ML20151B003

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

ML20151B003
Person / Time
Site:	Fort Saint Vrain
Issue date:	03/31/1988
From:	Stachew J EG&G IDAHO, INC., IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY
To:	NRC
Shared Package
ML20151A991	List: ML20151A988 ML20151A992 ML20151A996 ML20151B003
References
CON-FIN-D-6023 EGG-NTA-7289, TAC-47416, NUDOCS 8807200198
Download: ML20151B003 (51)

Text

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I EGG-NTA-7289 i March 1988

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INFORMAL REPORT

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TECHNICAL EVALVATION REPORT FOR THE PLANT National- -

-: PROTECTIVE SYSTEM TRIP SETPOINTS FOR Eng/neering , ' 'Ji FORT ST. VRAIN NUCLEAR GENERATING STATION Laboratory .

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DISCLAIMER This book was prepared as an account of work sponsored by an agency of the UNted States Govwnment. Neither the Uruted States Govenme.t nor any agency thereof, nor any of thew employees, makes any warranty, express or wnched, or assumes any legal haoihty or respons,tHhty for the accuracy, comp 4teness, or usefulness of any information, accaratus, product or process disclosed, or represents that its use would not minnge pnvately owned nghts. References herein to any specife commerc$

product, process, or serwce by trade name, trademart, manufacturer, or otherwise, does not necessardy constitute or imply its endorsement, recomtnendation, or favonng by the United States Government or any agency thereof. The views and opiruons of authors expressed herein do not necessanly state or reflect those of the United Statte Government or any agency thereof.

D. ,.

.. LEGG-NTA-7289 TECHNICAL EVALUATION REPORT.FOR THE

' PLANT PROTECTIVE SYSTEM TRIP SETPOINTS FOR FORT ST. VRAIN NUCLEAR GENERATING STATION, 4

J. C. Stachew Published-March 1988 Idaho National Engineering Laboratory EG&G. Idaho, Inc.

Idaho Falls, 10 83415 .

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. Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C. 20555 i Under DOE Contract No. DE-AC07-76I001570 FIN No. 06023 1

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ABSTRACT ,

This EG&G Idaho, Inc., report evaluates submittals provided by Public Service Company of Colorado for the Fort St. Vrain Nuclear Generating Station. The submittals are in response to requests that the trip setpoints specified in the Technical Specifications should account for instrumentation uncertainties.

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 Assistance.

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

Docket No. 50-267 TAC No. 47416 ii

CONTENTS ABSTRACT .............................................................. ii FOREWORD .............................................................. 11

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

2. DISCUSSION AND EVALUATION ........................................ 3 2.1 Methodology ................................................ 3 2.2 Evaluation of Reanalyzed Trip Setpoints .................... 3 2.2.1 Primary Cool ant Pre s sure - Low . . . . . . . . . . . . . . . . . . . . . 4 2.2.2 Primary Coolant Pressure - High .................... 5 2.2.3 Superheat Header Temperature - Low ................. 7 2.2.4 C i rc ul a to r S p e ed - Low . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 2.2.5 Fi xed Feedwate r Fl ow - Low . . . . . . . . . . . . . . . . . . . . . . . . . 9 2.2.6 Loss of Circulator Bearing Water ...... ............ 10 2.2.7 Circulator Speed - High ............................ 11 2.2.8 Neutron Flux - High .............. ................. 12 2.3 Evaluation of Proposed Technical Specification Changgs ..... 13 2.3.1 Limiting Safety System Settings (LSSS)

(Section 3.3) .................................... . 13 2.3.2 Protection System Instrumentation, Limiting Conditions for Operation (LCOs) (Section 4.4.1) . . . . 16

3. REMAINING' ISSUES ASSOCIATED WITH THE PLANT PROTECTIVE SYSTEM INSTRUMENTATION .................................................. 30 3.1 Ci rculator Trip on Programmed Feedwater Flow - Low . . . . . . . . . 30 3.2 Circulator Trip on Fixed Feedwater Flow - Low . . . . . . . . . . . . . . 31 3.3 Rod Withdrawal Prohibit at 30% Rated Thermal Power ......... 31 3.4 PPS Permissible Bypass Conditions .. ....................... 32

. 3.5 Technical Specification Upgrade Program Related Changes .... 32 3.5.1 TS Section 2.0, Definitions ..... ..............'.... 33 3.5.2 TS Section 4.0, limiting Conditions for Operation .......................................... 34 3.5.3 TS Section 5.0, Surveilitnce Requirements .......... 34 iii

3 5.4 TS Section 4.4.'1, Plant. Protective System. - .

Instrumentation LCOs ............................... ~34 3.5.5 TS Section-5.4.1,. Plant Protective System Instrumentation Surveillance and Calibration Requirements ......................................, 36

4. CONCLUSIONS ...................................................... 37 4.1 Propo:ed Changes Judged Acceptable . . . . . . . . . . . . . . . . . . . . . . . . . 38 4.2 Other FSAR Recommended Changes as a Result of Thi: Review ................................................ 39-4.3 Remaining-PPS Issues ....................................... 39-

5. REFERENCES ............................................... . ..... 42 ,

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. U TECHNXCAL EVALUAT!ON: REPORT FOR THE 1 i

PLANT PROTECTIVE-SYSTEM TRfP SETPO!NTS FOR FORT ST. VRAIN NUCLEAR GENERATING STATION'

1. INTRODUCTION 4

By letters dated June 21, 1985,1 May'15, 1986,2 August.28, 1987'3 and February 8, 19884 the Public Service Company of Colorado (PSC) p'r oposed numerous; changes.to the Technical Spect fications- (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 revised trip setpoints that include an additional margin of conservatism to account for instrumentation uncer^ainties. The revised trip setpoints were determined using as guidance Instrument Society of American Standard S67.04-1982,5 "Setpoints for Nuclear Safety-Related Instrumentation used 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 l operating conditions.

This Technical Evaluation Report provides an evaluation of the proposed trip setpoints and the reanalysis provided to reduce potential for j inadvertent safety actions, as transmitted-in PSC's revised letter of l February 8, 1988 and as supplemented by the earlier PSC submittals. The earlier PSC submittals were responded to by NRC letters dated January 24, ,

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1986 October 16, 1986,7 November 26, 1986,8 and November 25, .*987.9 The NRC letter of January 24, 1986 recommended that l

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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 Licensee's letter of August 28, 19873 responded to the NRC's seventeen issues. The NRC letter 9

of November 25, 1987 evaluated the PSC August 28, 1987 submittal and included a draft Technical Evaluation Report that identified continuing problem areas. A meeting was held with the Licensee on December 3, 1987 during which many of the issues were resolved and the resolutions are summarized in the PSC letter of February 8, 1988. Also, the PSC letters of August 28, 1987 and February 8, 1988 continue to rely on information presented in the earlier PSC letter of June 21, 1985.1 Many PPS functions presented in the PSC Ju'e n 21, 1985 letter were deleted in-the latest August 28, 1987 and February 8, 1988 submittals. 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 sctpoint 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.

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,b the Westinghouse STS,10 the NRC Staff draft Safety Evaluation Report (SER) in letter dated January 24, 1986,6 and other Licensee supplied documentation (PSC letters of March 9, 1984,11 June 21, 1985,1 May 15, 1986,2 and August 28, 1987,3 and February 8, 1988.4 .

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2. DISCUSSION AND EVALUATION 2.1 Methodoloay The Licensee submittal of February 8,1988 made proposed changes to 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 equipment protection is assured. By letter dated March 9, 1984 11 the Licensee provided a copy of a specification outlining the reevaluation of the Plant Protective System sespoints to account for instrumentation inaccuracy. This Licensee document incorporates the requirements of ISA Standard S67.04-1982, for establishing trip setpoint values. Therefore, the Licensee has established a methodology which is acceptable for determining Trip Setpoints and Allowable Values based on safety analyses for the Fort St. Vrain N clear Generating Station as documented in the FSAR.

2.2 Evaluation of Reanalyzed Trip Setooints Attachment 3 to the Licensee's letter of June 21, 1985 provided a Significant Hazards Consideration Analysis that addressed the results of new analyses for selected safety functions. The Licensee's letter of February 8, 1988 repeated these hazards consideration analyses except for the Fixed Feedwater Flow-Low Circulator Trip and Neutron Flux-High Scram.

These later two trips are still reviewed here for completeness. The conclusions of these analy:es were previously evaluated by the NRC Staff in Refe'rence 6 and have been updated here to be current with FSAR Rev. 5.

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2.2.1 Primary Coolant Pressure - Low -

The present setpoint for the low primary coolant pressure scram is programmed with load (circulator inlet temperature) to initiate scram when reactor coolant pressure is 50 psi celow normal. The low primary coolant pressure scram provides protection for inadequate core cooling that could result in temperature limits being exceeded. For rapid depressurization accidents, a scram would occur instantaneously, and changes in the low pressure setpoint would not have an impact on the consequences of the accident.

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 Sect'ons 4.3.5 and 14.8. For this accident, which is assumed to occur at 100% pcwer, and as currently anslyzed 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 90 psi below normal, 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 above normal. After the reactor scram, core average outlet temperature decreases with continued core cooling.

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

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A 4 j,.-- L.p.,. 4 4 a, ri Jg _.4a It was concluded that, since neither a safety limit nor an equipment' I .

design limit is. exceeded, the assumption of a lower. primary coolant pressure for initiation.of a' reactor scram is acceptable. l 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

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new trip setpoint is 68.6 psi below 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.

and dumo 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 sollows 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 4 the wrong loop is dumped. The only safety action initiated on 4

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 j graphite. With the assumptior. of a higher pressure trip (70 psi above normal) the depressurization program is initiated at i 120 see with a total steam ingress of 15,000 lb and-there is no ,

change in the amount that reacts with core graphite.

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2. FSAR 14.5.3.4 Case 4 - Subheader Rupture eith Moisture Monitor ,

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Failure and Correct-Loop Dump. It is assumed that no safety actions are initiated by the moisture monitors. On high primary coolant pressure, a reactor scram is init?ated, and the l 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

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of which 855 lb react with core graphite. With the assumption of a higher pressure trip (70 psi above normal) safety action is initiated in 157 see 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 Hoisture Monitor Failure and Wrono Looo 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 faulty 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 p.rimary coolant system. In the current analytis, 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, i

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the total steam ingress is 9,200 lb of dich 1,200 lb reacts with I core graphite.

The overall inpact of the change from 53 psi to 70 psi above normal for the high primary coolant pressure trip is an increase of ab. , 30% in the amount of moisture that reacts with core graphite in those cases for which nultiple failures of the protective system are assumed. While the impact of increased steam / graphite reaction was not specifically analyzed, '

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 cha'nge in the peak primary coolant pressure which could challenge the primary coolant systa relief valve rcoture 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 prest.ure.

2.2.3 Superheat Header T.mperature - Low Low superheat header temperature initiates a loop shutdown at a >

present setpoint of 800*F coincident with high diff>cential tem;2 rature 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 assun.ad that the trip on loop superheat ten erature is initiated at a superheat temperature of 780*F with a differential between loops of 65'F or ,

greater. The impacts of these assunptions were considered for two cases:

30% power an:1 100% power.

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There are tro basic consideratiors that are applicable to this safety equipment protection function. First the trip should be initiated prior to reaching floodout temperatures. Since the saturation temperature at normal operating pressure of 2400 psig is 660*F, the assumption of 780*F for mitigative action'provides an adequate margin of' safety prior to reaching the saturation temperature. The second consideration ~is that loop shutdown should occur before a turbine trip is initiated on low main steam temperature. 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 resalting 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 controlled 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 tur6ine mixed inlet steam temperature will then be 830*c. wnich assures that the loop temperature di....ence 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 temoerature is controlled at 1000*F. For this case, the temperature difference between loops is 220*F, and tFe main steam temperature is 890*F when the trip occurs. Thus, the available margins are greater than at 30%

power.

Based on this vsview, it is concluded that this analysis provides an acceptable basis to justify a change in the bases for determining the -

setpoint fo' tnese protection system channels. With the allowance for instrumecc uncertainty, the new trip setpoints are 798*F for low superheat header temperature at a 44.8*F differential temperature between loops.

2.2.4 Circulator Soeed - Low d

The present setpoint for the low circulator speed circulator trip is 1910 rpm below normal, as programmed by load (feedwater flow). The l

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circulator trip results in a reduction in plant lead'when operating at full

load conditions. Also the low feedwater flow setpoint, which i programmed by circulator speed, is lowered to preclude a trip of the oper'. ting circulator. Under conditions for ' single circulator operation the ratio of circulator speed to feedwater flow is about a factor of two g. eater than during normal operation.

For the reanalyzed case, it was assumed that a trip does not occur until a reduction of circulator speed occurs to 2390 rpm below normal. The coastdown from 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 addi'. ion, the trip includes a fixed 5 second delay to avoid spurious trips 6 e to changes in circulator speed during normal operation. In contetst, the response of the steam generator superheat header tempertcure to changes in helium flow is about 30 seconds. There u re, it was concluded that the assumption of a circulator trip at 2390 rpm below normal is acceptable.

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, 1986,6 this setpoint is not being changed in the proposed amendment request (Referenen 4). The discussion below is enclosed only for

, completeness and pertains to the PSC letter of June 21, 1985. l 1

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

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

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Two basic operating conditions tere addressed in the revised ~ analysis I

to support an assumption that the fixed low feedwater flow trip occurs a't 5% of rated feedwater flow. The first condition addressed a sudden total loss of feedwater flow to a steam generator during both one and two loop operation. Under such conditions feedwater flow is reduced to zero flow instantaneously. Due to a built-in 5 sec delay, loop isolation occurs 5 see following the occurrence of these events. Under this condition the consequences 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 boilirg 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 cottom with no mixing, it cannot be concluded thtt 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.

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

2.2.6 Loss of Circulator Bearing Water The present circulator trip on the loss of bearing water is initiated when the bearing water differential pressure, with respect to primary coolant pressure, is reduced to a low differential pressure of 475 psid.

This provides protection for the circulator bearings on a loss of the l normal and backup bearing water supply systems. In addition to a trip of l the helium circulator, the protective action includes the actuation of the l 10 l 1

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bearing tater accuculators to provide a source of bearing tater duriGg I circulator coastdown and operation of the circulator brake and seal system, t

as well as isolation of the circulator auxiliary, system service lines. The latter ensures the integrity of the primary coolant system when the dynamic seal provided by the bearing water system is not available.

The reanalysis of the operation of the loss of bearing water protection was undertaken based on the assumption .that the safety action is 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 minimum clearance of 0.001 inches will be maintained. ,

Based on this review, it is concluded that an acceptable basis has been pecvideo 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 1 trip of the helium circulator steam turbine drive was 11,000 rpm. This provided protection to assure that the circulator did not 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 er 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 transient overspeed for a loss of restraining torque event (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 of 13,267 rpm. Based on these analyses, it is extrapolated that Ih assumed 11

overspeed trip at 11,700 rpm would result in a maximum transiene, overspeed ,

of 13,370 rpm or less. The 11,700 rpm trip value was subsequently approved 12 by-the NRC Staff in Amendment No. 52 to the Facility Operating Licensee.

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 i 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 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 administratively 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 i protective actions and to note that these settings are te be established 4 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 cefine these setpoints were to have been retained in the subsequent PSC recubmittal 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 l 3 4 submittals of August 28, 1987 and February 8, 1988 the setpoint curve j was included for the high neutron flux scram. The Linear Channel-High Power RWP (Channels 3, 4 and 5) a s deleted in these last submittals per NRC direction to focus on only PPS functions that exist in the current FSV Technical Specifications, l 12

Based on the above evaluation, it'is concluded.that the neutron I flux-high scram trip setpoint and allowable'value g, resented in TS Figure 3.3-1 meets +.he intent of accommodating instrumentation inaccuracy.

2.3 Evaluation of Proposed Technical Specification Changes 2.3.1 Limiting Safety System Settinos. (LSSS) (Section 3.3)

The Licensee letter of February 8, 1988 .4proposed 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, and 3.3-8. These revised pages replaced existing pp. 3.3-1, 2, 3, 4, 5, 6, 7 and 8. Bacause of the protracted effort involved with the proposed amendment request, some issues have been deferred and others would more appropriately be addressed in future correspondence. Therefore, the evaluation below has focused primarily on only the issue of accounting for instrumentation inaccuracy. Observations have been made that are intimately connected with the PPS trip setpoints -

but these observations have been segregated from the central issue of accounting for instrumentation inaccuracy in the proposed amendment

request. Evaluation of the individual changes is given below.

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

In Table 3.3.-1, Limiting Safety System dettings, a Trip Setpoint and Allowable Value are specified for each scram, loop shutdown / steam water l l dump, and pressure relief trip function. Figures 3.3-1 and 3.3-2 were added for the Linear Cha.1nel-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 i

13 l

programmed eigh circulator in et tempes;.ture. These setpoints and S

allowable values are as presented by PSC in their le..er of June 21, 1985 and as updated to respond to the NRC letters of October 16, 19867 anc November 26, 1986.0 These latter N; letters recemmended that P,C distinguish between all Trip Setpoints and Allowable Values by accounting for setpoint tolerance and instrumentation drift based on the annuz,1 or refueling interval measured drift.

On op. 3.3-4 to 3.3-8 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. It has been clarified in the bu es that the minus tolerance on the setpoint, minimum setpoints, in Table 3.3-1,. Item 2c), 2d), and 2e) are to protect the assumption in the safety analysis that the subject pressure reliefs will not be actuated for certain classes of potential accidents and that there will be no fission product release from the primary coolant system from these accidents. The easis 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 lette of June 21, 1985.1 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 to Section 3.3 of the February 8,1988 PSC amendment request are acceptable.

The following comment (NRC Request 3) to Section 3.3 is provided for PSC's consideration for the FSAR but is not required for approval of the subject February 8, 1988 amendment reque n.

I NRC Request .1, The PSC response in letter dated August 28, 1987,3 resolves the 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

"Allowance should be made for graphic artists' %olerance.in I

transcribing data to curves. Other possible causes of minor apparent i discrepancies are that in some cases the steam graphite reaction may

. not be completed at the time of cut-off at the right side of the 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 TABLE 14.5-3. STEAM IN PCS Difference Total H 2O Total H 2O of Inleakage (nleakage Reacted and Reacted Figure-Steam in PCS i Case (Ib) (ib) (16) (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 7,161 6,800 (Figure 14.5-6)

~.

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. In the PSC/NRC meeting of December 3, 1987, PSC agreed to )

include resolution to this issue in the next FSAR revision.  !

1 1

15

2.3.2 Protectica System Instrumentation, Limiting Conditions for Operation I'

(LCOs) (Section 4.4.1)

The Licensee letter of February 8,1988 4proposed 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, 4b, 4c, 4d, Sa, 5b, 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. 4.4-64, 6b, and 6c on the Steam Line Rupture Detection and Isolation System (SLRDIS) are unchanged as is p. 4.4-9. Because of the protracted effort involved with the proposed amendment request, some issues have been deferred and others would more appropriately be addressed in future correspondence. Therefore, the evaluation below has focused primarily on only the issue of accounting for instrumentation inaccu+acy. Observations have been made that are intimately connected with the PPS trip setpoints but these observations have been segregated from the central issue of accounting for instrumentation inaccuracy in the proposed amendment request. Eva'uation of the individual changes is given below.

The added definitions on p. 4.4-1 of Trip Setpoint 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 it 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.

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 16

Redundancy, and Permissible Bypass Conditions. Primarily, the changes are.

to account for instrumentation inaccuracy as presented by PSC in their letter of June 21, 1985 and as updated to respond to the NRC letters of October 16, 19867 and November 26, 1986.8 These latter NRC letters recommended that PSC distinguish between all Trip Setpoints and Allowable Values by accounting for setpoint tolerance and instrumentation drift based on the atinual or refueling interval measured drift.

On pp. 4.4-10, 10a, 10b, 10c, 11, lla, 12, 12s, 12b, 12c and 13 the Basis for specification 4.4.1 is given. The setpoint methodology for determining Trip Setpoints and Allowable Values is as described for the LSSS basis in the previous section of this report (2.3.1). Each trip for scru , loop shutdown, circulator trip, and rod withdrawal prohibit functions are described and are consistent with FSAR Section 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 S m ral of the proposed changes in Section 4.4.1 are not directly associated with accounting for instrumentation inaccuracy. Changes for

p. 4.4-2 for when the PPS moisture monitor trips are disabled have already been discussed (see previous page). The remaining changed items not directly associated with accounting for instrumentation inaccuracy are discussed below.

Table 4.4-1 (Part 1), Item 10 Plar.t Electrical System-Loss, Note (j) was deleted from the Trip Setting column and replaced with tne '

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 correctly describe the undervoltage system design. Note (e) appears for the Plant Electrical System-Loss scram function in Table 4.4-1, Part 2, under Minimum Operable Channels. Updating of Note (e) is consistent with the FSAR Rev. 5 l description of the Plant Electrical System-Loss scram function in Section 7.1.2.3 and FSAR Table 7.1-2 and is therefore acceptable. ,

l

^

17  !

\

i

8n Table 4.4-1, Part 2, 2%em 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 recogni:es an existing Permissible Bypass Condition in LCO 4.9.2. Note (hl) is unchanged from the previous Permissible Bypass Condition note (h) for the Loop Monite .

In Table 4.4-2 (Part 2), p. 4.4-4d, Item 7c., High Differential Temperature Between loop 1 and Loop 2, the Permissible Bypass Condition has been 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 Bypass 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 suomitted as Item 15 to the Enclosure of NRC letter dated October 16, 1986.7 PSC's response in letter dated August 28, 1987 stated that removal of the footnote is more conservative as the applicability is 1

,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 have been exercised and so deleted the footncte. This deletion is in the conservative direction and removes a flexibility that in retrospect was unwarranted and therefore the deletion is acceptable.

18

In Table 4.4-4, Part 1, Item 1., p.4.4-7a, the trip setpoint has been I

changed from 22.5 cps to 3 4 2 eps and an allowable value of 33 2 cps has been added. The change on trip setpoint and addition of an allowable value are both conservative and are, therefore, acceptable.

Other minor editorial changes (commas, hyphens, consistency in titles, capitali:ation, etc.) have been made but were not specifically listed in PSC's "summary of proposed changes." These editorial changes are acceptable as are the expanded bases of the scram, loop shutdown, circulator trip, and RWP functions.

Based on the above evaluations and evaluations of Section 2.1 and 2.2 of this report, it is judged that the changes to TS Section 4.4.1 proposed-by the PSC February 8, 1988 amendment request are acceptable.

The following comments are provided .for PSC's consideration but are not reqJired for approval of the subject February 8,1988 amendment request.

i permissible Byoass 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. i The following reanalyzed trip functions did not have their bypass' s conditions justified: Primary Coolant Pressure-Low, Superheat Header Temeerature-Low, and Circulator Speed-Low. The bypass cendition for Fixed Feedwater Flow-Low has already been commented on in previous correspondence (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 er whether instrumentati:n inaccuracy is f

J 19

accounted for in the value is related to bu% not direcQly par % of .

accounting for instrumentation inaccuracy in the trip settings, it is judged that this issue does not need to be resolved for approval of the subject submittal. However, PSC should pursue this broader issue in a future submittal.

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

In Table 4.4-4, Part 2, Items 2a. and 2b. , Linear Channel-Low Power RWP and Items 3a and 3b., Linear Channel-High Power RWP (Channels 3, 4, and 5 and Channels 6, 7, and 8), under "permissible Bypass Condition," the bypass condition should reflect the RWP trip defeat actuated by movement of the Interlock Sequence Switch. For example, the bypass condition for Item I should be "ISS out of startup." And for Items 2a. arid 2b. it should be ISS in power position and power level greater than 4%." And for Items 3a. and 3b. it should be "ISS in startup" and "ISS in power position and power level greater than 10%."

NRC Request 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 FSAR) clarified that it is discusseJ 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 1 and Loop 2, that FSAR 20

Section 7.1.2.4 be clarified tu explicitly state the coincidence requirement as opposed to the present less clear reference to a comparator circuit and interlock.

NRC Reouest 7. Deletion of Curve for Circulator Soeed-Low In NRC letter dated October 16, 1986, request for additional information Number 7, PSC was asked to justify deletion of reference to 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 direction that the revised amendment request should only include those parameters which now exist in the present Technical Specifications.

In addition, those new parameters would not havn 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, I 5, Primary Coolant Moisture High Level Monitor and Loop Monitor, l 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, i

21

However, Circulator Speed-Low for the circulator trip is in the existing FSV Technical Specifications. The programmed curve for this trip should be supplied as has been done for other existing FSV functions that were missing programmed curves in the existing TS but which are 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 HighinTable3.3-1, Item 2.a)]. It was agreed in the December 3, 1987 PSC/NRC ueeting that this issue would be resolved in a future submittal.

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 most 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:

i "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, 1986,6 Enclosure 3, raised the concern that there was no apparent safety analysis to justify bypass.at less than 30 percent power on circulator trips on fixed feedwater flow-low. This concern and the parallel concern for the Programmed 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 insecuracy is submitted. l l

i NRC Request 13 Rod Withdrawal Prohibit (RWP) at 30% Rated Thermal power (RTP)

The Licensee's position regarding not providing instrumentation ineccuracy for the $ 0%

3 of rated power RWP setpoint remains unacceptable. j P.6, Attachment 3 to the PSC letter of June 21, 1985,1 stated that the rod I 22

---

L_ --_a

.-=

ciqhdracal prohibits were not analy8ed as par % of the program 90 coGply ci%h the guidance of the ISA Standard $67.04, because no credit is taken for them in accident analyses. This Licensee position was challenged in NRC letter

- dated October 16, 1986 which requested additional information to clarify why, at least, the 530% 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 deleted, an operater single failure in positioning the Interlock Sequence Switch (ISS) would not bypass required reactor protection trip functions.

The NRC letter stated that:

"Without the rod withdrawal prohibit, high power operation

(>20%) could 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 1

appears to be a required safety function to prevent this l occurrence. Therefo*e, at least this function of the RWP should I have had instrument uncertainty taken into account for the setpoint. Otherwise, additional safety analyses are required to demonstrate safe operaticn 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:

"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 similar to the BWR Reactor Mode Switch (NUREG 0123 Rev. 3). The i 50% RWP is 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. FSAR

Table 7.1-6 is an analysis of improper ISS settings and the effect on rise to power.

23

"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 protactive 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. And if the operator were to neglect placing the ISS in the Power

• position and exceed 30% power, it is highly unlikely that an accident wou'd occur in tnis 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 pa ameters even though the later will coritinue to be in the Technical Specifications. Finally, the Licensee states that the turbine generator is brought on-line with the external electrics) grid at approximately 28%

reactor power,.and this action needs to ba accomplished with stability without being encumbered with a rod-withdrawal prohibit setting in the sar.:e range, and reducing the RWP setting would also intrude into the 26% to 28%

range where secondary heat balances are made (heat balances are less accurate if performed at lower power levels).

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

1. The fundamental question is whether or no*. the safety analysis for protecting against accident situations remains valid if the operator were to inadvertently oroceed to power levels above 30%

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

24

~

2. The FSV Enterlock Sequence Switch (ISS) and Reactor Moae Switch I (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 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 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 f' unction and setpoint is bypassed because the ISS is positioned to Low Power (530% 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 .l had not been analyzed to account for its instrumentation inaccuracy, but i 1

with the .SS in the correct position. The Licensee stated. that "accident 1 l'

consequences for power range accidents are analyzed at conservative upper power limits" and that "the consequences of accidents occurring from 1.ower power levels have not generally been analyzed." The Licensee needs to make  ;

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 25

that secondary heat balances are made in the 26% 2o'282$ 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

$30% 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 ci.cumstance due to the short time spent in the 30% to 34% power range. This last argument is unacceptable as the accepted reactor protection sy' stem 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 W *o exceeding 30% reactor power.

. This Licensec argument is the exact reson that the 5 0% 3 RWP setpoint should be rigorous and inclede instrumentation inaccuracy. It is the single operator error of not positioning the ISS to the power position while exceeding 30%

RTP that constitutes a single failure defeat of some of the PPS functions.

Per General Oc>1gn Criteria 19 and 20 (see Appendix C, Rev 5 of the FSV FSAR) and as argued by the Licensee in stating that these criteria are met, the PPS has high functional reliability and radundancy to 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 de not provide the same level of protection ~  !

against iiadvertent 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 I

~

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 26

Neutron Flux Change, is bypassed in the Loe Power position of the 15$'and f 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 cf 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 interlocks 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. Accounting for instrumentation inaccuracy in reactor protection system interlocks or bypasses is required by the Standard Review Plan (Chapter 7.0),13 the STS (Section 2.2),10 and IEEE Std. 279.14 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.

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 sdtpoint 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 may be at 26%.RTP i 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 appropriate solution would appear to be doing additional safety analysis to raise the 30's RTP RWP set;;oint so that even af ter the 4%

instrumentation inaccuracy is accounted for, the setpoint is still sufficiently high (say 34% R1P) so that interfering with the turbine generator and secondary heat balance is avoided.

27

O ft is recommended that the Licensee reevaluate the 5303 RTP RWP .

setpoint to account for instrumentation inaccuracy as discussed above.

However, it should also be emphasized that accounting for instrumentation inaccuracy in the setpoint would only be warranted if first an analysis is done to justify the 30% RWP setpoint itself (see previous comment on p.19 titled "Permissible Bypass Condition"). As the 30!. RWP setpoint can be reevaluated without tha - delay the inclusion of instrumentation

.. other PPS M p setpoints, it is recommended that the 30%

RWP setpoint reevaluation be handled as a separate issue. In the interim, the remaining PPS setpoints for which instrumentation inaccuracy has already been accounted for could be approved now for facility use.

480 V AC Essential Bus Undervoltage Protection Trip Setooints In PSC letter dated August 24, 1987 (P-87272),15 Attachment 2, p. 2, NRC Comment (1) and the PSC Response are:

"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 tables es Tables 3.3.1.5 and 4.3.1.5. We note that the time dial setting for Functional Unit 3 changed from 6 to 5 in the p,meess. ..The licensee should verify the correct settings of these ?ndervoltage relays tnd commit to having thne tables and notes in the upgraded Electrical Technical Speci'ications. It should include nominal setpoints and allowable ilmits (where voltage and time tolerances exist)."

"PSC Resoonse:

"This comment is associated with the PPS Technical Specification amendment and willo 'e addressed as part of the PPS submittal."

Contrary to PSC's stated response the only information on essential bus undervoltage is Item 10. of Table 4 4-1 (Part 1) where the Trip Setroint, 1278 V and 531.5 seconds, ar.d Allowable Value, 1266 V and

$35 second are listed. No Trip Setpoints, Delay Times, and Allowable 28

l Values are provided for Degraded Voltage, Loss cf Voltage Automatic Throu

~? Over (ATO), or loss of Voltage-0G Start, Load Shed and Load Sequence. This latter informatien had been presented-in Tables 3.3.1.5 and 4.3.1.5 of the November 30, 1985 Oraft Technical Specifications 16 and is still required. It was agreed during the December 3, 1987.PSC/NRC' meeting that this issue would be addressed in a future submittal. 1 e

29 e - -

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3. REMA8HING XSSUES ASSOC 8ATED 08TH THE PLANT PROTECT!VE . l SYSTEM INSTRUMENTATION l This section is provided as a summary of tha remaining issues that have developed or have bien deferred as a result of the protracted effort on the Plant Protective .ystem instrumentation Technical Specification changes. The completion of these remaining issues to be summarized.here are net 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 facilitate 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 Feedwa+.cr Flow - Low was discovered too late to complete the analysis to incorpotste 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 letter. In the June 21, 1985 letter, PSC proposed to use the existing setpoint for the interim. The NRC letter of January 24, 1986,6 agreed with this position. In the subsequent PSC resubmittals of May 15, 4

1986,2 and August 28, 1987,3 and February 8, 1988 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 567.04-1982 methodology are still outstanding for the Programmed Feedwater Flow-Low circulator trip. Also, 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).

30

3.2 Circulator Trip on Fixed Feedwater Flow - Low 6 1 The NRC Staff found unacceptable PSC's June 21, 1985 proposal to change 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 economizer-evaporator-superheater. In the interim, the NRC Staff recommended continued use of the existing setpoint of 20% of rated full load. Consequently, in the subsequent PPS submittals of May 15, 1986, August 28, 1987, and February 8, 1988 PSC retained the setpoint of 20% of 3

rated full load. In the August 28, 1987 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 than the present 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 red withdrawal prohibits. As a minimum, as discussed in detail in Section 2.3.2 of this j report the rod withdrawal prohibit trip setpoints and allowable values at 30% rated thermal power should be reanalyzed for instrumentation inaccuracy I per the ISA S67.04-1982 methodology. The rod withdrawal prohibit trip setpoints are the method for executing the Permissible Bypass Conditions in l

i 31

the scram, loop shutdown, and circulator trip functions in Part 2 of I

Tables 4.4-1, 4.4-2, and 4.4-3, respectively. The Permissible Bypass Conditions themselves are treated directly below. It is emphasized that accounting for instrumentation inaccuracy in the RWP trip settings is only warranted if first the trip settings themselves are justified by analyses (see the next section directly below on permissible bypass conditions).

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 broader than accounting for instrumentation inaccuracy in the trip setpoints, they may be pursued without holding up approval of the trip satpoint revisione discussed in Section 2.3 of this report. Although many PPS parameters were not reanalyzed in the PSC June 21, 1985 letter, their existing permissible bypass conditions have the same problem.

Namely, the bypass condition itself has never been justified nor has the value for the bypass included instrumentation inaccuracy.

3.5 Technical Specification Upgrade Program Related Changes The PSC submittal of June 21, 19851 inc'luded extensive Technical SnaH firation 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, 1986,6 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, i 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 32 i

i

~ -

resubmittals of May 15, 1986, August.28, 1987, and February 8, 1988 tho

? upgrade related changes had been deleted. As part of the overall Technical Specification Upgrade Program,10 these upgrade related changes to the plant protection system instrumentation are still desirable and are still~

outstanding.

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 letter are enumerator below. These outstanding categories of changes were also briefly discussed with PSC in the telecommunication of July 15, 1987.

3.5.1 _TS Section 2.0, Definitions 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 2.la Action 2.lb Allowable Value 2.lc Channels to Trip 2.id Minimum Channels Operable 2.le Operational Mode-Mode 2.lf Total No. of Channels 2.lg Trip Setpoint 2.1h Actuation Logic Test 2.11 Channel Functional Test 33

' Items 2.1, 2..la, 2.lb, 2.le, 2.19, 2.1h, and 2.11 have already been I

addressed in the TSUP (PSC Oraft'TS of November 30, 198516) and any. l further action is being pursued in the TSUP. Items 2.lc, 2.1d, and 2.lf do )

not appear in the Standard Technical Specifications and are considered  !

optional for any additional action by PSC.

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

3.5.3 TS Section 5.0, Surveillance Requirements The proposed changes to Section 5.0, Surveillance Requirements, were on (numbering is the same as that of the PSC June 21, 1985 letter, Attachment 7): 5.0.1 through 5.0.7. Items 5.0.2 through 5.0.7 have already been addressed in the TSUP and any further action is being pursued in the TSUP. Item 5.0.1 does not appear in the Standard Technical Specifications and is considered optional for any additional action by PSC.

3.5.4 TS Section 4.4.1, Plant Protective System Instrumentation LCOs The proposed changes to the Plant Protective System Instrumentation section related to upgrade considerations were on added trip functions and format. The added trip functions were those that appear _ in the'FSAR (Chapter 7.0) but that were not in the existing FSV Technical Specifications. These added trip functions were (numbering is the same as that of the PSC June 21, 1985 letter):

p.4.4-2, Table 4.4-1 (Part 1), Wide Range Channel Rate of Change-High I

i 34 l

p.4.4-3a, Table 4.4-2 (Part 1), Primary Coolant Moisture High Level f Monitor and loop Monitor p.4.4-Sa, Table 4.4-3 (Part 1), Circulator Speed-Low (programmed curves for-4 circulators and'2 circulators in operation) p.4.4-5, Table 4.4-4 (Part 1), -Rod withdrawal prohibit for Startup Channel Rate of Change-High, Vide Range Channel Rate of Change-High, Linear Channel-High Power RWP (Channels 3, 4, and 5) p.4.4-5a, Table 4.4-4 (Part 1), Linear Channel-High Power RWP (Channel 6, 7, and 8), Multiple Rod Pair Withdrawal Table 4.4-5 Trip Setpoints, Delay Times, and Allowable Values for Degraded Voltage, Loss of Voltage Automatic Throw Over (ATO), and Loss of Voltage-0G Start, load Shed and Load Sequence. i Per NRC direction in a telecommunication, PSC deleted these trip functions I from their subsequent PPS submittals of May 15, 1986, August 28, 1987, and February 8, 1988, r.so, the upgrade format changes to the trip tables of  !

Section 4.4.1 on channel operability, applicable modes, and act.ic.u 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 35

~

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.

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 Requirements Section provided sis type testing specifications such as Channel Check, Channel Functional Test, Actuation Logic Test and Applicable 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 Sectio:s 3.3.4 of this report. Per NRC direction in a telecommunication, PSC deleted these upgraded surveillance and calibration requirements from their subsequent PPS resubmittals of May 15, 1986 and August 28, 1987.

The upgraded surveillance and calibration requirements of Section 5.4.1 are still outstanding and require addressing by-PSC. In any 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.

36

4. CONCLUSZONS An evaluation has been made of Lthe PSC submittal of -Feoruary 8,1988 on the proposed Technical Specification changes for the trip setpoints for the plant protective system to account for instrumentatien inaccuracy per the methodology of ISA S67.04-1982. PSC's earlier. letter of June 21, 1985 proposed a number of additional changes.related to upgrade considerations of the Technical Specifications. These earlier changes were primarily a part of an overall upgrade program to provide an improved statement of requirements consistent with the format of. Technical 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 the current specification requirements do not include adequate margins for instrumentation uncertainty. Therefore, the trip setpoint changes per NRC direction were resubmitted in PSC's letters of May 15, 1986 and August 28, 1987 as a proposed amendment to Appendix A of Facility Operating License, No. OPR-34. After additional comments by the NRC Staff in letters dated October 16, 1986, November 26, 1986, and November 25, 1987 PSC resubmitted the setpoint related changes in their letter-of February 8,1988.

Based en evaluation of PSC's resu'mittal, o it is concluded that the proposed changes related to accounting for instrumentation inaccuracy in.

the trip setpoints for the plant protective systems.are acceptable.

Also, it is concluded that the remaining' issues related to upgrade considerations, and a few segregated issues intimately connected:with the PPS trip setpoints, although deservir:g of comment and still outstanding, lmay 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 newly identified items or are oeferred items from the Licensee's earlier submittals of June 21, 1985 and May 15, 1986.

1 37

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

Discussion and Evaluation and 3, Remaining Issues Associated with the Plant  ;

Protective System Instru:rentation, j l

4.1 Prorcsed Changes Judged Acceptable i

The proposed Technical Specification changes on accounting for instrumentation inaccuracy in the trip setpoints in Enclosure 2 of of the Licensee's letter of February 8,1988 were found acceptable. 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, and 8) and Section 4.4, Instrumentation and Control Systems-Limiting Conditions for Operation (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 567.04-1982. Other related changes were the addition of definitions for Trip Setpoint and Allowable value and clarified and exptnded 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 LC0 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 Circulator Speed-High Water trip; and editorial changes.

38

g.

-4.2 Other FSAR Recommended Changes as a Result of This Review These other changes for the FSAR 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, and 6 in Table 14.5-3 and Figures 14.5-3, 14.5-Sa, 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 7b.

4.3 Remaining PPS Issues The following remaining issues have been separated out of the original proposed changes of June 21, 1985 per written direction in the NRC letter of January 24, 1986, per subsequent telecommunications on July.30, 1986 and July 15,1987 (see Section 3 of this report for a detailed discussion) and per the PSC/NRC meeting of December 3, 1987. These remaining issues are to be addressed in future submittals and are not necessary for approval of the PSC February 8, 1988 proposed Technical Specification changes.

o Justify any intended permissible bypass condition'such as "below 30% power." Provide analyses to account for instrumentation

, inaccuracy per ISA 567.04-1982 methodology for the: circulator trip function Programmed Feedwater Flow-lov.

39

o Resubmit the analyses to account for instrumentation inaccuracy g per ISA S67.04-1982 methodology for the circulator trip function 1

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 permiss.ible bypass condition for this trip furetion of "below 30%

power."

o Provide analyses to account for instrumentation inaccuracy per' ISA 567.04-1982 methodology for the rod withdrawal prohibit trip setpoints and allowable values in Table 4.4-4 (Part 1), Items 3a.

and 3b. as a minimum, p. 4.4-7a. Accounting for instrumentation inaccuracy in the RWP trip setpoints is only warranted, however, if first the PPS Permissible Bypass Conditions are justified (see next comment below).

o The PPS Permissible Bypass Conditions themselves need justified.

As part of the justification, account for instrumentation inaccuracy in the Permissible Bypass Conditions for the PPS trips. '

o In Table 4.4-4 (Part 2), Items 2a., 2b., 3a., and 3b. provide permissible bypass conditions consistent with the defeats of RWP trips by movement of the interlock sequence switch.

l o Provide upgrade to STS standards for setpoint and allowable  !

values for:

1 l

(i) -

Wide Range Channel Rate of Change-High Scram, Primary Coolant Moisture High Level Monitor and Loop Monitor Circulator Speed-Low (programmed curves)

Loop shutdown and rod withdrawal prohibits for Startup Channel Rate of Change-High, Wide Range Channel Rate of Change-High Linear Channel-High Power RWP (Channels 3, 4, 5, 6, 7, and 8) and Multiple Rod Pair Withdrawal.

l 40

'(11) . Provide upgraded PPS surveillance and c'alibration I requirements,

_(iii) Provide Trip Setpoints,. Delay' Times, and Allowable Values' for Degraded. Voltage, Loss of Voltage Automatic Throw Over.

-(ATO), and Loss of. Voltage-0G. Start, Load Shed and Load ,

Sequence.

i i

i J

i 4

P 9

41

5. REFERENCES

1. O. 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 J. A. Calvo, "Technical Specification Change Request to the Plant Protective System Trip Setpoints," Public Service Company of Colorado, P-87278, August 28, 1987.

4. R. O. Williams letter to J. A. Calvo, "Technical Specification Change Request to the Plant Protective System Trip Setpoints," Public Service Company of Colorado, P-88025, February 8,1988.

5. ISA-567.04, "Setpoints for Nuclear Safety-Related Instrumentation Used in Nuclear Power Plants," Instrument Society of America, 1982.

6. H. N. Berkow letter to R. F. 'Wsiker, "Fort St. Vrain-Plant Protection System Trip Setpoints," Of fice of Nuclear Reactor Regulation, U.S.

Nuclear Regulatory Commission, January 24, 1986.

7. K. L. Heitner letter to R. O. Williams, "Request for Additional Information for Plant Protective System Trip Setpoints and Surveillance Requirements for Fort St. Vrain Nuclear Generating Station," Office of Nuclear Reactor Regulatory, U.S. Nuclear Regulatory Commission, October 16, 1986.

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

9. K. L. Heitner letter to R. O. Williams, "Draft Technical Evaluation Report Concerning Plant Protective System Setpoints For Fort St.  !

Vrain," Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory. l Commission, November 25, 1987.

10. NUREG-0452, Rev. 4, Standard Technical Specifications for Westinohouse Pressurized Water Reactors, published by the Division of Licensing,

. Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Fall 1981.

11. 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.

42

12. K. L. Heitner letter to R. 0.' Williams, "Fort St. Vrain Nuclear f , Generating Station, Amendment No. 52 to Facility Operating License No. 0PR-34," Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, April 6, 1987.

13. NUREG - 0800, Standard Review Plan for the Review of Safety Analysis Reoorts for Nuclear Power Plants, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, June 1987.

14. IEEE Std. 279, Criteria for Protection Systems for Nuclear Power Generating Stations, The Institute of Electrical and Electrorics Engineers, Inc.

15. 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, 1987.

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

Public Service Company of Colorado, P-85448, November 27, 1985.

s 43

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