ML20235V887

Draft Plant Protective Sys Trip Setpoints for Fort St Vrain Nuclear Generating Station, Technical Evaluation Rept

ML20235V887
Person / Time
Site:	Fort Saint Vrain
Issue date:	10/31/1987
From:	Stachew J EG&G IDAHO, INC., IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY
To:	NRC
Shared Package
ML20235V868	List: ML20235V865 ML20235V887
References
CON-FIN-D-6023 EGG-NTA-7289-01, EGG-NTA-7289-1, EGG-NTA-7289-DRF, TAC-47416, NUDOCS 8710150368
Download: ML20235V887 (51)

Text

__ _.

,, ; 4 0

3. .= y l

D T EGG -NTA-7289:

L i r-  !

l ,-

TECHNICAL $ VALUATION REPORT.~FOR.THE 3

PLANT PROTECTIVE SYSTEM TRIP SETPOINTS-FOR FORT.ST. VRAIN NUCLEAR GENERATING STATION.

t h 'J. C. Stachew Published October 1987 Idaho National Engineering Laboratory EG&G Idaho, Inc.

Idaho Falls, ID 83415 i

l Prepared for the .

U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Under DOE Contract No. DE-AC07-761001570 FIN No. 06023 P

pgogg[g e

-.7

'o4 s; '

-ABSTRACT-LTnis EG&G Idaho, Inc., report evaluates submittals provided by Public Service Company t of' Colorado.for:the Fort St. Vrain Nuclear Generating ll Station. The'submittals are in response to requests that the trip l setpoints specified in the Tech'nical Specifications should account for

. instrumentation uncertainties. 1

)

1

< FOREWORD This report is supplied as 'part of the " Technical Assistance for

' Operating. Reactors Licensing Actions," being conducted for the'U.S. Nuclear Regulatory Commission', Washington D.C. , by EG&G Idaho, Inc., NRC Technical Assist'ance.

The U.S. Nuclear Regulatory Commission funded the work under 00E' contract No'. DE-AC07-761001570 FIN-No. 06023.

i i

I Docket No 50-267 TAC No. 47416 f

ii 1 l'

c' ,

M i

CONTENTS i l

iii AS S T RA C T . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . .

1 iii 20 REWORD ..............................................................

1. INTRODUCTION ............................... .................... 1

2. DISCUSSION AND EVALUATION .............. ............ ............ 3 3

2.1 Methodology .................................. .............

2.2 Evaluation of Reanalyzed Trip Setpoints ........ ...... 3 2.2.1 Primary Coolant Pressure-Low ........... . ......... 3  ;

2.2.2 Primary Coolant Pressure-High ...................... 5 l 2.2.3 Superheat Header Temperature-Low ........... ....... 7 2.2.4 Circulator Speed-Low ............................... 8 2.2.5 Fi xed Feedwater Fl ow-Low . . . . . . . . . . . . . . . . . . . . . . . . . . 9 2.2.6 Loss of Ci rculator Bearing Water . . . . . . . . . . . . . . . . . . 10 11 2.2.7 Circulator Speed-High ........................... ..

2.2.8 N e u t ro n Fl u x-H i g h . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 2.3 Evaluation of Proposed Technical Specification Changes ..... 13 2.3.1 Limiting Safety System Settings (LSSS) 13 (Section 3.3) .................................... ,

2.3.2 Protection System Instrumentation, Limiting ]

Conditions for Operation (LCOs) (Section 4.4.1) . . . . 19 l l

l

3. REMAINING ISSUES ASSOCIATED WITH THE PLANT PROTECTIVE SYSTEM 34 INSTRUMENTATION ................................ .................

Circulator Trip on Programmed Feedwater Flow-Low . 34 3.1 ...... .

Circulator Trip on Fi xed Feedwater Flow-Low . . . . . . . . . . . . . . . . 35 3.2 Rod Withdrawal Prohibit at 30% Rated Thermal Power ......... 35 3.3 36 3.4 PPS Permissible Bypass Conditions .......................... .

1 Technical Specification Upgrade Program Related Changes . .. 36 3.5 37 3.5.1 TS Section 2.0, Definitions ..... ..... .. ... . .

2.5.2 T5 Section 4.0, Limiting Conditions for Operation .. . .. .... .. .. .... . .. .. . ... .. 38 l TS Section 5.0, Surveillance Requirements . 38 3.5.3 . ......

iii i

l I

1

[ ._ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _

e ,: ;r ,

y c

73;V '3.5.4 TS Section 4.4.1, Plant Protective System .

Instrumentation LCOs ................................. 38:  !

TS-Section 5.4.1, Plant Protective 1 System 3.5.5 Instrumentation Surveillance and Calibration Requirements .......................................- 39 4 CONCLUSIONS ...........',.......................................... 41 Proposed Changes Judged Acceptable . . . . . 42 4.1l . . . . . . . . . . . . . . . . . . .

14.2. Proposed Changes Judged Not Acceptable ..................... 43 4.3 Other! Recommended Changes as -a. Resul t of Thi s ' Rev'f ew . . . . . . .. 44 i

4 '. 4 ' Remaining PPS Issues ........................................ 1

.5. REFERENCES ....................................................... 47 I

.)

I l

s ,?

iv i

.o-i TECHNICAL EVALUATION REPORT FOR THE PLANT PROTECTIVE. SYSTEM TRIP SETPOINTS ,

1

- FOR FORT ST. VRAIN NUCLEAR GENERATING STATION l l

I

1. INTRODUCTION I l

By letters dated June 21, 1985,1 May 15, 1986,2 and August 28, 3

1987 :the'Public Service Company of Colorado (PSC) proposed numerous changes.to the_ Technical Specification's (TS) for the Fort St. Vrain (FSV)

Nuclear Generating Station. The primary purpose of the proposed changes was to. modify the trip setpoints for the Plant Protective System (PPS) such that the, values specified included a sufficient. allowance for uncertainties

' associated with'the_ instrument' systems. Currently,'the setpoints for the PPS are specified at the same values .for which the' safety analyses assumed mitigative actions would be initiated. The proposed changes result in I revised trip setpoints that . include an additional margin of conservatism to account for. instrumentation uncertainties. The revised trip setpoints were determined using as guidance Instrument Society of American Standard 567.04-1982,4 '"Setpoints 'for Nuclear Safety-Related Instrumentation Used l in Nuclear Power Plants."

As a result of the Licensee's evaluation program to determine  ;

appropriate' values for instrumentation trip setpoints, the values for some ,

-trip functions were found to offer the potential for increased inadvertent scrams, loop shutdowns, or circulator trips. In these cases, the results  !

of a ' reanalysis were provided to justify the use of trip setpoints that  !

provide a greater margin between the trip setpoint value and normal operating conditions.

This Technical Evaluation Report provides an evaluation of the proposed trip 'setpoints and the reanalysis provided to reduce potential for inadvertent safety actions, as transmitted in PSC's revised letter of  !

August 28. 1987 and as supplemented by the earlier PSC submittals. The earlier PSC submittals were responded to by NRC letters dated January 24, 19865 October 16, 19866 and November 26, 1986.7 The NRC letter of 1

i' _ _ - - _ - - - - . - - - _ _ _ _ _ _ _ _

January. 24, 1986 recommended that the Technical Specifications for the trip setpoint reanalysis to account for instrumentation inaccuracy be separated

.from the' format upgrade issues. The NRC letters of October 16, 1986'and November 26,-1986 responded to.the PSC submittal that made the requested separation.(PSC letter of May 15,1986). These latter NRC letters were

~

requests ~ for additional information .or guidance to clarify seventeen . issues in the Licensee's-May 15, 1986 letter. The present PSC;1etter of August 28,1987. continues to rely on information presented in the earlier l PSC' letter of June 21, 1985.1 Many PPS functions presented in the PSC June 21, 1985 letter were deleted in the latest August 28, 1987 submittal.

Again, this was-per NRC direction.to focus attention on only those PPS functions that are currently in the existing FSV Technical Specifications.

Finally, it is emphasized that the NRC evaluation of January 24, 1986, on the reanalyzed trip setpoints that were made to justify the use of a greater margin between the trip setpoint and normal operating conditions.,

has been relied upon and has essentially been duplicated here in this  ;

report. No independent evaluation was made related to which setpoint changes required additional safety analyses or the correctness of such added safety analysis. Only an update was made to' bring the NRC discussion in the January 24, 1986 submittal current with Rev. 5 to the Fort St. Vrain FSAR.

l Evaluation of the Licensee's justification for change was based primarily on review against the Fort St. Vrain FSAR, Rev. 5, ISA S67.04-1982,4 the Westinghouse STS,8 the NRC Staf f draft Safety Evaluation Report (SER) in letter dated January 24, 1986,5 and other Licensee supplied documentation (PSC letters of March 9, 1984,9 June 21, 1985,1 May 15, 1986,2 and August 28, 19873 ,

l

\

2

d .,: , l

2. DISCUSSION'AND EVALUATION l:

l 2.1 Methodology The Licensee submittal'of. August 28, 1987'made proposed changes to .;

p Technical Specification Section 3.3, Limiting Safety System Settings, and 4.4.1, Plant Protective' System Instrumentation. These proposed changes were basically.to account for instrumentation inaccuracy in establishing the Trip Setpoints for the scram, loop shutdown, and circulator trip functions. In addition to the previously specified "as left Trip Setpoint"

(' 'an "as found Allowable Value" limit is also specified. The as found Allowable Value limit-is chosen to ensure that the analysis value used in the safety analysis to initiate the trip actions is not exceeded. The analysis value is that trip value used in the safety analysis which demonstrates the associated safety limit will not be exceeded or that 9

equipment protection is' assured. By letter dated March 9, 1984 the Licensee provided a copy'of a specification outlining the reevaluation of-the Plant Protective System setpoints' to account for instrumentation <

inaccuracy. This Licensee document incorporates the requirements of ISA Standard S67.04-1982, for establishing trip setpoint values. Therefore, the Lice'nsee has established a methodology which is acceptable for determining Trip Setpoints and Allowable Values based on safety analyses for the Fort St. Vrain Nuclear Generating Station as documented in the FSAR.

2.2 Evaluation of Reanalyzed Trip Setpoints Attachment 3 to the Licensee's letter. of June 21, 1985 provided a Signif_icant Hazards Consideration Analysis that addressed the results of new analyses for selected safety functions. The conclusions of this ,

analysis was previously evaluated by the NRC Staff in Reference 5 and has been updated here to be current with FSAR Rev. 5.

I 2.2.1 Primary Coolant Pressure - Low k

The present setpoint for the low primary coolant pressure scram is j programmed with load (circulator inlet temperature) to initiate scram when I 3 J

a=

. , 3

• l Jt reactor coolant pressure is 50 psi below normal. The low primary coolant I

pressure scram provides protection for inadequate core cooling that could result in temperature limits being exceeded. For rapid cepressurization accidents, a scram would occur instantaneously, and changes in the low pressure setpoint would not have an impact on the consequences of the 1

accident. 1 Two cases were reanalyzed based on the assumption that a scram occurs at a pressure of 90 psi below normal. The first case reanalyzed was the offset rupture of a 2-inch line in the helium purification regeneration piping, as currently analyzed in FSAR Sections 4.3.3 and 14.8. For this accident, which is assumed to occur at 100% power, and as currently analyzed a scram occurs at 50 psi below normal pressure in about 120 sec, primary coolant flow is 97% of rated, and the peak core average outlet temperature is 13 F above normal. Under the reanalysis assumption that a {

scram does not occur until primary coolant pressure is 9 psi below normal, i primary coolant flow will have been reduced to 92.5% of rated in 220 sec.

And the core average outlet-temperature peaks at 44 F ab::ve normal. After j the reactor scram, core average outlet temperature decreases with continued f core cooling.

The second caso reana!yzed was the effect of continued plant operation at both 100% and at 25% power with reduced primary coolant pressure just above the assumed scram value of 90 psi below normal. For these two conditions, circulator speed increases in response to the decreased helium I

inventory; however the core power-to-flow ratio only changes by 0.01 at both 25 and 100% power. The impact on helium temperature at the inlet to the steam generators is an increase of 9 F at 100% power and 2 F at 25%

power.

It was concluded that, since neither a safety limit nor an equipment design limit is exceeded, the assumption of a lower primary coolant pressure for initiation of a reactor scram is acceptable. (

l l

4 l

l

DRAFT Based on the review of these results,.it is concluded that this analysis provides an acceptable basis- to justify a lower trip setpoint for this safety function. With the allowance for instrument uncertainty the new trip setpoint is 68,6 psi belw normal primary coolant pressure.

2.2.2 Primary Coolant Pressure - High The present setpoint for.the high primary coolant pressure scram is ,

programmed with load (circulator inlet temperature) to initiate a scram when the reactor coolant pressure is 7.5% (approximately 53 psi) above normal. The high primary coolant pressure scram and preselected steam generator dump are a backup for the primary coolant moisture monitor scram 1 and dump of a leaking steam generator. The FSAR Section 14.5.3. safety analyses address six accident cases related to steam ingress with various postulated failures of the protection system. Of the six accident cases analyzed, only four involve safety actions initiated on high primary coolant pressure. Each case was reanalyzed as follows based on the assumption of a high pressure scram at 70 psi' above normal.

1. FSAR 14.5.3.2 Case 2 - Subheader Rupture and Wrong Loop Dump. It is assumed that the moisture monitors initiate a scram; however the wrong loop is dumped. The only safety action initiated on high pressure is the initiation of the steam generator depressurization program which reduces steam ingress by lowering steam generator pressure. The current analysis indicates that the safety action is initiated after about 80 sec, with a total steam ingress of 14,890 lb of which 180 lb react with core graphite. With the assumption of a higher pressure trip (70 psi j above normal) the depressurization program is initiated at 120 see with a total steam ingress of 15,000 lb and there is no change in the amount that reacts with core graphite.

2. FSAR 14.5.3.4 Case 4 - Subheader Rupture with Moisture Monitor Failure and Correct Loop Dump. It is assumed that no safety actions are initiated by the moisture monitors. On high primary I'

l' i

1 5

L__=_

4 t ,

coolant pressure, a reactor' scram is initiated, and the preselected loop dump isolates the leaking steam generator. The

. current-analysis indicates that there is a scram and steam generator dump in 95 sec, with a total steam ingress of 2,160 lb l

of which 855.lb react with core graphite. With the assumption of.

I a higher pressure trip (70 psi above normal) safety action is initiated in 157 sec with a total steam ingress of 3,200 lb of which 1,112 lb react with core graphite.

3. FSAR 14.5.3.4 Case 5 - Subheader Rupture with Moisture Monitor Failure and Wrong Loop Dump. This case is the same as (2) above; however, it is assumed 'that the intact loop is dumped. The current analysis indicates a total steam ingress of 16,040 lb of

'which 900 lb react with core graphite. With the assumption of a higher pressure trip, the total steam ingress is 15,600 lb of which 1,162 lb react with core graphite.

Although the reanalysis shows 'a lower total steam ingress, it.was noted that the original analysis was conservative since it assumed that the leakage was terminated 30 min after the time a scram was initiated, rather than 30 min after the time of the accident.

4. FSAR 14.5.3.4 Case 6 - Subheader Rupture with Moisture Monitor Failure, Correct Loop Isolation and Failure to Dump. This case ,

is the same as (2) above; however, it is assumed that the f aulty steam generator is isolated only, not dumped. Thus, the only difference between this case and case (2) is that the entire 6,000 lb inventory of the steam generator is assumed to enter the primary coolant system. In the current analysis, the total steam ingress is 8,080 lb of which 919 lb react with core graphite.

With the assumption of a higher value for the high pressure trip, the total steam ingress is 9,200 lb of which 1,200 lb reacts with core graphite.

l 6

DRF The overall impact of. the change from 53 psi to 70 psi above normal

-for the. high primary coolant pressure trip is an increase of'about 30% in the amount of moisture that reacts with core' graphite in those cases for wb Mh multiple failures of the protective system are assumed. While the impact of increased steam / graphite reaction was not specifically cnalyaid, the present' analysis of steam graphite reaction as noted in FSAR Section 14.5.2.2, demonstrates that these effects are not safety significant with regard to the structural integrity of graphite core

. support posts, bottom reflector blocks, or core support blocks. In addition, there would not be a safety significant . change in the effect on fuel particles or potential fission product release to the primary coolant system. More importantly, the consequences of increased steam ingress do not result in any significant change in the peak primary coolant pressure which could challenge the primary coolant system relief valve rupture disc.

Based only on the review of the reanalysis results, this analysis appears to provide an acceptable basis'to justify a higher value to establish the setpoint for the high primary coolant pressure scram. With the allowance for instrument uncertainty, the new trip setpoint is <46 psi above normal primary coolant pressure.

2.2.3 Superheat Header Temperature - Low Low superheat header temperature initiates a loop shutdown at a present setpoint of 800 F coincident with high differential temperature between loop 1 and 2 at a setpoint of 50 F. This provides protection to preclude a floodout of the steam generators due to an increase in feedwater flow or a reduction in helium flow to a loop. In the reanalysis it is assumed that the trip on loop superheat temperature is initiated at a superheat temperature of 780 F with a differential between loops of 65 F or greater. The impacts of these assumptions were considered for two cases:

30% power and 100% power.

s 7

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - - - - - - - - m-

c w ,

DRh7 There are two basic considerations that are appifcable to this safety e:;uipment protection function. First the trip should be initiated prior to reaching floodout temperatures. Sinc'e the saturation temperature at normal Operating pressure of 2400 psig is 660'F, the assumption of 780*F for citigative action provides an adequate margin of safety prior to reaching

ne' saturation temperature. The second consideration is that loop shutdown should occur before a turbine trip is initiated on low main steam

emperature. This turbine protection is initiated when the main steam temperature (i.e., the temperature of the combined loop steam flow) falls to 800 F.

Since the superheat header temperature for each loop is maintained by controlling primary coolant flow in that loop, a malfunction resulting in low superheat temperature for one loop would not result in a change in-superheat temperature for the other loop. At 30% power, steam temperature is contro11ed 'at about 880 F. Therefore, if loop isolation occurs at a superheat header temperature of 780*F, the temperature difference will be 100*F. The turbine mixed inlet steam temperature will then be 830'F, which assures that the loop temperature difference will satisfy that portion of the trip logic and loop isolation will occur prior to the occurrence of a turbine trip on low main steam temperature. At 100% power, steam temperature is controlled at 1000*F. For this case, the temperature difference between loops is 220*F, and the main steam temperature is 890 F i

when the trip occurs. Thus, the available margins are greater than at 30% l Dower.

Based on this review, it is concluded that this analysis provides an acceptable basis to justify a change in the bases for determining the setpoint for these protection system channeis. With the allowance for instrument uncertainty, the new trip setpoints are 798'F for low superheat header temperature at a 44.8 F differential temperature between loops. I 2.2.4 Circulator Speed - Low l

The present setpoint for the low circulator speed circulator trip is l l

1910 rpm below normal, as programmed by load (feedwater flow). The 8

, i t circulator trip results in a reduction in plant load when operating at full

. load' conditions. Also the low feedwater flow setpoint, which is programmed by circulator speed, is lowered to preclude a trip of the operating circulator. Under.. conditions for single circulator operation the ratio of circulator speed to feedwater flow is about.a factor of two greater than l during normal operation, t

l For the reanalyzed case, it was assumed that a trip does not occur j until a reduction of cir:ulator speed occurs to 2390 rpm below normal. The coastdown f rom rated speed of the circulator by 2390 rpm (25%) is only a matter of a few seconds. At part load conditions, the time to reach this value is about 4 seconds. In addition,-the trip includes a fixed 5 second delay to avoid spurious trips due to changes in circulator speed during normal operation. In contrast, the response of the steam generator superheat header temperature to changes in helium flow is about 30 seconds. Therefore, it was concluded that the assumption of a circulator trip at 2390 rpm below normal is acceptable. i Based on this review, it is concluded that this analysis provides an acceptable basis to justify a change in the bases for determining the trip setpoint for these protective system channels. With the allowance for instrumentation uncertainties, the trip setpoint is 1850 rpm below normal .

as programmed by feedwater flow.

2.2.5 Fixed Feedwater Flow - Low Because of the draft SER transmitted to PSC by letter dated January 24, 1986,5 this setpoint is not being changed in the proposed amendment request (Reference 3). The discussion below is enclosed only for completeness and pertains to the PSC letter of June 21, 1985.

The setpoint for the fixed low feedwater flow circulator trip is 20%

of rated feedwater flow. Since both circulators in a loop are tripped on low flow, this results in a loop shutdown, which provides protection against steam generator operation at tube temperatures above design values.

9

i

.:* u Two basic operating conditions were addressed in the revised analysis t support an assumption that the fixed low feedwater flow trip occurs at 5'. of ratedifeedwater flow. The first condition addressed a sudden total loss of. feedwater flow to a steam generator during both one and two loop cperation. Under such conditions feedwater flow is reduced to zero flow instantaneously. Due to a built-in 5 sec delay,- loop isolation occurs 5 sec following the occurrence of these events. Under this condition the c:nsequences of these events are the same as indicated by the original FSAR analysis, and tube temperatures remain below design limits.

The second condition addressed was continued operation at reduced feedwater flow. However, under this condition, the minimum feedwater flow rate considered was 14% of rated flow. With regard to static boiling stability conditions, it is noted that even if unstable boiling conditions are encountered at flow rates below 18.6%, the maximum helium temperature available at the Superheat II inlet would be less than 957 F, and, thus, could not result in significantly exceeding the maximum allowable temperature of 952 F at the limiting tube location. While it is noted that this analysis is conservative, since it postulates that a hot gas streak could penetrate the entire economizer-evaporator-superheater bundle from top to bottom with no mixing, it cannot be. concluded that this analysis justifies an assumption of loop isolation at feedwater flows as low as 5%

of rated flow.

Based on this analysis, an acceptable basis has not been . set forth to 1

- support the proposed change in the low feedwater flow trip setpoint.

I 2.2.6 Loss of Circulator Bearing Water 4

l The present circulator trip on the loss of bearing water is initiated l

[

when the bearing water dif ferential pressure, with respect to primary l

coolant pressure, is reduced to a low differential pressure of 475 psid.

This provides protection for the circulator bearings on a loss of the i

rormal 'and backup bearing water supply systems. In addition to a trip of i the helium circulator, the protective action includes the actuation of the 10

.I bearing water accumulators to' provide a source of bearinc water during circulator coastdown and operation of the circulator brake and seal system, .

as well'.as isolation of the circulator auxiliary system service lines. The f latter. ensures the integrity of the primary coolant system when the dynamic j seal provided by the bearing water system is not available. f The reanalysis of the operation of the loss of bearing water protection was undertaken based on the assumption that the safety action is. j initiated at a differential pressure of 450 psid. From prior testing of the bearing water system, the minimum differential pressure during a  ;

transient response of the system was 375 psid. From this data it is concluded that a 25 psid reduction in the trip setpoint would result-in a transient minimum differential pressure of 350 psid. Based on this value, analyses and tests demonstrate that the bearing acceptance criterion of a einimum clearance of 0.001 inches will be maintained.

Based on this review, it is concluded that an acceptable basis has been provided to justify a lower setpoint for this safety action. With an allowance for instrument uncertainty; the new trip setpoint is 459 psid.

2.2.7 Circulator Speed - High At the time of the PSC June 21, 1985 submittal, the setpoint for the trip of the helium circulator steam turbine drive was 11,000 rpm. This provided protection to assure that the circulator did nct exceed the design speed limit of 13,500 rpm. For steam line ruptures down stream of the.

circulator steam turbine, the maximum speed is 13,264 rpm with no control action or overspeed trip. Therefore, this event does not establish a limit '

for an acceptable high speed setpoint.

}

With the 11,000 rpm assumed overspeed trip value, the maximum j i

transient overspeed for a loss of restraining torque ever.t (compressor section blade shedding) was 13,050 rpm. Reanalysis with an assumed overspeed trip value of 11,500 rpm results in a maximum transient overspeed l of 13,267 rpm. Based on these analyses, it is extrapolated that an assumed t

i l

11

_ _ _ = _ _ ____ _ ___ _ )

cverspeed trip.at 11,700 rpm would result in a maximum transient overspeed cf 13,370 rpm or less. The 11,700 rpm trip value was subsequently approved 10 by the NRC Staff in' Amendment No. 52 to th'e Facility Operating Licensee.

l {

Based on this analysis and previous approval, it is concluded that an assumed overspeed trip value of 11,700 rpm provides an acceptable basis' for determining the trip setpoint for this protection function. With the allowance for instrument uncertainty, the overspeed trip setpoint is 11,495 rpm.

" 2.2 8 - Neutron ' Flux - High e

The setpoint for the high neutron flux scram is 140% of rated thermal power. As a consequence of uncertainties in the reactor power measurement, the setpoint for the high neutron flux scram has been administrative 1y  ;

controlled and adjusted at conservative values based on indicated reactor power. The Licensee provided curves that are currently being used to control the setpoint for the high neutron flux scram as well' as the high neutron flux rod withdrawal prohibit. In the PSC June 21, 1985, letter, the Licensee proposed to delete the values for the trip setpoints for the

~

crotective actions and to note that these settings are to be established for each fuel cycle and implemented based upon the approval of the Nuclear Facility Safety Committee. The NRC staff found that this proposal was unacceptable since these changes potentially could create an unreviewed safety question. Therefore the curves which define these setpoints were to nave been retained in the subsequent PSC resubmittal of May 15, 1986.

however, the high neutron flux rod withdrawal prohibit curve was not included in the PSC May 15, 1986, resubmittal. In the latest PSC submittal 3

of August 28, 1987 the setpoint curve was included for the high neutron flux scram. The Linear Channel-High Power RWP (Channels 3, 4 and 5) was deleted in this last submittal per NRC direction to focus on only PPS functions that exist in the current FSV Technical Specifications.

12

_ _ _ = _ _ - _ __-

' Based on the above evaluation, it is. concluded that the neutron flux-high scram trip: setpoint and allowable value presented in TS Figure 3,3-1. meets the intent of accommodating instrumentation inaccuracy, e.

2.3' Evaluation of Proposed Technical Specification Changes 2.'3.1' Limiting Safety System Settings (LSSS) (Section 3.3) 3 The Licensee letter of August 28, 1987 proposed changes to Technical Specification.Section 3.3, Limiting Safety System Settings,

, Proposed revisions were on TS pp. 3.3-1, 3.3-2a, 2b, 2c, 3.3-3a, 3b, 3.3-4, 3.3-5, 3.3-6, 3.3-7, 3.3-8, and 3.3-9. These revised pages replaced existing pp. 3.3-1, 2, 3, 4, 5, 6, 7 and 8. Evaluation of the individual changes is given below.

The added definitions for Trip Setpoint .and Allowable Value on TS p. 3.3-1 clarify them as the least conservative "as lef t" and "as found" value respectively, for a channel to-be considered operable. These definitions are in agreement with the guidance given in comments 2 and 5 of Enclosure 4 to the NRC letter of January 24, 1986.

In Table 3.3.-1, Limiting Safety System Settings, a Trip Setpoint and All wable Value are specified for each scram, loop shutdown / steam water dunp, and pressure relief trip function. Figures 3.3-1 and 3.3-2 were added for the Linear Channel-High Neutron Flux and Primary Coolant Pressure-Programmed Low and High. Figure 3.3-1 accounts for the detector decalibration for Cycle 4 as a function of indicated thermal power.

Figure 3.3-2 gives the allowable high and low primary coolant pressure j programmed with circulator inlet temperature. These setpoints and f allowable values are as presented by PSC in their letter of June 21,198S f 6

and as updated to respond to the NRC letters of October 16, 1986 and November 28, 1986.11 These latter NRC letters recommended that PSC I 1

distinguish between all Trip Setpoints and Allowable Values by accounting i l

for setpoint tolerance and instrumentation drift based on the annual or 1 refueling interval measured drift.

13 ,

I l

P On pp. 3.3-4 to 3.3-9 the Basis for Specification LSSS 3.3 is given.

The setpoint methodology for determining Trip Setpoints and Allowable Values is described as well'as the basis for each limiting safety system parameter. The. basis descriptions are consistent with FSAR, Sections- 7.1.2.3, 7.1.2.4 and 7.1.2.5 and-the licensing basis and discussion presented in Attachment 4 to the PSC letter of June 21, 1985.I Based on the above evaluation and the evaluations of Section 2.1 and 2.2 of this report, it is judged that the proposed changes are acceptable with the following exceptions.

Reheat Steam Temperature-High In Table 3.3-1, Item 1.b) and Table 4.4-1 (Part 1), Item 5, Reheat Steam Temperature-High cram, the allowable value is <1067"F whereas in the PCS letter of May 15, 1986 this value is <1061 F. The transmittal letter (of PSC letter of August 28,1987) does not call out this change. Nor is i this change addressed anywhere else in the attachment. As this function l previously had drif t accounted for, it is not obvious why the value has  !

changed.

l Drimary Coolant Pressure vs. Circulator Inlet Temperature In Figure 3.3-2, Primary Coolant Pressure vs.. Circulator Inlet Temperature, in the most upper left and most lower right legend block, the first entry should be " Allowable Valte" not just the word " Allowable."

This change was made in the PSC submittal of May 15, 1986 per NRC direction in letter dated January 24, 1986 but was left off in the latest submittal.

NRC Request 3 The PSC response in letter dated August 28, 1987,3 resolves the l

major discrepancy in Case 2 of 14,580 lb in FSAR Table 14.5-3 versus

~20,000 lb in FSAR Figure 14.5-2. Further PSC stated:

14

,, R f -

" Allowance should be made for graphic artists'. tolerance in D W

~ transcribing data to curves. Other possible causes of minor apparent l diser.epancies are .that in some cases the' steam graphite reaction may 1

.not be completed at the time of cut-off at the right side of the-i l . figures, and/or the. drainage of water from the steam generator into ]

'the.PCRV'may not have been completed at that time."

' FSAR Table 14.5-3, Cases 3, 5, and 6 still differ from their respective Figures 14 5-3,14.5-5, and 14.5-6 in the value of " Steam in Primary Coolant System" (see..below):

FSAR Section 14.5 i

TABLE'14.5-3. STEAM IN PCS Total H 0 ' Total H 0 o ne e 7 7 i Inleakage Reacted and Reacted Figure-Steam in PCS Case- -(lb) (lb) (lb) (ib) 3 6,240 185 6,055 4800 (Figure 14.5-3) 5 16,040 900 15,140 15,800 (Figure 14.5-Sa or b) 6 8,080 919 , a1 6,800 (Figure 14.5-6) 1 These differences in " steam in the primary coolant system" between FSAR Table 14.5-3 and the FSAR Figures are much larger than what should be allowed for graphic artists' tolerance, and since the Figure values for Cases 3 and 6 are still decreasing at the time of cut-off at the right side of the figures, the figure values would deviate by even more' than indicated in the above table. PSC should make the " steam in the primary coolant system" consistent between the Table 14.5-3 and Figure values for Cases 3, 5, and 6.

15

(-

L . -.

'i

\= .. .

Reactor Vessel Pressure Limiting Safety System Setting l i l-In Table 3.3-1, Items 2.c), 2.'d), and 2.e) for all entries in the Trip Setpoint and Allowable Value columns there is a plus and minus setpoint and a single allowable value. This does not agree with STS practice or with j 8

'the following PSC. statements of STS practice and NRC guidance (see p.1 l L

of Attachment 3 to PSC letter dated June 21, 1985):

l "Setpoints in the STS are defined as limits with either greater than or less than, in contrast tc the tolerances with plus or

~m inus'used by PSC. In addition, PSC defined a reportable occurrence as exceeding an Absolute Value, as opposed to an Allowable Value. As a result, in their letter The Commission recommended that FSV PPS setpoints be specified in terms of an Allowable Value and a Trip Setpoint, " expressed as either greater than or less than as well as equal to the.value specified."

PSC should reevaluate the above Trip Setpoint and Allowable values accommodating the quoted NRC guidance.

Basis for Specification LSSS 3.3 Under the heading " General Methodology" on p. 3.3-5, the' phrase "the greater value of" is applicable to the subsequent Items a. , b. , and c. but shouldn't be. Items a. , b. , and c. should be cumulative.

Also in the first two paragraphs, p. 3.3-5, there appear to be two definitions of how the " Allowable Value" is separated from the " Analysis Value." The first is:

"The remaining three factors contributing to instrument error and

[

used to determine the Allowable Values are:

16

T ,,

7 The greater value of:

.a. ' Accuracy of components not calibrated when the'setpoint is measured; or actual drif t data from calibrations.

b. Accuracy of~ test equipment used to calibrate instruments, and c, Design drift allowances (including environmental effects) on equipment accuracy."

The second-is:

"A " total inaccuracy"'value which was calculated, based on the refueling. surveillance frequency, was used to determine the margin between the Analysis Value and the Allowable Value."

Also per ISA Standard S67.04-1982, in Item a., the accuracy and drift should be accounted for not accuracy or drif t of the subject components.

i The last paragraph of p. 3.3-5 states: j "The Trip Setpoint was determined by accounting for the " total inaccuracy" of that portion of the instrument channel not tested during the monthly functional test plus the drift of that portion of the instrument channel which is tested during the monthly functional test. The value obtained by adding these factors is the margin between the Analysis Value and the Trip Setpoint."

This is not per the ISA Standard S67.04-1982 nor per the NRC guidance most recently given in letter dated November 26, 1986 of "We recommend that you propose TS based on the annual (or refueling interval) allowable values." Also this present PSC distinction continues to ignore the recommendation made in the NRC letter dated October 16, 1986 in which it was emphasized that the separation between " Allowable Value" and " Trip

{

17

.. ,. y }

i ,

i

^Setpoint" per the ISA 567104 Standard is to segregate that part of the i p' instrumentation inaccuracy that is_ subject to change with time,.namely  !

drift. The " inaccuracy" of'the channel not tested during the monthly l

functional test is usually a fixed known value that doesn't change and should not be part of the separation'.between " Allowable Value" and " Trip. j Setpoint."

i Basis For Helium Circulator penetration Interspace Pressure On p. :3.3-9 in the basis for the helium circulator penetration interspace pressure, the third sentence reads:

"The rupture discs would burst in the pressure range of 809 psig

(-2%) to 842 psig (+2%). The safety valves would open in the ~

range of 781 psig (-3%) to 829 psig (+3%) and would relieve at full capacity'at 886 psig (10% accumulation)."

Whereas in Table'3.3-1, p. 3.3-2c, Item 2.d), the' Trip Setpoint is quoted as "825 psig plus or minus 17 psi." The Lower limit burst pressure is thus 809 psig in the basis but.825 psig - 17 psi = 808 psig in the Table. These values should be consistent. If the 17 psi is a conservative roundoff' for 2% of 825 psig (namely 16.5 psi) on the low limit side then for consistency, 16 psi should be used for the high limit side for a value of 825 + 2% of 641 psig. It is recommended that the. Table values be left as is but the basis value be changed from 809 psig to 808 psig.

Basis For Steam Generator penetration Interspace Pressure Same comment as for the Helium Circulator Penetration Interspace pressure on 808 psig versus 809 psig for the -2% lower limit on the 825 psig trip setpoint. Make the rupture disc lower limit burst pressure consistent between Table 3.3-1, p. 3.3-2c, Item 2.e) and the value in the basis on p. 3.3-9.

18

0 ,

+  :.

i L 2.3.2 Protection System Instrumentation'n,- Limiting Conditions for Operation

~

(LCOs) (Section 4.4.1)' .;

The Licensee letter of August 28, 1987 3proposed changes'to Technical Specification Section 4.4.1, Protective System Instrumentation, Limiting. Conditions for Operation. Proposed revisions were on TS pp. 4.4-1, 2, 3a, 3b, 3c, 4a, 46, 4c, 4d, Sa, Sb, Sc, 7a, 7b, 8,~ 10, 10a,

-10b,.10c, 11, lla, 12, 12a',.12b, 12c, and 13. These revised pages. replaced!

existing pp. 4.4-1.through' 4.4-8, 4.4-10, 11, 12, and 13 . Existing-pp. 4.4-6a,.6b, and 6c on the Steam Line Rupture Detection and Isolation System (SLRDIS) are unchanged as is p. 4.4-9. Evaluation of the ir,dividual changes is given below.

The added definitions on p. d.4-1 of Trip Se'tpoint and Allowable Value are-as discussed earlier (Section 2.3.1 of this report) to distinguish between "as-left" and "as found" values, respectively.

On p. 4.4-2, clarification is made that LCOs 4.2.10 and 4.2.11 apply

-during the time that the PPS moisture monitor' trips are disabled. This is just a reminder to the operators since LCOs 4.2.10 and 4.2.11 would apply with or without this clarification. On this same page, the action for inoperable channels for Table 4.4.3, circulator trip, now provides a choice of either reactor shutdown or circulator' shutdown rather than the previous requirement of just circulator shutdown. Reactor shutdown is a more stringent action than just circulator shutdown and is therefore acceptable, i

In Tables 4.4-1, 4.4-2, 4.4-3, and 4.4-4 reformatting was provided by splitting each Table into Part 1, containing Trip Setpoint and Allowable Value, and Part 2, containing Minimum Operable Channels, Minimum Degree of Redundancy, and Permissible Bypass Conditions. Primarily, the changes are to account for instrumentation inaccuracy as presented by PSC in their 1 1

letter of June 21, 1985 and as updated to respond to the NRC letters of October 16, 19866 and November 26, 1986.11 These latter NRC letters  :

1 recommended that PSC distinguish between all Trip Setpoints and Allowable j Values by accounting for setpoint tolerance and instrumentation drift based j en the annual or refueling interval measured drift.

19

' On pp. 4.4-10,10a,10b.10c,11, lla,12,12a,12b,12c and 13 the Basis for specification 4.4.1 is given. The setpoint met odology for determining Trip Setpoints and Allowable Values is as des:ribed for the LSSS basis in the previous section of this report (2.3.1). Each trip for scram, loop shutdown, circulator trip, and rod withdrawal prohibit functions are described and are consistent with FSAR Sect'on 7.1.2.3, 7.1.2.4, 7.1.2.5, and 7.1.2.6 and the licensing basis and discussion presented in Attachment 4 to the PSC letter of June 21, 1985.1 Several of the proposed changes in Section 4.4.1 are not directly associated with accounting for instrumentation inaccuracy. Scme of these other changes have already been discussed for p. 4.4-2. The remaining items are discussed below.

In Table 4.4-1 (Part 1), Item 10, Plant Electrical System-Loss, Note (d) was deleted from the Trip Setting column and replaced with the Trip Setpoint and Allowable Value and correspondingly note (d) was deleted on p. 4.4-8, Notes for Tables 4.4-1 Through 4.4-4. Also for this same scram function, note (e) on p. 4.4-8 was updated to corre:tly describe the undervoltage system design. Note (e) appears for the Platt Electrical System-Loss scram function in Table 4.4-1, Part 2, under Miniinum Operable Channels. Updating of Note (e) is consistent with the FSAR Rev. 5 description of the Plant Electrical System-Loss scram fur: tion in Section 7.1.2.3 and FSAR Table 7.1-2 and is therefore acceptable.

In Table 4.4-1, Part 2, Item 4. , Primary Coolant Moisture High Level Monitor and Loop Monitor, under Permissible Bypass Conditions, the existing "none" and note (h) were clarified as note (h2) for the High Level Monitor and note (hl) for the Loop Monitor. Addition of note (h2) for the High Level Monitor just recognizes an existing Permissible By; ass Condition in LC0 4.9.2. Note (hl) is unchanged from the previous Permissible Bypass Condition note (h) for the Loop Monitor.

20

i .

DRF In Table 4.4-2 (Part 2), p. 4.4-4d, Item 7c., High Differential Temperature Between Loop 1 and Loop E, the Permissible Bypass Condition has Deen changed from "none" to "less than 30% rated power." This change is acceptable ~ as High Differential Temperature Between Loop 1 and Loop 2 is a coincident requirement-[see Footnote (p) on p. 4.4-8]'for Item 7a., Low Superheat Header Temperature, Loop 1, and for Item 7b., Low Superheat Header Temperature, Loop 2. As the existing Fort St. Vrain TS Permissible Sypass Conditions for Items 7a. and 7b. are both "less than 30% rated power," it is only consistent that the coincidence requirement, Item 7c.,

have.the same Permissible Bypass Condition (see the exceptions beginning on the next page for further comment on justification for bypass conditions).

The "*" footnote has been deleted for Circulator Speed-High Water under the " Minimum Operable Channels" and " Minimum Degree of Redundancy,"

Table 4.4-3 (Part 2), Item 9. A request for additional information on this item was submitted as Item 15 to the Enclosure of NRC letter dated l October 16, 1986.6 PSC's response in letter dated August 28, 1987 stated that removal of the footnote is more conservative as the applicability is now to have channels operable for each circulator versus the one per loop allowed'with the footnote. PSC determined that the previous allowed flexibility of operable channels for only one circulator per loop would not nave been exercised and so deleted the footnote. This deletion is in the conservative direction and removes a flexibility that in retrospect was unwarranted and therefore the deletion is acceptable.

In Table 4.4-4, Part 2, Items 3a. and 3b. , Linear Channel-High Power RWP (Channels 3, 4, and 5 and Channels 6, 7, and 8), under " permissible Bypass Conditions" "none" was changed to "above 30% rated power." If this RWP function is bypassed above 30% rated power but the Interlock Sequence Switch is lef t in the Low Power Position, then the block on outward rod motion is defeated even though it shouldn't be. This change is therefore unacceptable.

l l

I I

l 21 L

L_- l

1  !

l NT.

Other minor editorial changes (commas, hyphens, consistency in titles,.

capitalization, etc.) have been made but were not specifically listed in

'PSC's " summary of-proposed changes." These editorial changes are l acceptable as are the expanded bases of the scram, loop shutdown, circulator trip, and RWP functions, i

Based on the above evaluations and evaluations of Section 2.1 and 2.2 of this report, it is judged that-th'e changes to TS Section 4.4.1 are acceptable with the following exceptions.

l Permissible Bypass Condition Several trip functions in Tables 4.4-1, 2, and 3 have permissible bypass conditions of "less than 30% rated power." -Some of these trip functions were reanalyzed to increase the margin between the trip setpoint and the normal operating value of the subject parameter. There is no indication that the permissible bypass condition value of "less than 30%

rated power" had instrumentation inaccuracy accounted for. Further, there is no explicit justification in the reanalyses for any permissible bypass condition let alone for instrumentation inaccuracy' in any specific value.

The following reanalyzed trip functions did not have their. bypass conditions justified: Primary Coolant Pressure-Low, Superheat Header Temperature-Low, and Circulator Speed-Low. The bypass condition for Fixed Feedwater Flow-Low has already been commented on in previous correspondence I

(NRC letters of January 24, 1986 and October 16,1986) and is individually covered in another item below. As the issue of justification of the permissible bypass condition value or whether instrumentation inaccuracy is

- accounted for in the value is related to but not directly part of accounting for instrumentation inaccuracy in the trip settings, it is judged that this issue does not need resolved for approval of the subject submittal. However, PSC should pursue this slightly broader issue in a future submittal.

22

m I

. Linear Channel-High Power RWP (Channels 3, 4, and 5 and Channels 6, 7, and 8[

l.

In Table; 4.4-4, Part 2', Items 3a. and 3b. , Linear Channel-High Power

. RWP (Cha'nnels 3, 4i - .d 5 and . Channels 6, 7,' and 8), under " permissible-

.Sypass Condition," the bypass cond1 tion should remain as "none" and not as L >

ihe proposed "above 30% rated power."

NRC Reguest 6 PSC response to this NRC Request (to justify why.the High Differential Temperature Between Loop 1 and Loop 2 loop shutdown function is not in the FtAR) clar'ified that it'is discussed in FSAR Section 7.1.2.4 as a comparator circuit between the two loops and an interlock. Also footnote "(p)" in TS Table 4.4-2, Part 2, Items 7a., 7b., and 7c. and p. 4.4-8 states that:

" Item 7a. must be accompanied by Item 7c. for. Loop 1 shutdown.

Item 7b. must be accompanied by. Item 7c. for Loop 2 shutdown."

This is the only clear indication in either the Technical' Specifications or FSAR that coincidence is required with High Differential Temperature Between Loop 1 and Loop 2 to get loop shutdown on the Low Superheat Header Temperature trip for either Loop 1 or Loop 2. It is recommended that when the FSAR is revised to add the trip setpoint for the High Differential Temperature Between Loop l'and. Loop 2, that FSAR Section 7.1.2.4 be clarified to explicitly state the coincidence requirement as opposed to the present less clear reference' to a comparator circuit and interlock.

NRC Foquest 7, Deletion of Curve for Circulator Speed-Low i

In NRC letter dated October 16, 1986, request for additional information Number 7, PSC was asked to justify deletion of reference to 23

<1 . .

i.

Figures 4.4-la and;4.4-lb for the circulator Speed-Low trip.

PSC's response was to see'their response to NRC request 4. In their. Response 4, PSC referenced discussion.with the NRC Staff in the July.30,1986 telecon.

PSC stated:

"In a followup telecon, the NRC staff provided PSC with the directicn that the revised amendment request should only include .

.those parameters .which now exist ir, the present Technical Specifications. In addition, those new parameters.would not have had approved surveillance requirements had we included them."

This direction is acceptable for the response to the following listed NRC requests (as all of these functions are not in the existing FSV Technical Specifications):

4, Wide Range Channel Rate of Change-High, 5, Primary Coolant Moisture High Level Monitor and Loop Monitor, 8, Programmed Feedwater Flow-Low, 9, Rod Withdrawal Prohibits for Startup Channel Rate of Change-High and Wide Range Channel Rate of Change-High, 10, Rod Withdrawal Prohibit for Linear Channel-High power RWP (above 30% power) and 12, RWP Multiple Rod Pair Withdrawal.

However, Circulator Speed-Low for the circulator trip is in the  ;

existing FSV Technical Specifications. The programmed curve for this i trip should be supplied as has been done for other existing FSV l.

I 24 l

{ - . _ _ = . -_________a

DRAFT functions that were missing programmed curves in the existing TS but

~

• nich were supplied in the change request [for example, Primary

' Coolant Pressure - Programmed Low in Table 3.3-1, Item 1.c) and Primary' Coolant Pressure-Programmed High in Table '3.3-1, Item 2.a)].

l L NRC Request 8 PSC's response to this request was to "see PSC Response 4." PSC Response 4'was basically that trip functions not in the existing FSV Technical Specifications were left out of PSC's August 28, 1987.

submittal. While this reference to Response 4 is appropriate for cost of the comments on the Programmed Feedwater Flow-Low function in NRC Request 4, it is not appropriate for the question posed in the last. sentence of the NRC Request 8, namely:

"Also, the NRC letter of January 24, 1986, did request additional analyses for the Fixed Feedwater Flow-Low setpoint, but PSC did not provide or mention these latter analyses in their letter."

The NRC letter of January 24, 1986,5 Enclomre 3, raised the concern that there was no apparent safety analysis to justify cypass at less than 30 percent power on circulator trips on fixed feedwater flow-low. This concern and the parallel concern for the Programmed j Feedwater Flow-Low function (when it is submitted) remain unanswered.

This issue may be pursued by PSC in the future when the Programmed .

Feedwater Flow-Low function analyses to support instrumentation inaccuracy is submitted.

_ ___ _______________ - _____ Q

[  :

NRC Request 13 Rod Withdrawal Prohibit (RWP) at 30% Rated Thermal Power (RTP) e l

The Licensee's. position regarding not providing instrumentation

. inaccuracy'_for the 530% of rated power RWP setpoint, remains unacceptable.

P.6, Attschment'3 to the PSC letter of June 21, 1985,I stated that the rod

• 'thdrawal prohibits were not analyzed as part of the ' program to comply with

~

tne guidance of the ISA standard S67.04, .because no credit is taken for them-it accident analyses. This Licensee position was challengeo in NRC letter cated October.16, 1968 which requested additional information to clarify.

wny, . at least, the $30% of rated power RWP setpoint. does not require instrument uncertainty to be taken into account, P.4.4-6a, Table 4.4-4

'(Part 1) and to also, reevaluate the other RWPs to ensure that if they were del 'ed, an operator single failure in positioning the Interlock Sequence Switch (ISS) would not bypass required reactor protection trip functions.

Tne NRC letter stated that: i "Without the rod withdrawal prohibit, high power operation l

(>30%)fcould be commenced with the interlock sequence switch in the low power position with four scram functions and two circulator trip functions bypassed (FSAR Section 7.1.2.8). As this is an operator single failure. defeat of part of the reactor protection system at high power, the 30% of rated power RWP appears to be a required safety function to prevent this occurrence. Therefore, at least this function of the RWP should have had instrument uncertainty taken into account for the setpoint. Otherwise, additional safety analyses are required to demonstrate safe operation with the above reactor protection system functions bypassed."

The Licensee in letter. dated August 28, 19873 argued that backing  !

off the 30% RWP to accommodate instrument inaccuracies is inappropriate and unwarranted. The Licensee stated:

26

DWT l "The ISS, as explained in FSAR Section 7.1.2.8, is an administratively controlled method for operating protection

' # system bypasses during rise to power. In this regard it is l similar to the BWR Reactor: Mode Switch (NUREG 0123 Rev. 3). The~ f 30% RWP is incibded as a second line of defense (or added remindcr) to the reactor operator to place the ISS in the correct position prior to exceeding 30% reactor power. FSAR Table 7.1-6 is an analysis of improper ISS settings and the effect'on rise to power. l "The underlying rationale for not applying instrumentation uncertainties to the RWP circuitry is that of avoiding- the potential for initiating protective actions when system conditions do not warrant it. To apply uncertainties to these parameters, especially the 30% RWP, would mean backing off from this value, thus resulting in a setpoint somewhat less than 30%. By doing so, certain plant protective functions would be

" enabled" prior to system operating parameters (pressures, flows, temperatures, etc.) being within normal operating conditions."

The Licensee further clarified that the linear and wide range nuclear instrument channels, which input signals proportionate to reactor power to the RWP circuitry, are calibrated against a secondary heat balance prior to reaching 30%~ power (in the range of 26 to 28% power). And due to the accuracy of the secondary calorimetric and the RWP circuitry, reactor power would not exceed about 34% without actuating the RWP. If the operator were to neglect placing the ISS in the Power position and exceed 30% power, it is highly unlikely that an accident would occur in this circumstance, due to the short time spent in the 30% to 34% power range before the RWP would be received during the rise-to power. Also, the Licensee states that the Steam Line Rupture Detection / Isolation System (SLRDIS) is now relied on rather than the Hot Reheat Pressure-Low and Main Steam Pressure-Low scram parameters even though the later will continue to be in the Technical Specifications. Finally, the Licensee states that the turbine generator is brought on-line with the external electrical grid at approximately 28%

i 27

, E .g i 1

I reactor power,'and this action needs to be accomplished with stability without being encumbered with a rod-withdraw prohibit. setting in the same -

range., and reducing the RWP setting would also intrude into the 26% to 28%

L I' range'wherej secondary heat balances are made (heat balances 'are less accurate if performed at lower' power levels). 3 1

I The Licensee in its response has failed to consider several aspects relating,to the $30% RWP setpoint.. These aspects are as follows:

1. The fundamental question is whether or not the safety analysis for protecting against accident situations remains valid if the operator were to inadvertently proceed to power levels above 30%

RTP without positioning the ISS to the'" power" position,

2. The FSV Interlock Sequence Switch (ISS) and Reactor Mode Switch (RMS) do not provide the same level of protection against

-inadvertent operation outside intended bounds as does the BWR Reactor _ Mode Switch or the PWR Reactor Protection System interlocks, and

3. The Licensee has argued about the difficulties of lowering the i RWP setpoint below 30% RTP but has not pursued increasing it above 30% RTP.

Each of these aspects will be addressed individually below.

The fundamental point of accounting for inaccuracy in the 30% RWP setpoint is to guarantee that the safety analysis for the reactor trip l system remains valid. This is the same rationale for pursuing accounting for' instrumentation inaccuracy in the other plant protective system (PPS) setpoints (scrams, loop shutdowns, and circulator trips). Certainly, accounting for inaccuracy in PPS setpoints, as has been done and as has been the intent of this Technical Specification change effort, is of little consequence if the PPS function and setpoint is bypassed because the ISS is 28

I i L +, positioned to Low. Power (5 30% RTP)-when operation may actually be occurring at Power (>30% RTP). The RWP setpoint' inaccuracy may permit operation in such an unanalyzed condition the same as if one of the other PPS setpoints .;

.had not been analyzed to account for its instrumentation inaccuracy, but with the ISS'in the correct position. The Licensee' stated that " accident.

' consequences for power. range' accidents are analyzed at conservative upper power limits" and that "the-consequences of accidents occurring from lower  ;

power levels have not generally been analyzed." The Licensee needs to make i this more precise. The' power level at which the PPS trip functions, those that are bypassed in the Low Power ISS position, are needed should be well defined. Simply performing accident analysis at conservative upper power limits while demonstrating the PPS trips are adequate to ensure protection at worst case conditions does not establish at what low power level.'the trips may be ' bypassed. Likewise stating that lowering the RWP 30% RTP.

setpoint would cause difficulties because the turbine generator is brought

-on-line at approximately 28% reactor power and that secondary heat balances are made in the 26% to 28% power range does not constitute a valid basis for allowing potential single-failure defeat of PPS trips. The Licensee

~

has also argued that not accounting for the 530% RWP setpoint instrumentation inaccuracy may at most place the plant at risk in about a 4% power interval centered around 30% RTP and that it is highly unlikely that an accident would initiate in this circumstance due to the short time spent in the 30% to 34% power range. This last argument is unacceptable as the accepted reactor protection system practice is to provide protection over the full allowable power range without exception. The Licensee further argued that the ISS and RWP are included as a second line of defense (or added reminder) to the reactor operator to place the ISS in the correct position prior to exceeding 30% reactor power. This Licensee argument is the exact reason that the 530% RWP setpoint should be rigorous and include instrumentation inaccuracy. It is the single operator error of not positioning the ISS to the power position while exceeding 30% RTP that constitutes a singic failure defeat of some of the PPS functions. Per General Design Criteria 19 and 20 (see Appendix C, Rev. 5 of the FSV FSAR) l and as argued by the Licensee in stating that these criteria are met, the PPS has high functional reliability and redundancy to 29

-__-_-_____________-____a

4 s-

. assure that no single failure will result in loss of the protection function. As will,be discussed immediately below other reactor designs will result.in automatic ' scram if the operator attempts to go to higher power than~that permitted by the Reactor Mode Switch. As the FSV design does not provide for such automatic scram, the RWP provides the required backup.

The FSV ISS and RMS do not provide the same level of protection against' inadvertent operation outside intended bounds as do BWR or PWR systems. At FSV the operator may proceed to power levels above 30% RTP with the ISS;in the Low Power position and thus defeat various PPS functions intended to be operable above 30% RTP. This same situation does not exist.in BWR and PWR designs. In a BWR the Reactor Mode Switch has four positions: Startup, Run, Shutdown, and Refueling. If the operator inadvertently tried to go to power with the RMS in Startup, the plant would automatically scram on the Intermediate Range Monitor-High trip. In contrast, at FSV the startup trip, the High Wide Range Channel Rate of Neutron Flux Change, is bypassed in the Low Power position of the ISS and therefore will not cause an automatic scram. In the PWR design, on increasing reactor power, the P-6 and P-10 interlocks allow manual block of the Source Range trip, the Intermediate Range trip, and the Low Setpoint Power Range trip. On increasing reactor power, the P-7, P-8 and P-9 inter' locks automatically enables reactor trips that are intended to be operable as various higher power levels are reached. The operator cannot defeat these automatically enable reactor trip function interlocks. At most, the operator could fail to block the startup trips when allowed but these would only cause automatic reactor scram as their setpoints were reached on progressing to higher powers. Because of this significant difference in the FSV design that allows the operator to inadvertently bypass certain reactor protective functions when they were intended to be  ;

operable, the RWP takes on an added safety significance for FSV to block outward rod motion when the ISS is not correctly positioned.

I 1

i 30 I l

i

-_ -- ._ D

~. ,

l- .

The Licensee has pursued and ' explained the difficulties of lowering the RWP setpoint below 30%'RTP but has not pursued increasing it above 30%

RTP, For other PPS trips the Licensee has performed additional safety analysis to allow raising the involved setpoint so as to avoid inadvertent actuations when the instrument uncertainty is accounted for. Also, if the present'30%-RWP setpoint is subject to the approximate 4% instrumentation uncertainty stated by the Licensee then the actual power say.be at 26% RTP when the 30% RWP setpoint is actuated. This would appear to confuse the

. issues brought up by the Licensee of potentially' interfering with bringing the turbine generator on-line at about 28% reactor power and doing secondary heat balances between 26% and 28% reactor power. An all around more appropris.e solution would appear to be doing additional safety analysis to raise the 30% RTP RWP setpoint so that even after the 4%

instrumentation inaccuracy is accounted for, the setpoint is still sufficiently high (say 34% RTP) so that interfering with the turbine generator and secondary heat balance is avoided.

It is recommended that the Licensee reevaluate the 5 0%3 RTP RWP setpoint to account for instrumentation inaccuracy as discussed above. As the 30% RWP setpoint can be reevaluated without the need to delay the inclusion of instrumentation inaccuracy in the other PPS trip setpoints, it is recorrnended that the 30% RWP setpoint reevaluation be nandled as a separate issue. In the interim, the remaining PPS setpoi ,ts for which instrumentation inaccuracy has already been accounted for could be approved now for facility use.

480 V AC Essential Bus Undervoltage Protection Trip 5etpoints In PSC letter dated August 24, 1987 (P-87272),II Attachment 2, p. 2, NRC Comment (1) and the PSC Response are:

4 31

q m- - -_ --. - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ . _ _ _ _

S,.' ,a i:

p NRC Comment (1):

" Tables ~.4.4.5 and 5.4.5.and associated notes.were to be added to the Technical Specifications. ~ Reference [3] included these I tables as Tables 3.3.1.5 and 4.3.1.5. We note that the time I

-dial setting for Functional Unit 3 changed from 6 to 5 in the process...The licensee should verify the correct settings of these undervoltage relays and commit to having these tables and I

notes in the upgraded Electrical Technical Specifications. It should' include nominal 'setpoints and allowable limits (where voltage arid time tolerances exist)."

_"PSC Response:

"This comment is associated with the PPS Technical Specification amendment and will be addressed as part of the PPS submittal."

Contrary to PSC's stated response the only information on essential b'us undervoltage is Item 10. of Table 4.4-1 (Part 1) where the Trip Setpoint, 3278 V and $31.5 seconds, and Allowable Value, 3266 V and

$35 second are listed. No Trip Setpoints, Delay Times, and Allowable Values are provided for Degraded Voltage, Loss of Voltage Automatic Throw Over(ATO),orLossofVoltage-D.G. Start,LoadShedandLoadSequence.

This latter information had been presented in Tables 3.3.1.5 and 4.3.1.5 of I

the November 30, 1985 Oraf t Technical Specifications and is still 4 required.

High Reactor Building Temperature (Pipe Cavity)

In Table 4.4-1 (Part 1), Item 12, P.4.4-3b, High Reactor Building Temperature (Pipe Cavity), the allowable value has been changed from 165 F, PSC letter of May 15, 1986, to 166 F. There is no explanation for this change. As this function previously had drift accounted for, it is not obvious why the value has changed.

32

Low Superheat Header' Temperature DRAF l

In Table 4.4-2 (Part 1), Items 7a., 7b., and 7c., a footnote "(p)"

resignation should be added under the " Functional Unit" description for l

each entry: Low Superheat Header Temperature, Loop 1 and 2, and High Oif ferential Temperature Between Loop 1 and Loop 2~. The. footnote'"(p)"

. indicates that Low Superheat Header Temperature in a Loop is required coincident with High Differential Temperature Between Loop 1 and Loop 2 in order to get'a scram. Although the. footnote "(p)" appears in Table 4.4-2 (Part 2) for these Functional Units, .the footnote is 'also applicable to the

_ trip information in Part 1 of Table 4.4-2. In fact normal STS practice would require that Item 7c. not be a separate item, but instead be directly called out as a coincidence requirement in.both Items 7a. and 7b.

Circulator Speed-Low In Table'4.4-3,' Item 1., Circulator Speed-Low, the allowable valu'e of

$2035 rpm below normal was $1974 rpm below normal in the PSC letter of May 15, 1986. As this function previously had drift accounted for, it is not obvious why the value has changed.

l Circulator Seal Malfunction In Table 4.4-3 (Part 1), Item 8., Circulator Seal Malfunction, the allowable value of 576.1" H2 0 was $75.6" H 2 0.in the PSC letter of May 15, 1986. As this function previously had drift accounted for, it is not obvious why the value has changed.

Circulator Speed-High Witer In Table 4.4-3 (Part 1), Item 9. , Circulator Speed-High Water, the allowable value of $8,786 rpm was $8,670 rpm in the PSC letter of May 15, 1986. As th); function previously had drift accounted for, it is not obvious why the value has changed.

33

'i 6 DRE

3. REMAINING ISSUES ASSOCIATED WITH THE PLANT PROTECTIVE SYSTEM INSTRUMENTATION This section is provided as a tummary of the remaining issues that have developed or have been deferred as a result of the protracted effort

~

on the Plant Protective System instrumentation Technical Specification changes. The completion of these remaining issues to be; summarized here are not necessary for approval of the Licensee's proposed Technical Specification changes addressed:in the earlier sections of this report on including instrumentation inaccuracy'in the trip setpoints. Because of the protracted effort involved with the PPS proposed changes, many issues of investigation were separated out of the original. proposal to accommodate instrumentation inaccuracy in the trip setpoint and, have been pursued under separate cover letters for future submittal. Each of these remaining separate issues of investigation are summarized below to status them and to faci.litate future tracking of them.

3.1 Circulator Trip on programmed Feedwater Flow-Low Per PSC's letter of June 21, 1985,1 the absence in the Technical Specifications of Programmed Feedwater Flow-Low was discovered too late to complete the analysis to incorporate instrumentation inaccuracy per the ISA 567.04-1982 methodology. PSC therefore committed to complete the analysis and submit the revised Trip Setpoints and Allowable Values by separate j letter. In the June 21, 1985 letter, PSC proposed to use the existing  !

setpoint'for the interim. The NRC letter of January 24, 1986,5 agreed 2

- with this position. In the subsequent PSC resubmittals of May 15, 1986 and August 28, 1987,3 this function was left out per a telecon agreement with the NRC Staff. Therefore, Trip Setpoints and Allowable Values .and the supporting analysis per the ISA S67.04-1982 methodology are still outstanding for the Programmed Feedwater Flow-Low circulator trip. Also, l PSC should provide justification for any intended bypass of this trip such as "below 30% power" (see similar comment for the Fixed Feedwater Flow-Low trip function in Section 3.2 of this report).

i 34

3.2 Circulator Trip on Fixed Feedwater Flow - Low 5 1 rThe NRC Staff found unacceptable PSC's June 21, 1985 proposal to cnange the' circulator trip on Fixed Feedwater Flow-Low. The NRC Staff.

noted concerns with PSC's discussion of the superheater II high inlet temperatures reached due to a hot gas streak penetrating the . entire i

- e onomizer-evaporator-superheater. In the interim, the NRC Staff recommended continued use of the existing setpoint of 20% of_. rated full )

laad. Consequently, in the subsequent two PPS submittals of May 15, 1986 and August 28, 1987, PSC retained the setpoint of.20% of rated full load.

.In the August 28, 19873 letter, PSC committed to submit a revised setpoint and supporting analysis for Fixed Feedwater Flow-Low (per discussion with the NRC Staff in a telecommunication conference on July 15, 1987). Therefore, this item is still outstanding, ,

Further the NRC Staff identified a second concern with circulator trip on Fixed Feedwater Flow-Low. In Enclosure 3 to the letter of January 24,

.1986, the NRC Staff requested PSC to justify the bypass condition of the Fixed Feedwater Flow-Low trip below 30% power. In a telecommunication conference on July 30, 1986, PSC said they interpreted this request to be applicable to the issue of Fixed Feedwater Flow-Low for setpoints for less inan the pres'ent 20% of normal full load. As it is not obvious that PSC has -correctly interpreted this issue it- has been reiterated in the earlier sections of this report. Therefore, this item is still outstanding.

3.3 -Rod Withdrawal Prohibit at 30% Rated Thermal Power There was no reanalysis'(Section 2.2 of this report) of trip settings to account for instrumentation inaccuracy for the rod withdrawal prohibits. As a minimum, as discussed in detail in Section 2.3.2 of this report the rod withdrawal prohibit trip setpoints and allowable values at 30% rated thermal power should be reanalyzed for instrumentation inaccuracy per the ISA S67.04-1982 methodology. The rod withdrawal prohibit trip setpoints are the method for executing the Permissible Bypass Conditions in  ;

i 4

35 I

l '

the-scram, loop shutdown, and circulator trip functions in Part 2 of Tables 4.4-1, 4.4-2, and 4.4-3, respectively. ' The Permissible Bypass Conditions-themselves are treated directly below.

3.4 PPS Permissible Bypass Conditions In the reanalysis (Section 2.2 of this report) of several trip functions ~ to increase the margin between the trip setpoint and the normal operating value, no justification was presented for the permissible bypass condition. There was no analysis for either ' accounting for instrumentation inaccuracy in the' bypass value or for the specific bypass value itself. As these issues are slightly broader than accounting for-instrumentation inaccuracy in the trip setpoints, they may be pursued without holding up approval of the trip setpoint revisions discussed in Section 2.3 of this report.

3.5 Technical Specification Upgrade Program Related Changes The PSC submittal. of June 21, 19851 included extensive Technical Specification Upgrade Program (TSUP) related changes in the content and format of the limiting conditions for operation and surveillance requirements for the plant protective system instrumentation, in addition to the setpoint changes due to instrumentation inaccuracy. In the NRC letter of January 24, 1986,5 the NRC Staff had a number of concerns related to'the proposed upgrade program related changes. The NRC Staff listed 30 comments in the January 24, 1986 letter's Enclosure 4. Also, because the setpoint changes due to instrumentation inaccuracy were significant safety concerns, the NRC Staff directed PSC to resubmit the setpoint changes early and to propose a separate schedule for the balance of the upgrade related changes. As a result, in the subsequent PSC resubmittals of May 15, 1986 and August 28, 1987, the upgrade related changes had been deleted. As part of the overall Technical Specification Upgrade Program,12 these upgrade related changes to the plant protection l system instrumentation are still desirable and are still outstanding.

l 36

Because the plant protective system instrumentation upgrade related changes in the PSC letter of June 21, 1985 were.so extensive and because of

.the protracted effort involved, the major categories of changes in the PSC q

~ 1etter are' enumerator below. These outstanding. categories of changes were j also briefly discussed with PSC in the telecommunication of July 15, 1987.

3.5.1 TS Section 2.0, Definitions 1

The ' proposed changes to the definitions were on (numbering is the same as that of the PSC June 21,-1985 letter):

2.1 Three Room Control Complex j- 2.la Action 2,1b Allowable Value 2.1c Channels to Trip 2.1d Minimum Channels Operable r 2.le Operational Mode-Mode 2.lf Total No. of Channels-2.lg Trip Setpoint 1

2.1.h Actuation Logic Test 2.1.1 Channel Functional Test l-l-

Items 2.1, 2.la, 2.lb, 2.le, 2.19, 2.1.h, and 2.1.1 have already been addressed in the TSUP (PSC Draft TS of November 30, 198512) and any further action is being pursued in the TSUP, items 2.1c, 2.2d, and 2.If do 37

~ -

~ not: appear'in the Standard. Technical Specifications and are considered--

DRE optional for any' additional' action by PSC.

I

1. .

l 3.5.2 TS Section 4.0, Limiting Conditions for. Operation The proposed changes to Section 4.0, Limiting Conditions for Operation,: were on (numbering is the same as that of the PSC June 21, 1985 letter): 4.0.1 through 4;0.6. Items 4.0.3 through 4.0 63 have already been addressed in' the TSUP and any further action is being pursued in the TSUP.

Items' 4.0.1 and 4.0.2 do not appear in the Standard Technical a.

Specifications and are considered optional for any additional action by PSC .

t. .

S.

3.5.3 TS Section 5.0, Surveillance Requi,rements s , , gq w ,

m,!

6 m The proposed changes to Section 5.0, Surveillance Requirements, were ,

J, on (numbering is the same as that of the PSC June,2h 1985 letter, Attachment 7)

5.0.1 through 5.0.7. Items 5.0.2threbgh.5.0.7have _

1

. already been addressed in the TSUP and any furtherM.tien is Seing pursued in the TSUP. Item 5.0.1 does not! appear in $he Standard Technical Specifications and is considered optional'for any additional act%n by PSC.

m l

? ,,

3 5.4 TS Section 4.4.1. Plant ProtectivW System Instrumentation':LCOs

~

]p The proposed changes to the Plant Protective System Inst umentation section related to upgrade considerations wer$ o,n added trip functions and l format. The added trip function % were those that=cppeer in the FSAR (Chapter 7.0) but that were not in (he existing FSV T6thnical l Specifications. These added trfp-functions were (numdering is the same as j that of the PSC June 21,1985letOr):

p.4.4-2, Table 4.4-1 (Part 1), " Wide Range Channel Rate of Change-High i p.4.4-3a, Table 4.4-2 (Part 1), Prdaary Cotlant' Moisture high Level Mdnitor and Loop Monitor  ;

l 1

38 l

)

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ i s

. o

[ F*P l -

L i

(. ,

p.4.4-5, Table 4.4-4 (Part 1), Rod withdrawal prohibit for Startup Channel Rate of Change-High, Wide Range Channel Rate of Change-High, Linear Channel-High Power RWP (Channels 3, 4, and 5) p.4.4-Sa, Table 4.4-4 (Part 1), Linear Channel-High Power RWP (Channel 6, 7, and 8), Multiple Rod Pair Withdrawal Per NRC direction in a telecommunication, PSC deleted these trip functions from their subsequent PPS submittals of May 15, 1986 and August 28, 1987.

Also, the upgrade format changes to the trip tables of Section 4.4.1 on channel operability, applicable modes, and actions were deleted as discussed earlier. Other deletions and or deferrals not directly related to upgrade issues that were in this Section 4.4.1 in the June 21, 1985 letter have already been discussed in other sections of this report (for example deletion of undervoltage protection).

The added trip functions and upgrade format considerations of Section 4.4.1 are still outstanding and require addressing by PSC. In any  ;

resubmittal of these added trip functions and upgrade format, PSC should address the comments made in Enclosure 4 of the NRC letter of January 24, ,

1986. I 3.5.5 TS Section 5.4.1, Plant Protective System Instrumentation Surveillance and Calibration Requirements The proposed changes to the PPS Surveillance and Calibration Requirenients Section provided STS type testing specifications such as Channel Check, Channel Functional Test, Actuation Logic Test and Applicable j Modes. These testing specifications were added for the existing trip ]

functions of Section 4.4.1 as well as the added functions discussed directly above in Section 3.5,4 of this report. Per NRC direction in a i

1 l

39

' ' uu p w

• L, telecommunication, PSC deleted these upgraded surveillance and calibration requirements from their subsequent PPS resubmittals of May 15, 1986 and j August 28, 1987. l I

The upgraded surveillance and calibration requirements of Section 5.4.1 are still outstanding and require addressing by PSC. In any J resubmittal of these upgraded surveillance and calibration requirements, PSC should address the comments made in Enclosure 4 of the NRC letter of January 24, 1986. .

2 l

40 l

f L____________.

r. 3 4

/

i[

4, CONC LUS10NS-

-An evaluation has been made~of the PSC' submittal of August 28,'1987 on the proposed Technical Specification changes for the trip setpoints for the p' ant protective system to account for instrumentation inaccuracy per the

~

trethodology:of.1SA S67.04-1982. , PSC's earlier letters of June 21, 1985 and l l

w May 15, 1986, proposed a number of additional changes'related to upgrade c:nsiderations of the lechnical Specifications. These earlier changes were p-imarily a part of'an overall upgrade program to provide an improved statement of requirements consistent with the format of. Technical 1 Specifications for light water reactors. The NRC staff had a number.of comments (see NRC letter dated January 24,1986) on'the specifics of these proposed changes.

Those changes related to trip setpoints are safety significant' in that  ;

I the current specification requirements do not include adequate margins for instrumentation uncertainty. Therefore, these changes per NRC direction 3 vere resubmitted in PSC's letter of May 15, 1986, as a proposed ~ amendment f to Appendix A of Facility Operating License, No. DPR-34. After additional comment by. the NRC Staff in letters dated October 16, 1986 and November 26, 1986, PSC resubmitted the setpoint related changes in their letter of ,

i A ugust 28, 1987.

Based on evaluation of PSC's res'ubmittal, it is concluded that the  ;

proposed changes related to the trip setpoints for the plant protective i i

systems are acceptable with the exceptions noted below. -)

Also, it .is concluded that the remaining issues related to upgrade considerations, and a few separated setpoint issues, although still 3 outstanding, may be pursued on a separate schedule. In fact, almost all of

- these remaining issues are not part of the proposed changes in the Licensee's latest resubmittal but are deferred items f rom the Licensee's I earlier submittals of June 21, 1985 and May 15, 1986.

i

l I

l 41 l

L____ _ _ . . . . _ _ . _ _ . _ . . _ _ _ _ . _ _ . . . _ _ . . _ . _ _ .______ __.__ _.____.______._ _ __ ___ _ _ __

a e Below is a summary of,the: specific conclusions reached in Sections 2, I

Discussion and Evaluation and 3, Remaining Issues Associated with the Plant Protective system Instrumentation.

4.1 Proposed Changes Judged Acceptable  ;

The proposed Technical Specification' changes in Enclosure 2 of

- Attachment 2 of, the Licensee's letter of August 28, 1987'were found acceptable except for certain specific items that are'noted in the next.

Section 4.2 of this report. Included in these acceptable proposed changes'

- are.TS Section 3.3, Limiting Safety System Settings (pp. 3.3-1, 3.3-2a, b,-

c, 3.3-3a, b, 3.3-4, 5, 6, 7, 8', and 9) and Section 4.4, Instrumentation and Control Systems-Limiting Conditions for 0peration (pp. 4.4-1, 2, 3a, b, c, 4a, b, c, d,.Sa, b, c, 7a, b, 8, 10, a, b, c, 11, a, 12, a, b, c, and

~13). These changes basically replaced the existing Trip Setting with a Trip Setpoint and Allowable Value to account for instrumentation inaccuracy per the methodology of ISA S67.04-1982. Other related changes were the addition of definitions-for Trip Setpoint and Allowable value and clarified and expanded Bases and reformatting.

Other acceptable changes in these TS Sections not directly-related to instrumentation inaccuracy were: on p. 4.4-2 clarification of an action

.that LCO 4.2.10 and 4.2.11 are applicable when the PPS moisture monitor trips are disabled; also, on p. 4.4-2 the action for an inoperable channel

.of circulator trip was expanded to allow reactor shutdown or circulator trip; note (e) on p. 4.4-8 was updated to correctly describe the undervoltage system design; addition of 'a note (h2) on p. 4.4-8 to recognize LCO 4.9.2 as an existing bypass condition for the Primary Coolant Moisture High Level Monitor; addition on p. 4.4-4d of "less than 30% rated power" as a bypass condition on High Differential Temperature Between Loop-1 and Loop 2; deletion of the

• footnote on p. 4.4-5c on the l Circulator Speed-High Water trip; and editorial changes, j 42

e. e ->:

,4- 2 proposed Changes Judged Not Acceptable-1 o In. Figure 3.3-2, add "value" tofthe word " allowable."

o. In Table'3.3-1, Items 2.c), 2.d), and 2.e), express the. Trip'

.Setpoint and Allowable Value in terms of either greater' than or less than as well as equal to the value specified and not in terms of a value with a plus or minus or just a single value.

o On p. 3.3-5 under " general methodology" clarify the definition of

" Allowable Value," delete the words "the greater.value of,"

clarify the paragraph discussing the margin between the Trip 1 setpoint and the Analysis Value, and change " accuracy or drift"  !

to " accuracy and drif t" or justify why ISA 567.04-1982 methodology was-not followed for these differences.

o. For the Helium Circulator Penetration Interspace Pressure, make the rupture disc lower limit burst pressure consistent between Table 3.3-1, p. 3.3-2c, Item 2.d) and the value in the basis on

p. 3.3-9.

i o For Steam Generator Penetration Interspace Pressure, make the .

rupture disc' lower limit burst pressure consistent between  !

Table 3.3-1, p. 3.3-2c, Item 2.e) and the value in the basis on I'

p. 3.3-9.

o In Table 4.4-4, Part'2, Items 3a. and 3b., under the column headed " Permissible Bypass Conditions" change "above 30% rated power" back to "none."

o For the Circulator Speed-Low circulator trip, supply the programmed curve as a function of feedwater flow.

1 43 1

_J_L. _

.D: f *i : s' i l

o' In Table 4.4-4,. p. '4.4-7a, Items 3a. and 3b. , account for-

..in'strument inaccuracy in the Trip Setpoint and Allowable Values D.P ' l per. the' IS A .S67.04-1982 methodology.  ;

i o In Table 4.4-2 (Part 1), Items 7a.. 7b., and 7.c., add the footnote "p" des'ignation under the column titled " Functional Unit."

' I.

o Provide Trip Se'tpoints, Delay Times, and Allowable Values'for Degraded Voltage, Loss of Voltage Automatic Throw Over (ATO), and -

Loss of Voltage-D.G. Start, Load Shed and Load Sequence.

4.3. Other Recommended Changes as a Result of This Review-

.These other change's are recommended as a result of.the reviews performed but these changes are not necessary for approval of the proposed Technical Specification changes, o In the FSAR make the " steam in PCS" consistent between the cases 3,- 5, end 6 in Table 14.5-3 and Figures 14.5-3, 14.5-5a, and 14.5-6, respectively.

o In FSAR Section 7.1.2.4, for Low Superheat Header Temperature clarify that in order to get the trip, coincidence is required with High Differential Temperature Between Loop 1 and Loop 2 in TS Table 4.4-3, and consider deleting Item 7c. and making High Differential Temperature Between Loop 1 and Loop 2 an explicit coincidence requirement in each of Items 7a. and 76.

4.4 Remaining PPS Issues The following issues have been separated out of the original proposed changes of June 21, 1985 mostly per written direction in the NRC letter of January 24, 1986 and per subsequent telecommunications on July 30, 1986 and L  !

i 44 l,

E-_-_ __2-_-___ _

~yit..

1<

Fp-July 15,1987 (see Section 3 of this report for a detailed discussion).

LThese issues are to be addressed in future submittals, o' Provide analyses to account for instrumentation inaccuracy per ISA 567.04L1982 methodology for the circulator trip function Programmed Feedwater Flow-Low. Also, justify any intended-permissible bypass condition such as "below 30% power."

o Pesubmit the analyses to account for instrumentation inaccuracy per.JSA S67.04-1982 methodology for the circulator trip function Fixed Feedwater Flow-Low. This resubmittal should resolve the NRC Staff's concerns'with the superheater II high inlet temperature reached due to a hot gas streak penetrating the entire economizer-evaporator-superheater and should justify the permissible bypass condition for this trip function of "below 30%

power."

o Provide analyses to account for instrumentation inaccuracy per ISA $67.04-1982 methodology for the rod withdrawal prohibit trip setpoints and allowable values in Table 4.4-4 (Part 1), Items 3a.

and 3b. as minimum, p. 4.4-7a.

o Provide analyses to account for instrumentation inaccuracy in the Permissible Bypass Conditior.s for the PPS trip. The PPS l Permissible Bypass Conditions themselves also need justified.

o Provide upgrade to STS standards for setpoint and allowable values for: ,

(i) Wide Range Channel Rate of Change-High Scram, Primary Coolant Moisture High Level Monitor and Loop Monitor Loop shutdown j and rod withdrawal prohibits for Startup Channel Rate of Change-High, l i

45 I

I 1

ois o

l. ~

~l l

l' Wide Range Channel Rate of Change-High D1 k. I Linear Channel-High Power RWP (Channels 3, 4, 5, 6, 7, and 8) and Multiple Rod Pair Withdrawal, (ii) Provide upgraded PPS surveillance and calibration requirements.

l l

1 l

l l

l l

46  !

1 l

w r te . .j 1 -

5. REFERENCES 1; 0.'R.' Lee letter to.E. H. Johnson, " Proposed Changes to Sections 2.1, 3.3, 4.0, 5.0, LCO 4.4'.1, and SR,5.4.1 of.the Fort St. Vrain Technical Specifications," Public Service Company of Colorado, P-85214,

' June 21, 1985.

.2. R. F. Walker letter to H. N. Berkow, " Technical Specification Change Request to the Plant Protective System Trip Setpoints," Public Service Company of Colorado, P-86279, May 15, 1986.- <

3. R. O. Williams'. letter to Jose A. Calvo, " Technical Specification i Change Request to the Plant Protective System Trip Setpoint>," Public Service Company of Colorado, P-87278, August 28, 1987,

4. ISA-S67.04, "Setpoints for Nuclear Safety-Related Instrumentation Used in Nuclear Power Plants," Instrument Society of America,.1982.

i

5. H. N. Berkow letter to' R. F. Walker, " Fort St. Vrain-Plant Protection System Trip Setpoints," Office of. Nuclear Reactor Regulation, U.S.

Nuclear Regulatory Commission, January 24, 1986.

6. Kenneth L. Heitner letter to R. O. Williams, " Request for Additional Information for Plant Protective System Trip Setpoints and l '

Surveillance Requirements for Fort.St. Vrain Nuclear Generating-Station," Office of Nuclear Reactor Regulatory, U.S. Nuclear Regulatory Commission, October 16, 1986.

7. Kenneth L. Heitner letter to R. O. Williams, " Plant Protective System Setpoints," Office of Nuclear Reactor Regulation, U.S. ' Nuclear Regulatory Commission, November 26, 1986.

8. NUREG-0452, Rev. 4, Standard Technical Specifications for Westinghouse Pressurized Water Reactors, published by the Division of Licensing, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Fall 1981.

9. D. Warembourg letter to J. T. Collins, " Fort St. Vrain Plant Protective System Technical Specifications," Public Service Company of Colorado, P-84078, March 9, 1984.

10. Kenneth L. Heitner letter to R. O. William, " Fort St. Vrain Nuclear Generating Station, Amendment No. 52 to Facility Operating License ,

No. DPR-34," Office of Nuclear Reactor Regulation, U.S. Nuclear i

I Regulatory Commission, April 6,1987.

13. H. L. Brey letter to Jose Calvo, " Final Draft Upgrade Technical Specification Sections 3/4.8, Dated November 30, 1985,'" Public Service Company of Colorado, P-87272, August 24, I!87.

I 12. O. R. Lee letter to H. N. Berkow, " Upgraded Technical Specifications,"

l .: Public Service Company of Colorado, P-85448, A"a'abr 27,1985, 1

47 i

--~"--- - . _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ , _