, a speed = edifier, and a speed transmitrer which in turns sends a signal l to the speed switch icw. The feedwater por*:?aa consists of a ficw ele =ent, a l
flow transmitter, a flow = edifier, and a 1 ev ucdifier characterizar which in turns sends a signal to the speed swir. ) f y. , Using =anufac:urers published 1 I
tecuracy data and ce=bining the= by L-e 134 - squares =ethod the overall accuracy is t2.6%. At 100% feedwater flow, the progrs= ed circula:cr speed is currently 9550 rpm for 4 circulator (2 per loop) operations. This conver:s to an instrc-
= ntation accuracy of t250 rp=. Two se:s of calibration data are available to
- _ cbtain a ceasure of instru=ent decalibration or drif t. Ihese calibra
- icas were parfor=ed using calibration procedure 5.4.1.3.2d-R on 6-30-78 and 6-29-79. A total of twelve instru=ents were calibrated each at 100, 75, 10, and 25" feed-water flow or a total of 48 calibration poin:s. The differenes in the "as left" vorsus the "as found" cendition average abcut -4 rp= with the =axi=u= values bning -75 rps and + 45 rp=. These recorded differences could be ei:her calibra-tion inaccuracy or drift. Based on these data, it is concluded the: :he abcVe l instruments perfor=ance is consistent with the =anufacturers published accuracy l .dnta.
t 1
Attachment 5 Pagn 3 of 4 Conclusion and Recommendations A change in the low circulator speed PPS limit (Fort St. Vrain Technical Specifi-cation Table 4.4-3 of LCO 4.4.1) to provide allowances for instrument calibration error and drift has been analyzed. Although the instrucent trip setting remains at the present Technical Specification value of 1910 rpm (20% of design circulator 7
speed) below c.ormal operating speed progra==ed with loop feedwater flew, an abco-lute value of 2390 rpa (25% of design circulator speed) below nor=al operating speed progra==ed with loop feedwater flow has no significant effect upon the prob-ability of a loop trip in the event of a single circulator =alfunction. Cases 4
involving the rapid decay of circulator speed are unaffected. Protection against steam generator fl.codout continues to be =aintained by the low superheat header te=perature PPS trip for slow decay of both or a single circulator within a loop.
Therefore, consequences of a circulator speed decay are unchanged.
It is recoc= ended that the Technical Specification be revised as follows:
Table 4.4-3 Instru=ent Absolute Value Trio Setting
- 1. Circulator Speed-Low < 2390 rpm below normal 1910 + 75 below nor=al prograc=ed with loop progra==ed by loop feedwater flow feedwater flow The =inimum difference of 405 ' rpm from the instru=ent setpoint tolerance range is more than sufficient to accoc=odate instru=ent inaccuracy.
/
6 6
TABLE I EFFECTS OF SINGLE CI,RCtILATOR SPEED I.0SSES i
{
I SUDDEN FAILURE OF ONE CIRCtILATOR WI'llt STEAll P0h'ER INITIAL CONDITIONS TO TURilINE STOPPED IN 1 SECOND j NO. OF NO.OF PLANT HAIN TIME TO PPS CIRC IlI((jUIhIh
- g. N IN PLANT LOAD LOOPS CIRCULATORS LOAD STEAM TRIP AT AllSOLUTI- AT CONCI.USION AFFECTED LOOP CASE OPERATING OPERATING (%) TE.51P". ( F) VALU (%) I
. 1)URIliG TI:AXSIENT _
1 2 4 100 1000 <7 960 50 j 2 2 4 50 971 <8 945 50 3 2 4 25 880 <8 865 25 l 4 1 2 50 1000 <7 940 32.5 5 2 1/ LOOP 50 971 <8 910 32.5 .
6( ) 1 1 25 971 <9 970 0 (1) Includes 5-accond delay in PPS circuitry For this case with the last operating circulator M (2) reaching the trip level, the circulator is not D 2%
tripped. An automatic otart is made on the water 2E'E) $k turbine. *1bo loop trouble would result in a reactor E3 scram. Rh"
. >?niv ,.
(3) At time of reaching circulator trip absolute value W
3En b-
l
.-. - - - - - ~ . - . . . -_ _ . . . . . _ . _ _ _ _ . . .
l l
l i
I l
l 1 l
l l
l ATTACHMENT 6 SUPPORTIVE INFORMATION FOR EVALUATION OF FIXED FEEDWATER - LOW TRIP SETTING Previously Submitted to the NRC in P-77188, Reference 5 l
l I
i i
e e
2 Attachment 6 Page 1 of 14 4
i f Introduction 1
The fixed feedwater flow trip is used to prevent steam generator operation at tube temperatures above design values in the event of a sudden loss of j 'feedwater flow while the plant is operating with the Interlock Sequence i
Switch (ISS) in the power position (power greater than 25 to 30%). The
- Final Safety Analysis Report also discusses the use of this trip to preclude 4 continued operation of.the plant below feedwater flows which could result in I unstable flow conditions in the steam generator. At' power levels less than 25 to 30%, these trips are not considered since they are normally locked out, i.e., they are not active when the ISS is in the startup or low power posi- '
tion.
J
! There are two aspects to the trip for sudden loss of feedwater flow, the
] minimum magnitude of feedwater flow and an allowable time delay for trip ac-
- cuation. These two aspects have been examined. The results indicate that
! a feedwater flow of zero and a time delay of 5 seconds for trip actuation j will not result in the tube metal temperatures exceeding ASME-Code allowable values. Three general modes of operation were considered in analyzing plant response to loss of feedwater: two loop operation at 100% and 25% initial j- load condition and one loop operation at 50% plant load (100% loop load).
2 Reanalysis of potential unstable flow conditions in the steam generator due i to low feedwater flow have been performed using actual and projected plant
- operational data for the Fort St. Vrain startup program. Also used in the
- reanalysis are new experimental data which was not available at the time of j the Final Safety Analysis Report analysis. These analyses indicate that unstable flow conditions could occur below about 18% feedwater flow but such j instability would not cause steam generator damage even if they persisted 4 for long time periods.
! A reduction in low feedwater flow trip setting is desirable because operating j experience has indicated unnecessary loop shutdowns can occur due to feedwater
! flow less than 20% for five seconds during reactor scram and turbine trip
- transients. Therefore, to prevent future unnecessary loop shutdowns due to feedwater flow fluctuations, it is recommended that the Plant protective Sys-i tem trip setting as specified in the Technical Specification LCO 4.4.1 be reduced from 20% to 10% for the fixed feedwater flow - low. This setting is consistent with an absolute value of 0% and the associated instrument accuracies. ,
t
.An alarm would be provided at 20% feedwater flow to warn the operator that the plant is being operated below design conditions and thus alert the operator to take corrective action.
j J
Discussion
- 1. Sudden Loss of Feedwater Flow Analyses were performed to show that allowable generator tube metal tem-peratures would not be exceeded if feedwater wcre lost with the plant initially in:
h Attachment 6 Page 2 of 14 Discussion (continued)
- 1. Sudden Loss of Feedwater Flow (continued)
- a. Two loop operation at 100% power.
- b. Two loop operation at 25% power.
- c. One loop operation at 50% power.
Two types of failure were considered. These combinations bracket all credible cases. In the first case sudden feedwater isolation at the feed-water end of the steam generators would permit for the time period'of in-terest the inventory to boil off and continue to add to the steam produc-tion from the second loop. Helium circulation consequently remains high for the duration of the accident.
In the second case, a feedwater line rupture at the steam generator inlet could, in effect, produce a steam / water dump resulting in termination of steam production from the effected loop in about two seconds. Total steam available to drive the helium circulators quickly drops to 50% and the circulators accordingly must slow down.
From the above, the limiting conditions arise from the first case in which helium circulation remains high for the accident time period of interest.
Accordingly, this case determines the required protection for the steam generator and was selected for analysis. As an upper limit, the helium flow was assumed to remain constant at 100% of rated during the transient.
A. Sudden Loss of Feedwater Flow in a Single Loce During Two-Loon Ooera-tion The system transient analyses was perfor=ed using the TAP code for an assumed total loss of feedwater flow to a steam generator, starting at 100%'and 25% initial plant power. Due to linitations of the TAP code, it was necessary to simulate the sudden loss of feedwater in a three-second ramp from rated plant flow to zero flow. In addition, the TAP code provides only the response of a " nominal" tube within a ;
" nominal" steam generator module. Accordingly, it is necessary to i modify the reference data to account for (1) instantaneous loss of feedwater and (2) behavior of the hottest tube in the hottest module. l Allowable delay times for loop shutdown can then be established to i preclude tube failure (defined herein as not exceeding the allowable I tube temperature limit at the corresponding tube pressure). l
\
The three-second ramp down in feedwater flow rate was ccmpensated for !
. by conservatively assuming the hottest tube changes temperature at the I i
.sans rate as the nominal tube. The nominal tube temperature was com- )
pensated for by adding the appropriate aTs for the hottest tubes.
t
1 Attachment 6 Page 3 of 14 Discussion (continued)
- 1. Sudden Loss of 7eedvater Flow (continued)
A. Sudden Loss of Feedwater Flow in a Single Looo During Two-Looo Ooera-tion (continued)
The maximum allowable tube temperature is dependent upon the coinci-dent differential pressure to which the tube is subjected. Since the differential pressures to which the tube is subjected are signifi-cantly less than design values for the postulated accidents, the al-lowable temperatures are significantly higher than design. Ihis analy-sis assumes that the loss of feedwater results from closure of a feed-water block valve and establishes allowable tuSu temperatures at an arbitrarily selected time of six seconds after the loss of flow, at 2
which time steam production in the failed loop has fallen to 66% of normal. (This is a conservative approach since the reduction in steaming rate at longer ti=es would result in slightly higher allow-able tube temperatures.) The reduction in steam generator tube pres-sure and resultant increase in tube allowable temperature is illustrated in Figure 1. Peak tube temperatures are also shown at the time of com-pleting of loop shutdown and illustrate that Superheater I outlet is the critical location. The other way to lose feedvater (steam / water dump due to a line rupture) results in significantly higher allevable temperatures.
Once the loss-of-feedwater-flow trip point is reached and the built-in Plant Protective System delay time expires, a signal is generated to shut down the circulators in the effected loop. Circulator speed valve stroking time, full open to close, is 1.5 seconds. After closure of the speed valves, the affected circulators will coast down quickly due to the bearing and gas retarding torque. With the circulators in the other loop still operating and providing a loop flow resistance, the shutdown circulators will go into stall early in their coast down ti=e.
At this point, the shutoff valve will close thereby stopping all helium flow through their loop. The coast time from the speed valve closure to stall has been determined to be 0.3 and 0.7 seconds at 100% and 25%
plant power, respectively.
The results of the analyses as discussed above are presented in Table
- 1. The allowable Plant Protective System delay after reaching the proposed minimum 0% trip setting to reach the design allowable =etal temperature for the hottest tube exceeds the five-second Plant Protec-tive System delay currently incorporated into this Plant Protective System trip setting. The margins are much larger at the reduced initial 25% plant load conditions.
k !
l Attachment 6 l Page 4 of 14 ;
Df scussion (continued)
- 1. Sudden Loss of Feedwater Flow (continued)
B. Sudden Loss of Feedwater Flow in a Single Loop During Single Loop Operation l
1 It is assumed that the accident is initiated from a conditien where !
l only one loop is operating and the plant is producing about 50% of l rated power. Thus, the operating loop is running at about 100% of ;
, loop design power. The major difference between this case and'that I previously considered is that there is no additional source of steam available to maintain helium circulator speed and thus helium circu- I lation decays rapidly with onset of the accident if the accident is due to blockage of feedwater at the steam generator inlet. If due to a line rupture causing a steam water du=p, the almost instantan-eous depressurization of the steac generator would cause an even more rapid loss of helium circulation.
f l A second difference between this and the previous case is that there l is no restraining torque on the helium circulators in the failed loop due to continued operation of the circulators in the failed loop due to continued operation of the circulators in the second loop. Thus, the circulator coastdown time is longer and will con-tinue down to about 500 rpm at which time the helium shut-off valve
- closes under its own weight. The time for this to occur is 12 sec-onds as shown in Figure 2.
, This sudden loss of feedwater flow does not permit retention of the i conetant helium flow and thus limits the metal tube temperature rise.
The nominal temperature rise for this accident predicted by the TAP code is shown in Figure 3. The allowable Plant Protective System de-lay time for this assumed accident is presented in Table 2. The same corrections are made for the hottest tube as opposed to the i nominal tube and for increased allowable tube temperatures based upon !
the actual tube differential pressure being less than for design dif- j ferential pressure conditions. The allowable Plant Protective System j delay times for not exceeding maximum allowable tube temperatures all :
exceed five seconds with a trip setting at the proposed mini =um value I of 0%.
- 2. Steam Generator Boiling Stability A. Static Stability During the Final Safety Analysis Report design period, boiling stability of the steam generators at low feedwater flow conditions was a major concern because of possible damage by overheating of the Economizer-Evaporator-Superheater I section tubes. Since then, two changes have
h Attachment 6 Page 5 of 14 ;
Discussion (continued)
- 2. St4am Generator Boiling Stability (continuea) l A. Static Stabiliev (continued) occurred to significantly lessen this concern. These are (1) new ex-perimental data has become available permitting more accurate analysis of boiling instability and (2) plant operating conditions demenstrated by the Fort St. Vrain startup program are different and less conducive to boiling instability tube damage. ,
Static instability, in the extreme, could result in cessation of flow through a few tubes. Should this occur, the stagnant tubes would quickly approach the temperature of the ambient helium. Data from DC-1-4 (see Table 3) show that a helium temperature of about ll36*F was anticipated at the inlet to Superheater II, at 25% feedwater flow.
Since this temperature is greater than the maximum allowable te=pera-ture of the ZESI tubes (1080*F at the operating tube pressure stress),
plant operation in an area of potential instability (sirnificantly less than 25% flow) was precluded by imposition of a low feedwater loop trip set at 20% feedwater flow when the ISS is in the power position.
Plant operation data during the rise-to-power program, however, has shown that actual operating conditions are significantly different from the original anticipated conditions due to regenerative heat transfer between steam generator lead-in and lead-out tubes. Presently predicted operating conditions for three cases (26.1%,18.6%, and 15.8%
feedwater flow rates) are shown in Table 3. (For comparison, antici-pated operating conditions at 25% feedwater flow frem DC-1-4 (January 30, 1970) are also shown.) Analysis of these three cases shows that the steam generators are statiscally stable for 26.1% and 18.6% feedwater, but are unstable at 15.8% feedwater. Howover; potential instability below 18.6% feedwater is now of =uch less concern for the following reasons:
First, it should be noted that the EESI tubes operate at a significantly lower differential pressure stress (approximately 1800 psi) than design values (e.g., 2920 psi for the economizer lead-in tubes). Thus, the ASME code mmvimum allowable temperature is considerably higher than that corresponding to the design pressure as shown in Table 4. For the limiting location, the economizer lead-in tubes, the =ax1=um allowable mean wall temperature is about 952*F compared to the design temperature of 500*F corresponding to the design pressure.
Second, helium temperature at Superheater II inlet decreases with lead between 26.1% and 15.8% feedwater flow. Thus, even if unstable boiling conditions are encountered at flow rates below 18.6%, the maxi =um helium temperature available at Superheater II inlet would be less than 957'F and thus could not result in significantly exceeding the maximum allow-able temperature at the limiting tube location (952*F). Note that this
. . , .,e.mk .~.s .=vn~*.eemw--- -ws=="~ ~~~' ~ * * = ' ' ~ ~ ~ * ' ' " " " " * ^ '
1 Attachment 6 i
Page 6 of 14-Discussion (continued)
- 2. Steam Generator Boiline Stability (continued)
A. Static Stability (continued) approach is quite conservative in that it postulates that a hot gas streak could penetrate the entire EIS bundle from top to bottom with no mixing - a highly improbable situation. Any significant gas mixing would, of course, result in much lower gas temperatures and in corre-spondingly less concern about boiling stability. ,
S. Dynamic Stability Dynamic boiling instability is of concern since it can result in cyclic variation in stean/ water flow rates with corresponding cyclic variations in tube wall temperatures. In the extreme, such variations could re-sult in excessive tube bending stresses due to interference between the
, bundle and the tube support plates (" bear hug") which, in combination with cyclic radial temperature gradient, could leak to fatigue failures.
Three cases were examined for evidence of dynamic instability: 18.6%,
15.8" and 14% feedwater flow rate. Due to computer code limitati.ns, plant operating conditions at 14% feedwater were estimated from the data at the two higher flow rates. The steam generators were shown to be stable, by use of the dynamic stability code LOOP, down to 14%
feedwater flow. Nevertheless, the effect on the steam generators of dynamic instability was estimated, assuming that the steam generators were actually unstable below about 20% feedwater flow.
l The maximum temperatur variation of the tubes, occurring in the evapo- !
rator section, would result from fluctuation in the boiling regime from nucleate to film-type, regy} ting from tha cyclic variation in mass flow. 1 Generalized correlations show that, for the operating conditions of ,
interest, this could result in a maximum fluctuation of about 36*F in tube wall temperature. Test data on both the General Atomic and the French boiling stability test loops indicate that maximum mean wall l temperature swings will be on the order to 25 - 40*F. There also would !
be a 20*F variation in tube wall radial temperature difference. These {
conditions would not change the normal operating stress range and the tubes could thus withstand an essentially infinite number of such cycles without failing due to fatigue. Accordingly, it is concluded that dy-namic instability of the steam generators at low feedwater flow rates could not cause damage, even if an instability occurs.
(1) .M.
Lasarev, "The Effect of An Oscillating Dryout Point on Evaporator Tube Lifetime," Trans. ANS, 1972 Annual Meeting, June 18-22, 1972, pp. 396.
1
o Attachment 6 Page 7 of 14 Conclusions A reduction in the setpoint of the fixed feedwater flow - low Plant Protective System trip from the present Technical Specification value of 20% to 10% and retention of the present five-second delay does not jeopardize the operation of the steam generators in the event of sudden loss of feedwater flow. Time to reach steam generator maximum allowable tube temperatures under the postu-laced accident conditions all exceed the five-second Plant Protective System delay, the circulator speed valve closure, and the time required for stall or coast down of the associated circulators.
Unstable flow conditions in the steam generator, should they occur, would not damage the steam generator even if they should persist for long periods of time. However, an alarm will be provided at 20% feedwater flow to alert the operator that the plant is operating in an undesirable region.
The incorporation of the proposed. change will remove a major source of unnec-essary loop trips which have occurred during plant transients and thus i=-
prove plant reliabiilty.
4 TABI.E 1 ,
f Initial Tube Henn Maximum Wall Temperature Allowable Transient Max. Delay Cire. Shutdown Allowable PPS i gg) Delay, Sec.
location " Nominal" "llottest" Temp. AT, AT/Sec. Time, Sec. Time, Sec.
100% LOAD.
Sil I 1.cadout 755 862 1975 213 14,8 14.4 1.8 12.6 SH I Dutlet 865 979 1075 96 11.4 8.4 1.8 6.6 s' ;
" i Outlet 1155 1223 1300(2) 77 6.0 12.8 1.8 11.0 !
1,00% LOAD Sil I Leadout 870 959 1085 126 2,8 45.0 2.2 42.8 Sil I Outlet 920 992 1085 93 2 46,5 2.2 44.3 1
utlet 1100 1152 1300(2) 148 0,8 185 2.2 182.8 ffk
", aif n
(1) Sil II bundle temperature transient is less severe than the RII tubes and thus does not limit allowalle **
delay time. ,
(2) ASHE code does not specify material properties above 1300*F (design temperature) for Inconel 800.
1 I
TAllLE 2 :
Initial Tube Mean Maximur I Wall Temp., *F Allowable Equivalent !
" Nominal" Max. Delay Cire. Shutdown Allowable PPS I.ocation gy) " Nominal" "llottest" Temp. Tube Temp. Time, Sec. Time, Sec. Delar, Sec.
AT_ _
i 5 11 I Imadout 785 892 1075 183 968 49 13.5 35.5 .
511 I. i Outlet 880 994 1075 81 961 32 13.5 18.5 l t
Ril outlet 1135 1203 1300(2) 97 1232 43 13.5 29.5 i
f (1) Sil II bundle temperature transient is less severe than the Ril tubes and thus does not limit m>
allowable delay time.
((n (2) ASHE Code does not specify material properties above 1300*F (design temperature) for Inconel *d 800. $
~
$= o%
Attache.ent 6 Page 10 of 14 TABLE 3 PREDICTED PLANT OPERATING CONDITIONS DC-1-4 June - July, 1977 Januarv 30, 1970 Power 29.3% 20% 17% 28.4%
Feedwater Flow 26.1% 18.6% 15.8% 25%
~
Helium Flow 35.1% 32.1% 26.6% 28.6%
Helium Temperature at Reheater 1134 1030 1030 1236 Helium Temperature at SH II 1026 957 954 1136 Helium Pressure, psia 625 626 622 590 MS Temper.ture at SH II Outlet 962 942 938 1000 MS Temperature at T-G 880 850 .835 1000 RH Temperature at Reheater Outlet 972 920 924 1000 RH Temperature at T-G 930 900 900 1000 Economizer Inlet Pressure 2443 2432 2426 2440 Tube Pressure Stress, psi 1818 1806 1804 1850
O Attachment 6 Page 11 of 14 TABLE 4 MAXIMUM ALLOWABLE TUBE TEMPERATURES AT 18.6" FEEDWATER FLOW Economizer Lead-in Tubes 952*F j Economizer Initial Length 952*F
- Economizer Final Length 962*F ,
j Evaporator 980*F SH I 1080*F j SH II >1200*F i
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- Attachment 6 1 Page 12 of 14 i
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ATTACDINT] ,
SUPPORTIVE INFORMATION FOR EVALUATION OF LOSS OF CIRCULATOR BEARING WATER - PPS SETPOINT f
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Attachment 7 Pagr. 1 of 7 EVALUATION OF LOSS OF CIRCULATOR BEARING WATER l
- PPS SETPOINT INTRODUCTION The helium circulators have water lubricated bearings. Recirculating water is supplied to each circulator bearing set at about 170 gpm and at a pressure about 640 psi above primary coolant pressure. Each cir- ,
{ culator bearing set includes two journal bearings, a main thrust bear- !
ing and a reverse thrust bearing. The recirculating water is normally supplied'by 2 of 3 (1 standby) pumps and is referred to as the nor al bearing water (NBW). A backup source is available from the fee 6.ater
. system and is referred to as backup bearing water (SUSW). Given the sudden loss of bearing water from both of the above two sources, a third supply is available for safe shutdown of the circulator. This safe shut-down supply consists of a gas pressurizer and water accumulator capable of supplying bearing water for at least 30 seconds at design flow rate.
This is adequate for safe shutdown of the affected circulator.
T"ne setpoint at which the accumulator is fired given a sudden loss of bearing water is the subject of this evaluation. The setpoint needs to be sufficiently high to ensure' the bearings are not damaged and yet low enough not to cause unnecessary circulator trips during plant tran-sients, including transfers frem NSW to SUSW. Both an instrument trip setting and an " absolute" value will be established. The difference between the instrument trio setting and the absolute value must be at least equal to the instrument calibration error and drift.
DISCUSSION l This anal'ysis investigates the safe shutdown of a circulator, with the l assumed sudden loss of NSW and SUSW. Of interest is the absolute value
! at which the accumulator could be fired pennitting the circulator to be i shut down without damage. Firing of the accumulator above the absolute value would result in additional margin against circulator damage.
The analysis will address normal operation clearances within the water lubricated bearing and the minimum clearances to preclude damage during shutdown. Data will be provided for the bearing water pressure history
, follo' wing a sudden loss of normal bearing water and firing of the ac-l cumulator for safe shutdcwn. The consequences of the pressure history with the firing not occurring until the absolute value of the setpoint is recched will be addressed regarding bearing clearances. Finally, the acceptability of a 475 psid instrument trip setting with an absolute PPS trip value of >450 psid will be addressed.
Attacht::ent 7 Paga 2 of 7 BEARING CLEARANCES Normal operating clearances for the circulator bearings are:
Turbine Jou.nal Bearing (centered) 0.0025 in.
Compressor sournal Bearing (centered) 0.0035 in.
Main Thrust Bearing (centered) 0.00a5 in.
Reverse Thrust Bearing (centered) 0.0045 in.
A clearance of 0.001 in is conservatively selected as the minimum clearance to assure adequate lubrication during ' shutdown of a cir-culator frcm 100% speed.
CIRCULATOR SHUTDOWN TESTS Bearing pressure dynamics during circulator shutdown were measured as a part of the Reference 1 tests. During these tests, the NEW and SUBW supplies were terminated and an accumulator was fired at the 475 psid setpoint. Folicwing accumulator firing, mcmentary dips in becring cartridge differential pressure recorded on Brush recorders during the transfer to the accumulator water supply as shown in Figure 1 and 2.
Minimum pressures observed during these dips are:
Circulator Minimum pressure (asid)
A 405 8 4.05 !
C 375 0 378 For 'C' and 'D' circulators, these dips in aP occurred within 0.5 I seconds after firing of the. accumulator and recovered within one second i to above 400 psid and within 4 seconds to 450 psid.
From these data, i.t follows that if the accumulators were fired at 450 psid instead of 475 psid, the mcmentary pressure dips would also be 25 psi less, or a minimum of 350 psid for the 'C' circulator.
JOURNAL BEARINGS l Each circulator is equipped with two journal bearings. The purpose of the bearings is to center the shaft in the housing. Lead on the jour-nal bearings is solely a function of the imbalance on the rotor. prior l to assembly of the FSV circulators, the rotor is balanced so that the residual static imbalance and the dynamically imposed imbalance due to the coupling moments during rotation is less than 0.2 ounce inches in
! any plane.' This gives a displacement of 0.00004 in. a. 10,800 rpm with a bearing water flow generated AP of 700 psid across the bearing housing. Displacemer.t increases with increasing circulator speed.
Attachment 7 Pags 3 of 7 Since the stiffness of the bearings is generatd by the pressure dif-ferential across .them, as the pressure drops, the stiffness decreases and therefore, for the same imbalance load, the shaft displacement will increase. Thus,.at 350 psid, the shaft dis:lacement will only be 0.00008 in. at 10,800 rpm and imbalance of 0.2 ou.ce inches. Thus, it can be seen that the minimum AP that may occur during a sudden loss of beaHng water incident has practically nc effect on the journal bearing oparating clearances.
Further evidence as to the large margin of safety a the journal bearings was demonstrated in tests perfomed as part of Reference 1. For these '
tests, the circulators were shut down with accumulator water ficw generated pressure drops across the bearings of only 50 psid! This pressure drop is considered adequate to shut dcwn the circulator from 8000 RPM.
THRUST BEARINGS Each circulator is also equipped with two thrust bearings: a main thrust bearing, and a reverse thrust beari.ng. Extensive design margin in these thrust bearings occurs at the design operating speed where the generated thrust is -2000 lbs (i.e., the thrust load is carried by the reverse thrust bearing). The thrust goes through zero at 90% speed and up to a running maximum of +S000 lbs on the main thrust bearing at 30% speed.
(See Figure 3.)
The maximum load on the reverse thrust bearing would occur during a rapid pCRV depressurization event while the circulator is at 100% speed.
During this event, the cperating reverse thrust load would increase I from 2000 lbs to 2900 lbs in conjunction with the helium pressure decay while the reheat steam pressure in the circulator turbine would remain at operating pressure. For this reverse thrust load, a bearing water pressure of 263 psid is required to maintain a clearance of 0.001 in.
on the reverse thrust bearing.
The thrust load on the main thrust bearing as described above and illus-trated in Figure 3 varies with circulator speed being about 8000 lbs at 30% of rated speed. This bearing, hcwever, was designed to accept a maximum .hrust load 11,400 lbs. Assuming a minimum 350 psid bearing water pressure as discussed under " Circulator Shutdown Tests *, the minimum clearance of 0.001 is maintained with the maximum thrust load (11,400 lbs) at normal circulator cperating speeds. The clearance improves with increasing circulator speed as shown in the attached Table.
Circulator Speed Running Clearance for 350 psid -
(RPM) Bearino aP and 11,400 Thrust Load !
l 2,000 0.0010 4,000 0.0011 6,000 0.0012 8,000 0.0014 10,000 0.0018
Attachmnt 7 Pag's 4 of 7 There is one multiple failure case where the running clearanc'e on the main thrust bearing is reduced slightly below 0.001 in. This case is an offset steam. generator tube rupture with wrong loop dump producing high PCRV pressure but belcw the PCRV relief valves setpoint. A circulator in the shutdown loop is self-turbining at 300 rpm with atmospheric steam pressure in the downstream piping of the circulator steam turbine. Under
, the above conditions, the maximum thrust load of 11,400 lbs is experienced. l With sudden loss of bearing water, the rotor is stopped by application of the brake within 6 to 10 seconds. Assuming the minimum bearing pressure of 350 psid, the running clearance is reduced during the 6 to 10 second I shutdown time-to 0.0009 in. This would not cause any damage to the i main thrust bearing. In fact, the original (first prototype) circulator I tested at Valmont had no brakes and was normally stopped by reducing the bearing water flow and allowing the thrust runner to rub on the bearing surface. No damage occurred to the thrust runner as a result of this shutdown method. l Thus, at any speed above self-turbining, the shaft clearance will be maintained over the censervative 0.001 in. value during a circulator shutdown.
INSTRUMENTATION ACCURACY The bearing water pressure above primary coolant pressure is menitored by Barton differential pressure switches. There are three Earten instruments per circulator for a total of 12 instruments. The range' of these Barton instruments is 600 psid and are set to read between 50 - 650 psid. The manufacturers published setpoint accuracy is 2% or 212 psid while the repeatability is 20.2% or 1.2 psid.
Calibration.is specified in the Fort St. Vrain Technical Specification and is acccmplished by applying a kncwn pressure (monitored with a pressure gauge) and recording and adjusting as required the trip setting. One set of calibration data on the twelve instruments covering an interval of 14 months is available. The differences between the as-left and as-found reading were all positive with the average change being a +7.3 psi with the maximum change being '13 psi. These readings would include instrument drift as wel1 as setpoint accuracy and repeatability. Based upon these data, it is concluded that the instrument perfannance is consistant with the manufacturers published accuracy data.
RECOMMENDATION Based on the above analysis, it is recon = ended that the PPS " Circulator Bearing Water t.ow AP" instrument setpoint should be 475 +10, -0 psid with an absolute value of > 450 psid. The 25 psi differential is sufficient to acccamodate instruEent calibration error and drift.
, REFERENCES
- 1) . General Atomic Request for Test (RT-3653) to Cemonstrate by test that the Time , Delay Settings and valve pre-position programs for the emergency accumulator ficw control valves provide adequate
- circulator emergency shutdown protection at all anticipated operational conditions.
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i ATTACHMENT 8 l
3 SUPPORTIVE INFORMATION ON APPLICABILITY OF ABSOLUTE TRIP VALUES FOR 5% AND 30% ROD 'JITEDRAWAL PROHI3ITS IN THE s FORT ST. VRAIN TECHNICAL SPECIFICATION 9
i 0
4 4
F 0
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l Attachment 8 l . Pcgs 1 of 4 SUPPORTIVE INFORMATION ON APPLICABILITY OF ABSOLUTE TRIP VALUES FOR S AND 30". R00 WITH0RAWAL PROHIBITS IN THE FORT ST. VRAIN TECHNICAL SPECIFIrATION INTRODUCTION The purpose of the evaluation presented herein is to investigate the principle' function of the 5% and 30% power rod withdrawal prohibits (RWPs) listed in the Fort St. Vrain Technical Specification LC0 4.4.1, Table 4.4-4, Items Za, 2b, 3a and 3b. The evaluation will be directed at the applicability of specifying " Absolute Value" in the technical specifications for these RWPs.
DISCUSSION -
The Fort St. Vrain plant protective system (PPS) employs multiple lines of defense for reactivity excursion accidents particularly those that could be produced by a rod withdrawal accident. The rod withdrawal prohibits (RWPs) are discussed in FSAR Section 7.1.2.2 and illustrated in FSAR Figures 7.2-5 and 7.3-1. -
Some plant parameters monitored by the plant protective system, primarily those of the primary and secondarj coolant systems, must be bypassed until the parameter reaches nonnal power operation levels, which is about 30% power. Thus there is a need to bypass certain plant protective system actions during startup and then switch them in when the appropriate power levels are reached. At For'. St. Vrain, protective action bypasses are acccmplished with the Reactor Mode switch and the Interlock Sequence Switch as discussed in FSAR Section 7.1.2.8. Of interest regarding the 5% power RWP and 30% power RWP during a plant startup is the Interlock Sequence Switch (I.S.S.). Dual requirements must be satisfied to accomplish a rise-to-power. These dual requirements are the I.S.S. position applicable power range is consistent with the actual power level as measured on power channels III - VIII, and the pemissable arrangement of power channels III - VIII exists. FSAR Table 7.1-7 (attached) l provides a tabulation of system actions should the I.S.S. be in the I wrong position for the planned operation. l During a rise-to-power from a shutdown condition the operator places the I.S.S. in the STARTUP position. Tnis actuates startup channels I and II, which monitor neutron flux at very low power levels (FSAR Figure 7.2-5). If there is inadequate flux reading ($ 2.5 cps) then a RWP occurs preventing startup. Also wide range log and linear channels i III, IV and V and linear channels, VI, VII and VIII, which are never i bypassed, are in service. Channels III, IV,and V will result in a RWP !
and finally a scram action if the rate of neutron flux change is excessive. '
Startup is continued in this mode up to about 5% power at which time administrative procedures require the operator to advance the I.S.S. to the LOW POWER position. Should the operator attempt higher power operation without advancing the I.S.S., this would be precluded by a 5% i power RWP and a.larm. The annunciator window with the caption Incorrect l Interlock Sequence would be lighted to remind the operator to change l the switch position. Should the operator attempt startup with the
[ ___ . _ _ __ __
Attcchment 8
- Pcgs 2 of 4 i~ I.S.S. in the LOW POWER position, this would be precluded by the 5%
power RWP plus the above discussed alam and annunciation. If the operator attempts startup with the I.S.S. in the POWER POSITION, then an RWP is activated by linear channels III, IV, V, VI, VII and VIII being less than 30% power Again the above alam and annunciation would 1 be provided as a reminder to the operator. In addition, scram action
, would occur due to primary and secondary coolant parameters which are
- activated by placing the I.S.S. in the POWER position not being at the required operational levels.
After about 5% power has been attained, the operator advances the
- I.S.S. to the LOW POWER position. Plant operation is now permitted j between about 5% and 30% power and monitored by the 5% RWP and 30% RWP.
- Attempteo operation above 30% power is precluded by the 30% RWP while
] operation below 5% is alarmed with annunciation of Incorrect Interlock j Sequence Window.
At about 30% power, when primary and secondary coolant parameters have
! attained nomal power production levels, aditional plant protective
{
system actions are made operational, and the 30% RWP is removed by j advancing the I.S.S. to the POWER position. Specifically, these plant protective systems are primary coolant pressure-low, (programmed with j circulator inlet temperature), hot reheat header pressure-low, main j steam pressure-low, super-heat header temperature-low, circulator i speed-low (programmed with feedwater flow), fixed feedwater flow-low, 1 main feedwater pressure-low (programmed with circulator speed), main feedwater pressure-low, and emergency feedwater flow-low.
4 During the previously described rise-to-power sequence, a 120% power RWP and a 140% power scram have been operational. These are never bypassed by the I.S.S.
In an orderly plant shutdown, the sequence works generally in reverse to that described for startup. The I.S.S. is moved from the POWER position at about 30% and from the LOW POWER to the STARTUP position at 1 about 5% power. The alam and Incorrect Interlock Sequence window annunciation actuated by the 5% and 30% RWPs will serve as a reminder to the operator to follow the administrative procedures.
In Fort St. Vrain FSAR Chapter XIV, Section 14.2.2, analyses of rod withdrawal accidents both at source power and at full power are . presented.
- The many lines of defense employed in the plant design to protect against rod withdrawal accidents are discussed. One of the lines of defense for startup is the 5% and 30% RWPs if the Interlock Sequence 3
Switch is in the wrong position. However, the od withdrawal accident explicitly examined during startup assumed all other defense mechanisms fail and that the accident is terminated by the 140% power scram. For a rod withdrawal accident at full power, three methods of temination are examined. These are a 140% power scram, operator manual scram of 60 seconds after initiation of the event, and 105 seconds after initiation by the 1075'F reheat steam temperature scram. In all cases the consequences of rod withdrawal accidents were judged to be acceptable. Therefore, the 5% and 30% RWPs, while they could terminate a red withdrawal accident under some conditions, are not required to mitigate the consequences of these types of accidents.
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J Attachment 0
- Pago 3 of 4 The primary purpose of 5% and 30% RWPs is to assure operator compliance with administrative precedures to maintain the Interlock Sequence Switch in the proper position consistent with the actual power level.
This stated purpose is repeated in FSAR Section 7.1.2.2, Section 14.2.2.1 and in the Basis for Fort St. Vrain Technical Specifications L.C.0.
4.4.1. This is consistent with no credit assumed for thesa RWPs in the accident analyses as discussed above.
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[_ . - - . . _ _ , ~ , _ ~ . . _ _ . - _ _ _ _ . , , , . _ . . _ . . . _ . _ . . . . . _ . _ . . _ . , _ _ _ . . _ _ . , . , , _ _ , _ _ , _ _ . _ _ _ , , . _ _ , _ , _ _ _ _ _ _
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Attechment 8
, Pags 4 of 4 CONCLUSION AND RECOMME!:0ATIONS The Interlock Sequence Switch activates or bypasses selected plant protective system actions consistent with pcwer level during a rise-to-power or in an orderly shutdown. The~ function of the 5% and 30% RWPs is to ensure that the operator follows administrative procedures of maintaining the Interlock Sequence Switch in the proper position consistent with the measured pcwer level. In accomplishing this function both the 5% and 30% RWPs are in actuality a histable in which the measured power must be either below or above the bistable trip setting consistent with I.S.S. position (See Discussion). Therefore the current use of > 5% and 5 30% RWP describes only a portion of the function performed by these RWPs in ensuring ccepliance by the operator with administra-tive procedures. It is therefore reconnended that the instrument trip setting be revised to delete the 2 and 5 and be replaced with a setpoint with p.lus and minus tolerances.-
Also since these RWPs are not required to mitigate the consequences of red withdrawal accidents nor was credit assumed for these RWPs in Chapter XIV of the Fort St. Vrain FSAR, " Absolute Trip Value", in the proposed format change to Technical Specification LCO 4.4.1 is not applicable (N.A.).
Ihe reccz. ended revisions to Fort St. Vrain Technical Specification LCO 4.4.1, Table 4.t-4 for the Absoluta Value and the Instrument Trip Setting are as follows: ,
Instrument Absolute Trip No. Functional Unit Value Setting 2a Linear Channel - Lcw Pcwer RWP N.A. 5 + 2, - 2%
(Channels 3, 4 and 5) 2b Linear' Channel - Lcw Pcwer RWP N.A. ' 5 + 2. - 2%
(Channels 6,7and8) 3a Linear Channel - High Power RWP N.A. 30 + 2. - 2%
(Channels 3, 4 and 5) 3b Liriear Channel - High Power RWP N.A.
- 30 + 2. - 2%
(Channels 6, 7 and 8)
The balarice of LCO 4.4.1, Table 4.4-4 remain applicable including the notes specified therein.
It is racemended the "Sasis for Specification LCO 4.4.1" item d) Linear Channel (30% Power)" be revised as follows:
Linear Chann11 (30% Power) is 'provided to prevent control rod
,' withdrawal if reactor power exceeds the limit for the "Lcw Powera position or imprcper positioning of the I'.S.S.
s mW T 3 (
f l
The remai'nder of the "Sasis" item d) remain as is. ,
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