ML20236V319
| ML20236V319 | |
| 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 | |
| ML20236V296 | List: |
| References | |
| CON-FIN-D-6023 EGG-NTA-7289-02, EGG-NTA-7289-2, EGG-NTA-7289-DRF, TAC-47416, NUDOCS 8712040229 | |
| Download: ML20236V319 (51) | |
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i s TECHNICAL EVALUATION REPORT.FOR THE 3 PLANT PROTECTIVE SYSTEM TRIP SETPOINTS FOR FORT ST. VRAIN NUCLEAR GENERATING STATION .g 1 1 t
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J. C. Stachew J i Published October 1987~
, .i Idaho National Engineering Laboratory EG&G Idaho, Inc. l Idaho Falls, ID 83415 1
l I i i Prepared for the U.S. Nuclear Regulatory Commission- 'i Washington, D.C. 20555 Under DOE Contract No. DE-AC07-76ID01570 . FIN No.'D6023 ! 8712040229 871125 PDR ADOCK 05000267 y, ' ' p PDR -
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l 1 . ABSTRACT This EG&G Idaho, Inc., report evaluates submittals provided by Pubite 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 werk under DOE
- ontract No. DE-AC07-761001570 FIN No. 06023- .
Docket No. 50-267 TAC No. 47416 it i
CONTENTS r ABSTRACT .............................................................. 111 FOREWORD .............................................................. 111-
- 1. INTRODUCTION ..................................................... 1.
- 2. DISCUSSION AND EVALUATION ........................................- 3 2.1 Methodology ................................................ 3 2.2 Evaluation of Reanalyzed Trip Setpoints .................... 3 2.2.1 Prima ry Coolant Pressure-Low . . . . . . . . . . . . . . . . . . . . . . . 3 2.2.2 Primary Coolant Pressure-High ...................... 5 2.2.3 Superheat Header Temperature-Low ................... 7 2.2.4 Circulator Speed-Low ............................... 8 2.2.5 Fi x ed Feedwa te r Fl ow-Low . . . . . . . . . . . . . . . . . . . . . . . . . . . 9' 2.2.6 Loss of Circulate'r Bearing Water . . . . . .. ......... 10 l 2.2.7 Circulator Speed-High ................. ......... .. 11 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)
(Section 3.3) ...................................... 13 2.3.2 Protection System Instrumentation, Limiting Conditions for Operation (LCOs) (Section 4.4.1) . . .. 19
- 3. REKAINING ISSUES ASSOCIATED WITH THE PLANT PROTECTIVE SYSTEM INSTRUMENTATION ................................. ................ 34 3.1 Circulator Trip on Programmed Feedwater Flow-Low .. ........ 34 3.2 Circul ator Trip on Fixed Feedwater Flow-Low . . . . . . . . . . . . . . . . 35 l
3.3 Rod Withdrawal Prohibit at 30% Rated Thermal Power ......... 35 1 3.4 PPS Permissible Bypass Conditions .......................... .36 3.5 Technical Specification Upgrade Program Related Changes . . . . 36 3.5.1 TS Section 2.0, Definitions ........................ 37 , 3.5.2 TS Section 4.0, Limiting Conditions for > Operation .......................................... 38 3.5.3 TS Section 5.0, Surveillance Requirements .......... 38 111 j i
3.5.4 TS Section 4.4.1, Plant Protective System Instrumentation LCOs ............................... 38 3.5.5 TS Section 5.4.1,. Plant Protective System Instrumentation Surveillance and Calibration Requirements ....................................... 39
- 4. CosCtuSIoss ...................................................... 41 4
1 l; 4.1 Proposed Changes Judged Acceptable ......................... 42 4.2 Proposed Changes Judged Not. Acceptable ...................... 43 , 4.3 Other Recommended Changes as a Result of This Review ....... 44 4.4 Remaining PPS Issues ....................................... 44
- 5. REFERENCES ......................................................... 47 i
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I I TECHNICAL EVALUATION REPORT FOR THE ! PLANT PROTECTIVE SYSTEM TRIP SETPOINTS FOR FORT ST. VRAIN NUCLEAR GENERATING STATION 4
- 1. INTRODUCTION 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 Specifications (TS) 'for the Fort St. Vrain (FSV) Nuclear Generating St'ation. The primary purpose of the proposed changes was to modify the trip setpoints for the Plant Protective System (PPS) such J 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 I ritigative actions would be initiated. The proposed cheges result in evised trip setpoints that include an additional margia of conservatism to a: count for instrumentation uncertainties. The revise: trip setooints were
- etermined using as guidance Instrument Society of Ame ':an Stancard ,
567.04-1982,4 "Setpoints for Nuclear Safety-Related Instrumentation Used I in Nuclear Power' Plants." i 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 orovide a greater margin between the trip setpoint vai.e and nomal operating conditions. l 1 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
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 l November 26, 1986 responded to the PSC submittal that made the requested ' separation (PSC letter of May 15,1986). These latter NRC letters were l requests for additional information or guidance to clarify seventeen issues l in the Licensee's May 15, 1986 letter. The present PSC letter 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. i 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 setpo, int and normal operating conditions, has been relied upon and has essentially been duplicate: here in this report. No independent evaluation was made related to .nich 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 i FSAR. Evaluation of the Licensee's justification for change was based primarily on review against the Fort St. Vrain FSAR, Rev. 5, ISA 567.04-1982,4 the Westinghouse STS,8 the NRC Staff draft Safety Evaluation Report (SER) in letter dated January 24, 1955,5 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
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- 2. DISCUSSION AND EVALUATION 2.1 Methodology The Licensee submittal of August 28, 1987 made proposed changes to I
- 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 j
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 l 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 l Licensee provided a copy of a specification outlining the reevaluation of . l 1 the plant Protective System setpoints to account for instrumentation inaccuracy. This Licenfee document inco'porates r the requirements of ISA Standard S67.04-1982, for establishing trip setpoint values. Therefore, the Licensee has established a methodology which is acceptable for i determining Trip Setpoints and Allowable Values based on safety analyses l 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 Significant Hazards Consideration Analysis that addressed the results of l 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. l l 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 3
reactor coolant pressure is 50 psi below normal. The low primary coolant ; pressure scram provides protection for inadequate core cooling that could result in temperature limits being exceeded. For rapid depressurization i accidents, a scram would occur instantaneously, and changes in the low l 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 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,
-l 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',
1 Orimary coolant flow will have been reduced to 92.5% of rated in 220 sec. And the core average outlet-temperature peaks at 44 F a:ove normal. After the reactor scram, core average outlet temperature decreases with continued core cooling.
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i 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 power-to-flow ratio only changes by 0.01 at Doth 25 and 100'. power. The impact on helium temperatu-e at the inlet to the steam generators is an increase of 9 F at 100% power and 2 F at 25% power. I 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. 4
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 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.D' (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 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 f ailures of the prox.ection system. Of the six accident cases analyzed, only four involve safety actions initiated on high primary I coolant pressure. Each case was reanalyzed as follows based on the assumption of a high pressure scram at 70 psi above nor al.
- 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 j the wrong loop is dumped. The only safety action initiated on l 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 i the safety action is initiated after about 80 sec, with a total steam ingress of 14,890 lb of which 180 lb rea:t with core graphite. With the assumption of a higher pressure trip (70 psi ;
above normal) the depressurization program is initiated at 120 see with a total steam ingress of 15,000 lb and there is no l change in the amount that reacts with core graphite.
- 2. FSAR 14.5.3.4 Case 4 - Subheader Rupture with Moisture Monitor l
l Failure and Correct Loop Dump. It is assumed that no safety actions are initiated by the moisture monitors. On high primary 5 ; l
coolant pressure, a reactor scram is initiated, and the DRAFI preselected loop dump isolates the leaking steam generator. The current analysis indicates that there is a scram and steam i generator dump in 95 sec, with a total steam ingress of 2,160 lb- i 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 sec with a total steam ingress of 3,200 lb of which 1,112 lb react with core graphite. 4 l
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- 3. FSAR 14.5.3.4 Case 5 - Subheader Rupture with Moisture Monitor l 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, i J 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 af ter the time a scram was initiated, rather than 30 min af ter the time of the j 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 i 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. 6 _ _ _ _ _ _ _ _ _ _ _ ~
DRWT 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 ! which multiple 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 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 s: ram.. 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.: i 30% power and 100% power, j 7
. DlMFI I There are two basic considerations that are applicable to this safety )
equipment protection function. First the trip should be initiated prior to i reaching floodout temperatures. Since the saturation temperature at normal ! operating pressure of 2400 psig is 660*F, the assumption of 780*F for l sitigative action provides an adequate margin of safety prior to reaching the saturation temperature. The second consideration is that loop shutdown 'j 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. j lI Since the superheat header temperature for each loop is maintained by controlling primary coolant flow in that loop, a malfunction resulting in I l low superheat temperature for one loop would not result in a change in l superheat temperature for the other loop. At 30*4 power, steam temperature j i s controlled at about 880'F. Therefore, if loop isolation occurs at a i 1 superheat header temperature of 780 F, the temperature difference will be
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100*F. The turbine mixed inlet steam temperature will tr.en be 830 F, which assures that the loop temperature difference will satisfy that portion of q 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 when the trip occurs. Thus, the available margins are greater than at 30% power. j a l Based on this review, it is concluded that this aralysis provides an acceptable basis to justify a change in the bases for determining the j setpoint for these protection system channels. With the allowance for j instrument uncertainty, the new trip setpoints are 798*F for low superheat j header temperature at a 44.8 F differential temperature between loops. l l 2.2.4 Circulator Speed - Low 1 The present setpoint for the low circulator speed circulator trip is 1910 rpm below normal, as programmed by load (fecdwater flow). The j l 8 I l 1
DPAfI circulator trip results in a reduction in plant load when operating at full j 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 j circulator speed to feedwater flow is about a factor of two greater than during normal operation. , j For the reanalyzed case, it was assumed that a trip does not occur = until a reduction of circulator speed occurs to 2390 rps below normal. The coastdown from rated speed of the circulator by 2390 rps (25%) is only a matter of a few seconds. At part load conditions, _the time to reach this i l value is about 4 seconds. In addition, the trip includes a fixed 5 second I 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 assur:: ion of a circulator trip at 2390 rpm below normal is acceptable. Based on this review, it is concluded that this analysis provides an ] l l acceptable basis to justify a change in the bases for determining the' trip l 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 I Because of the draft SER transmitted to PSC by letter dated 1 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. l The setpoint for the fixed low feedwater flow circalator trip is 20% l of rated feedwater flow. Since both circulators in a icop are tripped on low flow, this results in a loop shutdown, which provides protection against steam generator operation at tube temperatures above design values. l l 9 l
Two basic operating conditions were. addressed in the revised analysis to support an assumption that the fixed low feedwater flow trip occurs at E% 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 cperation. 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 1 feedwater flow. However, under this condition, the minimum feedwater flow rate considered was 14% of rated flow. With regard to static boiling stability conditiens, it is noted that even if unstable boiling conditions are encountered at flow rates below 18.6%, the caximum he'ium temperature ; available at the Superheat II inlet would be less than 957*F, and, thus, could not result in significantly exceeding the maximum a'lowable temperature of 952 F at the limiting tube location. Whi:e it is noted that ; 1 this analysis is conservative, sini:e it postulates that a hot gas streak
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could penetrate the entire economizer-evaporator-superheater bundle from 4 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% cf rated flow. l Based on this analysis, an acceptable basis has not oeen set forth to support the proposed change in the low feedwater flow trio setpoint. - l 2.2.6 Loss of Circulator Bearing Water l The present circulator trip on the loss of bearing water is initiated l when the bearing water differential pressure, with respect to primary coolant pressure, is reduced to a low differential pressure of 475 psid. l This provides protection for the circulator bearings on a loss of the normal and backup bearing water supply systems. In addition to a trip of the helium circulator, the protective action includes the actuation of the 10
- s. (G~? ..
bearing water accumulators to provide a source of bearing water during circulator coastdown and operation of the circulator brae aid seal system,
/ i 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.
1 . The reanalysis of the operation of the loss of bearing water protection was undhtaken based on the assumption that the safety action is initiatedatadifferentialpressureof450isid. From prior testing of the bearing water sys" tem, the mininum differential pressure during a transient response of ths system was 375 psid. From this data it is concluded that a 25 psid reduction in ,tbe trip setpoint would rd, ult' in a transient minimum differential pressure of'350 psid. Eised o$ tnis value, j analyses and tests demonstrate that the bearing acceptante criterion of a f rinimum clearance of 0.001 inches will be maintepned. l . i ! / { f Based on this review, it is concluded that an acceptable basis has i I
- een provided to justify a lower setpoint for this safety action. With an j allowance for instrument uncertainty, the new trip setpoint is 459 psid. ]
l t 2.2.7 Circulator Speed - High ! At the time of the PSC June 21, 1985 submittal, the setpoint for the l trip of the helium circulator steam turbine drive was 11,.000 rpm. , This provided protection to assure, that the circulator did not exceep the design speed limit of 13,500 rpm. For steam line ruptures down stNani of the l :irculator steam turbine, the maximum speed is 13,264 r:m 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-transient overspeed for a loss of restraining torque event (compressor i sectionbladeshedding)was13,050rpnl. 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 an assumed 11 i ;
p a i g, . everspeed trip at 11,700 rpm would rtsuit in a maximum transient overspeed $ f of 13,370 rpm or less. The 11,700 rp.a trip value was subsequently approved l N to the Facility Operating Licensee. l by the NRC Staff in Amendment No. 52 ) u ,
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't Based on this analysis and previous, approval, it is concluded that an j 5
assumed verspeed trip value of '41,700 rpm' provides an acceptable basis for ketenining the trip setpoink for ihis pr'otectics function. With the alhkmce for instrumdt uncertainty, the oterbred trip setpoint is I i 11I fl5 rpm. f r 2.2.8 Jeutron Flux - Mg l l i
\c 7 The setpoint for the high neutron flux scram is 140% of rated thermal .
power. As $ consequence of uncertainties in t'r2 react.or power measurement, .{ the setpoint for the high neutron flux scram has been administrative 1y t :ortrolled and adjusted at conservative values based on indicated reactor [ oower. The Licensee provided curves that are currently being used to control the setpoint for the hign neutron flux se' rim as well as the,high ~ ~ neutron flux rod withdrawal prohibit. In the PSC June 21, 1985, letter, i the Licensee proposed to delete the values fo'r the trip setpoints for the protective actions 'and to note tut >these settings ar'e to be established for each fuel cycle and implemented based upon the approval of the Nuclear
- acility Safety Committee. The NRC staff found that this proposal was i unacceptable since these changes potentially could create an unreviewed-l 4 safety question. Therefo,rq the curves which define these setpoints were to trave been retained in the subsequent PSC resubmittal of May 15, 1986. I However,thehighneutronIllix rod withdrawal proMW. turve was not included in the pSC May 15,' 1936, resubmittal. In the latest PSC submittal 3
of August 28, 1987 the'$etpoint curve was included f'or the high neutron l flux scram. The Linear Channel-High Powe: RWP (Channels 3, 4 and 5) was f deleted in this last submittal per NRC direction to focus on only PPS l functions that exist in the current FSV Technical Specifications.
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-X lj 9 l 5 Ba'.md on the above evaluation, it is concluded that the' neutron -;
i fTurhiqh scram trip setpoint and allowable value presented in TS. -l Figure 3.3-1 meets the intent of accommodating instrumentation inaccuracy. I 2.3 Evaluation of Proposed Technical Specification Changes , 1 l i 2.3.1 Limiting Safety System Sett knas (LSSS) (Section 3.3) The Licensee letter of Iugu'st 28, 1987 3proposed changes to i 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 replace'd 5 i existing pp. 3.3-1, 2, 3, 4, 5, 6, 7 and 8. Evaluation of.the individual ! hanges is given below. ! The added definitions for Trip Setooint and A11owa:1e Value on j
'S p. 3.3-1 clarify them as the least conservative "as left" and "as found"
.alue respectively, for a channe' ,to be considered opera:1e. 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. I In Table 3.3.-1, Limiting Safety System Settings, a Trip Setpoint and Allowable Vr.he are specifieo for each scram, loop shutoown/ steam water j dump, and pressure relief trip function. ' Figures 3.3-1 and 3.3-2 were l added for the Linear Channel-High heron Flux and Primary Coolant 3ressare Programmed 1.ow and High. Figure 3.3-1 accounts for the detector decailbe.ition for Cycle 4 as a function of indicated thermal power. Figura 3.3-2 gives the allowable high arid low primary c:olant pressure l programe.ed with circulator inlet temperature. These se: points and allowable values are as presented by PSC in their letter of June 21, 1985 , l snd as updated to respond to the NRC letters of October 16, 19866and November 28, 1986.11 These latter NPC letters recommended that PSC distinguish between all Tria Setpoints and Ailowable Va:ues by accounting : for setpoint tolerance and instrumentation drift based on the annual or l refueling interval measured drift- I
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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 scram, the allowable value is $1067 F whereas in the OCS 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 M is change addressed anywhere else in the attachment. As this function oreviously had drift accounted for, it is not obvious why the value has changed. , l Orimary 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
'irst entry should be " Allowable Value" not just the word " Allowable."
"his change was made in the PSC submittal of May 15, 1986 per N:tC direction in letter dated January 24, 1986 but was left off in the latest submittal.
NRC Request 3 The PSC .ssponse 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' tolerance in transcribing data to curves. Other possible causes of minor apparent 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 Total H 0 Total H 0 o ne e t Inleakage Reacted and Reacted cigure-Steam in PCS Case (16) (Ib) ('b) (ib) j 3 6,240 185 6,055 4800 (Figure 14.5-3) j j 5 16,040 900 15,140 15,800 (Figure 14.5-5a orb) < 6 8,080 919 7,161 6,800 (Figure 14.5-6) These differences in " steam in the primary coolant system" between FSAR l l Table 14.5-3 and the FSAR Figures are much 4 rger than what should be allcwed for graphic artists' tolerance, ant! since the Figure values for I Cases 3 and 6 are still decreasing at the t me of cut-o'f 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 coclant system" consistent between the Table 14.5-3 and Figure values for Cases 3, 5, and 6. l 15
l Reactor Vessel Pressure Limiting Safety System Setting . l 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 8 the following PSC statements of STS practice and NRC guidance (see p.1 of Attachment 3 to PSC letter dated June 21,1985) .
"Setpcints in the STS are defined as limits with either greater l than or less than, in contrast to the tolerances with plus or minus used by PSC. In addition, PSC defined a reportable occurrence as exceeding an Absolute Value, as opposed to an 1 Allowable Value. As a result, in their letter The Commission- i i
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 ' 'ysis Value." The first is:
"The remaining three factors contributing to instrument error and used to determine the Allowable Values are:
l l 16 l
The greater value of:
- a. Accuracy of components not calibrated when the setpoint is i measured; or actual drift data from calibrations.
{ 1 Accuracy of test equipment used to calibrate instruments, and I b. j
- c. Design drift allowances (including environmental effects)'on equipment accuracy."
- 1 The second is
- J "A " total inaccuracy" value which was calculated, based on the refueling surveillance frequency, was used to determine the margin I between the Analysis Value and the Allowable Value." ;
Also per ISA Standard S67.04-1982, in Item a., the accuracy and drift snould be accounted for not accuracy or drift of the subject components. The last paragraph of p. 3.3-5 states:
"The Trip Setpoint was determined by accounting for the " total i inaccuracy" of that portion of the instrument channel not tested l
during the monthly functional test plus the drif t of that portion l 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 annuai (or refueling interval) allowable l values." Also this present PSC distinction continues to ignore the '! recommendation made in the NRC letter dated October 16, 1986 in which it I was emphasized that the separation between " Allowable Value" and " Trip i 17 l 1 1
Setpoint" per the ISA $67.04 Standard is to segregate that part of the instrumentation inaccuracy that is subject to change with time, namely drift. The " inaccuracy" of the channel not tested during the monthly 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 Setpoint." Basis For Helium Circulator Penetration Interspace Pressure l 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 cpen in the ( range of 781 psig (-3%) to 829 psig (+3%) and woul: relieve at full capacity at 886 psig (10% accumulation)." Whereas in Table 3.3-1, p. 3.3-2c, Item 2.d), the " rip Setpoint is ou'oted 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 = E:8 psig in the Table. These values should be consistent. If the 17 psi is a conservative coundoff 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 841 psig. It is recommended that the Table values be left as is but the basis value be changed from 809 psig to E28 psig. Basis For Steam Generator Penetration Interspace Press.-e 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) a d the value in the basis on p. 3.3-9. 18
. 1 2.3.2 Protection System Instrumentation, 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, Protecti've System Instrumentation, Limiting Conditions for Operation. Proposed revisions were on TS pp. 4.4-1, 2, 3a, 36, 3c, 4a, 4b, 4c, 4d, Sa, 5b, Sc, 7a, 7b, 8,10,10a, J 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 individual changes is given below. The added definitions on p. 4.4-1 of Trip Setpoint and Allowable Value i 1 are as discussed earlier (Section 2.3.1 of this report)*to distinguish 1 i between "as lef t" and "as found" values, respectively. ! On p. 4.4-2, clarification is made that 1.COs 4.2.10 and 4.2.11 apply l dtving 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, new provides a choice l of either reactor shutdown or circulator shutdown rather than the previous requirement of just circulator shutdown. Reactor shutdown is a more j stringent action than just circulator shutdown and is tnerefore acceptable. In Tables 4.4-1, 4.4-2, 4.4-3, and 4.4-4 reformatt'ng was provided by splitting each Table into Part 1, containing Trip Setpcint 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 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 recommended that PSC distinguish between all Trip Setpcints and Allowable Values by accounting for setpoint tolerance and instrumentation drift based on 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 methodology for i 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 'I , ) scram, loop shutdown, circulator trip, and rod w4thdrawal prohibit j
- functions are described and are consistent with FSAR Section 7.1.2.3, l 7.1.2.4, 7.1.2.5, and 7.1.2.6 and the licensing basis and discussion l presented in Attachment 4 to the PSC letter of June 21, 1985.1 i Several of the proposed changes in Section 4.4.1 are not directly 1 associated with accounting for instrumentation inaccuracy. Some of these other changes have already been discussed for p. 4.4-2. The remaining {
items are discussed below. l In Table 4.4-1 (Part 1), Item 10, Plant Electrical System-Loss, j Nete (d) was deleted from the Trip Setting column and replaced with the Trip Setpoint and Allowable Value and correspondingly note (d) was deleted
- p. 4.t.-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 j undervoltage system design. Note'(e) appears for the Plant Electrical System-Loss scram function in Table 4.4-1, Part 2, under Minimum Operable Cnannels. Updating of Note (e) is consistent with the FSAR Rev. 5 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. In Table 4.4-1, Part 2, Item 4., Primary Coolant Moisture High Level
% nitor anc 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 Bypass Condition in LCO 4.9.2. Note (hl) is unchanged from the previous Permissible Bypass Condition note (h) for the Loop Monitor.
20 l l l
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 l 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 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 1 l ' flexibility of operable channels for only one circulator per loop would not have been exercised and so deleted the footnote. This deletion is in the l conservative direction and removes a flexibility that in retrospect was ! unwarranted and therefore the deletion is acceptable. ! 1 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 l 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 left 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. 21
Other m' inor editorial changes (commas, hyphens, consistency in titles, capitalization, etc.) have been made but were not specifically listed in j PSC's " summary of proposed changes." These editorial changes are j 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 are l acceptable with the following exceptions. Permissible Bypass Condition I i l Several trip functions in Tables 4.4-1, 2, and 3 have permissible i 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% i i 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 (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 j 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 ;
Linear Channel-High Power RWP (Channels 3, 4, and 5 and Channels 6, 7, and 8) 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 1 Bypass Condition," the bypass condition should remain as "none" and not as the proposed "above 30% rated power." 1 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 i FSAR) clarified 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." i This is the only clear indication in either the Tecnnical 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 ecommended that when the FSAR is revised to add the trip setpoint for the High Differential Temperature Between Loop 1 and Loop 2, that FSAR Section 7.1.2.4 be clarified to explicitly state the coincidence
-equirement as opposed to the present less clear reference to a comparator circuit and interlock.
1 NRC Request 7, Deletion of Curve for Circulator Speed-Low In NRC letter dated October 16, 1986, request for additional information Number 7, PSC was asked to justify deletion of reference to 23
. )
Figures 4.4-la and 4.4-1b for the circulator Speed-Low trip. PSC's f 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. j l PSC stated: 1 l
"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 have had approved surveillance requirements had we included them." i 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, l 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 , trip should be supplied as has been done for other existing FSV 24
functions that were missing programmed curves in the existing'TS but which 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)]. MRC 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 .; i Technical Specifications were left out of PSC's August 28, 1987 submittal. While this reference to Response 4 is appropriate for j l most of the comments on the Programmed Feedwater Flow-Low function in l l NRC Request 4, it is not appropriate for the question posed in the 7ast 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-Lc setpoint, but PSC did not provide or mention these latter ana"yses in j their letter."
I The NRC letter of January 24, 1986,5 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 t.:e parallel concern for the Programmed Feedwater Flow-Low function (when it is submitted) remain unanswered. l l This issue may be pursued by PSC in the future whe- the Programmed l ' Feedwater Flow-Low function analyses to support instrumentation inaccuracy is submitted. l l
'1 1
25
NRC Request 13 Rod Withdrawal Prohibit (RWP) at 30% Rated Thermal Power (RTP) The Licensee's position regarding not providing instrumentation i inaccuracy for the $30% of rated power RWP setpoint remains unacceptable. P.6, Attachment 3 to the PSC letter of June 21, 1985,I stated that the rod withdrawal prohibits were not analyzed as part of the program to comply with the guidance of the ISA Standard S67.04, because no credit is taken for them in accident analyses. This Licensee position was challenged in NRC letter dated October 16, 1968 which requested additional information to clarify l why, 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 l (Part 1) and to also, reevaluate the other RWPs to ensure that if they were deleted, an operator single f ailure in positioning the Interlock Sequence Switch (ISS) would not bypass required reactor protection trip functions. Tne NRC letter stated that: j i
"Without the rod withdrawal prohibit, high power operation
(>30%) 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 appears to be a required safety function to prevent this occurrence. Therefore, at least this function of the RWP should i have had instrument uncertainty taken into account for the j 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 i _ _ _ -_ D
"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 30% 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.
"The underlying rationale for n'ot applying instrumentation uncertainties to the RWP circuitry is that of avoiding the potential for initiating protective actions wh'en 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 setpoir.: somewhat less than 30%. By doing so, certain plant protective functi: s vould be
" enabled" prior to system operating parameters (pressures, flows, temperatures, etc.) being within normal operating condit'ons."
The Licensee further clarified that the linear and wide range nuclear instrument channels, which input signals proportionate to reactor power to j the RWp circuitry, are calibrated against a secondary. heat balance prior to l reaching 30% power (in the range of 26 to 28% power). And due to the l I 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 - I 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 I 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 :arbine generator is brought on-line with the external electrical grid at ap;roximately 28% l 4 27 l,
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% range where secondary heat balances are made (heat balances are less _i accurate if performed at lower power levels). 4 The Licensee in its response has failed to consider several aspects relating to the 530% 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 i
- 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.
l Each of these aspects will be addressed individually below. ) i l i 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 function and setpoint is bypassed because the ISS is l i l i 28 l L-_--______---_ l
positioned to Low Power ($ 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 i had not been analyzed to account for its instrumentation inaccuracy, but ! with the ISS in the correct position. The Licensee stated that " accident l consequences for power range accidents are analyzed at conservative upper l power limits" and that "the consequences of accidents occurring from lower 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
- l i
limits while demonstrating the PPS trips are adequate to ensure protection at worst case conditions does not establish at what low power level the i Likewise stating that lowering the RWP 30% RTP l trips may be bypassed. setpoint would cause dif ficulties b'ecause the turbine generator is brought I 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 l
'or allowing potential single-f ailure defeat of PPS tries. The Licensee nas 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 circumstance due to the short time spent in the 30% to 34% power range. This last argument is unacceptable as j the accepted reactor protection system practice is to provide protection over the full allowable power range without exception. The Licensee l further argued that the ISS and RWP are included as a second line of defense (or added reminder) to the reactor operator te clace the ISS in the l l
correct position prior to exceeding 30% reactor power. This Licensee i argument is the exact reason that the 530% RWP setpoint should be rigorous l 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 single failure defeat of some of the PPS functions. Per General Design 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 l PPS has high functional reliability and redundancy to 29 l
assure that no single failure will result in loss of the protection l function. As will be discussed immediately below other reactor designs j 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 l 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 ; 1 against inadvertent operation outside intended bounds as do BWR or PWR l systems. At FSV the operator may proceed to power levels above 30% RTP - I 1 with the ISS in the Low Power position and thus defeat various PPS j
\
l functions intended to be operable above 30% RTP. This same situation does 1 I l 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 tr.e ISS and therefore will not cause an automat'ic scram. In the PWR design, on.
~
increasing ieactor power, the P-6 and P-10 interlocks al' low 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 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 f I most, the operator could f ail to block the startup trips when allowed but l I these would only cause autcmatic 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 . i operable, the RWP takes on an added safety, significance for FSV to block l outward rod motion when the ISS is not correctly positioned. I 1 l 30
o 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 may 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 t.he 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% RTP RWP setpoint so that even after the 4%
% strumentation inaccuracy is accounted for, the setpoint is still sufficiently high (say 34% RTP) so that interfering with the tarbine generator and secondary heat balance is avoided. I It is recommended that the Licensee reevaluate the 530% 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 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. !
I 1 430 V AC Essential Bus Undervoltage protection Trio Set:oints In PSC letter dated August 24, 1987 (P-87272),II Attachment 2, p. 2, NRC Comment (1) and the PSC Response are: l 31
4 4 i "NRC Comment (1): DRAf[ j
" 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 g dial setting "for Functional Unit 3 changed from 6 to 5 in the process...The licensee should verify the correct settings of I these undervoltage relays and commit to having these tables and f notes in the upgraded Electrical Technical Specifications. It j should include nominal 'setp'oints and ' allowable limits (where voltage and 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 informat'on on essential ) aus undervoltage is Item '10. of Table 4.4-1 (Part 1) where the Trip { Setpoint, >278 V and <31.5 seconds, and Allowable Value, >266 V and
$35 second are listed. No Trip Setpoints, Delay Times, and Allowable j Values are provided for Degraded Voltage, Loss of Voltage Automatic Throw j i
1 Over ( ATO), or loss of Voltage-D.G. Start, Load Shed and Load Sequence. This latter information had been presented in Tables 3.3.1.5 and 4.3.1.5 of the November 30, 1985 Draft Technical Specifications 12 and is still
*equired.
High Reactor Building Temperature (Pipe Cavity) l 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 accountec for, it is not obvious why the value has changed. , 32 l
Low Superheat Header Temperature In Table 4.4-2 (Part 1), Items 7a., 7b., and 7c., a footnote "(p)" designation should be added under the " Functional Unit" description for each entry: Low Superheat Header Temperature, Loop 1 and 2, and High Differential 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 value of
$2035 rpm below normal was $1974 rpm below normal in the PSC letter of Ma9 15, 1986. As this function previously had drift accounted for, it is not obvious why the.value has changed.
Circulater Seal Malfunction In Table 4.4-3 (Part 1), Item 8. , Circulator Seal Malfunction, the allowable value of 576.1" H2O was $75.6" H2 O in the PSC letter of May 15, 1986. As this function previously had drif t accounted for, it is not obvious why the value has changed. Circulator Speed-High Water 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 this function previously had drift accountec for, it is not obvious why the value has changed. l 33
- 3. REMAINING ISSUES ASSOCIATED WITH THE PLANT PROTECTIVE SYSTEM INSTRUMENTATION This section is provided as a summary 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 tc status them and to
'acilitate future tracking of them.
3.1 Circulator Trip on Programmed Fe.edwater Flow-Low Per PSC's letter of June 21, 1985',I 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 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, PSC should provide justification for any intended bypass of this trip such as "below 30's power" (see similar comment for the Fixed Feedwater Flow-Low trip function in Section 3.2 of this report).
~
34 (
o 3.2 Circulator Trip on Fixed Feedwater Flow - Low The NRC Staff found unacceptable 5 I 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 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 i 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, i 1986, the NRC Staff requested PSC to justify the bypass condition of the Fixed Feedwater Flow-Low trip below 30% power. In a tele' communication' 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 l 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 l 30% rated thermal power should be reanalyzed for instrumentation inaccuracy per the ISA 567.04-1982 methodology. Th'e rod withdrawal prohibit trip setpoints are the method for executing the Permissible Bypass Conditions in 35
-_________m__._-
the scram, loop shutdown, and circulator trip functions in Part 2 of Tables. 4.4-1, 4.4-2, an'd 4.4-3, respectively. The Permissible Bypass-
'I 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 i i 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 l approval of the trip setpoint revisions discussed in Section 2.3 of this report. 3.5 Technical Specification Upgrade Program Re'ated' Changes The bC submittal of June 21, 1985 1 included extersive Technical Specification Upgrade Program (TSUP) relate'd changes in :ne 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 l 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 ina::uracy 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 Techr' cal Specification Upgrade Program,I2 these upgrade related changes to the plant protection system instrumentation are still desirable and are still outstanding. 36 L l -_-_-_ -
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): 1 2.1 Three Room Control Complex 2.la Action 2.lb Allowable Value 2.lc Channels-to Trip 2.1d Minimum Channels Operable 2.le Operational Mode-Mode 2.lf Total No. of Channels 2.lg Trip Setpoint - 2.1.h Actuation Logic Test 2.1.1 Channel Functional Test Items 2.1, 2.la, 2.lb, 2.le, 2.1g, 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, 2.1d, and 2.lf do
.j l
37
9 not appear in the Standard Technical Specifications and are considered' l cptional for any additional action _by PSC. ! j
'3.5.2 TS Section 4.0, Limitina Conditions for Operation 9
l 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. ]
- tems 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. ;
J 3.5.3 TS Section 5.0, Surveillance Requirements The proposed changes to Section 5.0, Surveillance Requirements, were- j
- n (riumbering 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-i Specifications and is considered optional for any additional action by PSC. 3.5.4 TS Section 4.4.1, Plant Protective System Instrumentation LCOs-i 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 (number,ing 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 p.4.4-3a, Table 4.4-2 (Part 1), Primary Coolant Moisture High Level
. Monitor and Loop Monitor 38
... j f
p.4.4-5, Table 4.4-4(Part1), Rodwithdrawalprohibitforlcantup 0 Chtnnel Rate of Change-High, Wide Range , Ch'ai.nel Rate of Change-High, Linear i Channel-High Pome RWP (Charynels 3, 4,g
\
and 5) i
, )
i p.4.4-5a, Table 4.4-4 (Part 1), Linear Channel-High Power RWP < 1 (Channel 6, 7, and 8), Multiple Rod Pair Withdrawal 7 L
$fa j '
Per NRC direction in a telecommunication.oPSC deleted these trip functions 1 y ] from their subsequent PPS submittals of May 15, 1986 and August 28, 1987. J l Also, the upgrade format changes to the trip tables of Section 4.4.1 on I channel operability, applicable modes,. and actions were deleted as (
.s discussed earlier. Other deletions and or defe,rrals not directly related I
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 :nis 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, P3C 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 Systdi.i Instrumentation Surveillance and Calibration Requirements > q The proposed changes to the PPS Surveillance and Calibration Requirements Section provided STS type testing specifications such as Channel Check, Channel Functional Test, Actuation Logic Test and Applicable Modes. These testing specifications were addad 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 directior. in a r
^
39
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- telecomunicatiori, PSC' deleted these upgraded surveillance and calibration I #
requirteentsfromtheirsubsequentPPShesubmittalsofMay 15,1986//no' August 28, 1987. . J, j# g * , k The upgraded survei nance and calibration requirements of
* Section 5.4.1 are st 11 outstanding and , req 6f re ' addressing by 9SC,y In any redabndttal of these upgraded surveillance apd calibration requirements,)
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- 4. CONCLUSIONS An evaluation haibeen sadi of the PSC submittal of August 28, 1987 on the proposed Technical Specidcation changes for the trip setpoints for the )
plant protective system to' account for instrumentation inaccuracy per the ; methodology of ISA S67.04 1982. NSC's earlier letters of June 21, 1955 and N y 15, 1986, propc.ed a number of additional changes related to upgrade j considerations ,f the Technical' Specifications. These earlier changes were primarily a part cf an overail upgrade program to provide an improved statement of rew irements consistent with the format of Technical l Specifications for light water reactors. The NRC staff had a number of I comments see NRC letter dated January 24,1986) on the specifics of these proposeuI;(hanges. c i' hose changes related to trip setpoints are safety significant in that the current specification requirements do not include adequate margins for
'nstrmentation uncertainty. Therefore. these changes per NRC direction I were resubmitted ir. PSC's letter of May 15, 1986, as a p- posed amendment to Appencix A of Facility Operating License, No. DPR-34. After additional commert. ny the NRC Staff in letters datec October 16, 1986 and November 26, 1986, PSC resubmitted the setpoint related changes in their letter of August 28; 1987.
Based on evaluation of PSC's resubmittal, it is concluded that the propc.sd . changes related to the trip setpo!nts for the plant protective systems are acceptable with the exceptions noted below. Also, it' is concluded that the remaini.ng issues related to uporade I considerations, and a few separated setpoint issues, although still outstanding, may be' pursued on i separate schedule. In fact, almost all of these renaining issas are not part of the proposed changes in the j Licensee's latest resubmit.t.nl but are deferred items f rom the Licensee's earlier submittals of June 21, .U185 and F.ay 15, 1986. I i 41' I [. . u _ _ _ _ _ _ _ _ _ _ _
Below is a summary of the specific conclusions reached in Sections 2, 3 1 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 l 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 l 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 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 j 13). These changes basically replaced the existing Trip Setting with a l Trip Setpoint and Allowable Value to account for instrumentation 'naccuracy cer 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, - l 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 j 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 l 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-Se on the Circulator Speed-High Water trip; and editorial changes.
42
d 4.2 Proposed Changes Judged Not Acceptable , Oh I o In Figure 3.3-2, add "value" to the 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 margia between the Trip setpoint and the Analysis Value, and change " accuracy or drift" to " accuracy and drift" or justify why ISA 567.34-1982 methodology was not followed for these differe*:es, o For the Helium Circulator Penetration Interspa:e pressure, make s the rupture disc lower limit burst pressure cc sistent between Table 3.3-1, p. 3.3-2c, Item 2.d) and the value in the basis on
- p. 3.3-9.
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 l
- 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." For the Circulator Speed-Low circulator trip, supply the o programmed curve as a function of feedwater flew. 43
1 F o In Table 4.4-4, p. 4.4-7a, Items 3a. and 3b.', account for . instrument inaccuracy in the Trip Setpoint and Allowable Values per the ISA 567.04-1982 methodology. . j 9 o In Table 4.4-2 (Part 1), Items 7a., 7b., and 7.c., add the k footnote "p" designation under the column titled " Functional Unit." ) o Provide Trip Setpoints, 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 changes 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, 14.5-5a, and 14.5-6, respectively. . l 1 o In FSAR Section 7.1.2.4, for Low Superheat Header Temperature l 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. ) 4.4 Remaining PPS Issues The following issues have been separated out of the original proposed 4 changes of June 21, 1985 mostly per written direction in the NRC letter of i January 24, 1986 and per subsequent telecommunications on July 30, 1986 and 44
- _ _ _ _ _ _ _ _ - _ - _ _ . _ _ - . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ ._ - _ _ _ _ - ~
- et s i
July 15,1987 (see Section 3 of this report for a'detatied discussion). ] These issues are to be addressed in future submittals. o Provide ' analyses to account for instrumentation inaccuracy- per ISA 567.04-1982 methodology for the circulator trip function Programmed Feedwater Flow-Low. Also, justify any intended permissible bypass condition such as "below 30% power." i o Resubmit the analyses to account for instrumentation inaccuracy per ISA 567.04-1982 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 ft.nction 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 minimum, p. 4.4-7a. - o Provide analyses to account for instrumentation inaccuracy in the Permissible Bypass Conditions for the PPS trip. The PPS Permissible Bypass Conditions themselves also need justified, o Provide upgrade to STS standards for setpoir; and allowable values for: i (1) Wide Range Channel Rate of Change-High Scram, Primary Coolant Moisture High Level Monitor and loop Monitor i l Loop shutdown and rod withdrawal prohibits for Startup Channel Rate of Change-High, , 45
.a .;~ .
DRAFT Wide Range' Channel Rate of Change-High LinearChannel-HighPowerRWP(Channels 3,'4,5,6,7, and 8) and Multiple Rod Pair Withdrawal. (ii) Provide upgraded PPS surveillance and calibration requirements. 46
4. r ,
- 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 Jose A. Calvo, " Technical Specification Change Request to the Plant Protective System Trip Setpoints," 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.
- 5. H. N. Berkow letter to R. F. Walker, " Fort St. Vrain-Plant Protection System Trip Setpoints," Office of Nuclear Reactor Regu.lation, 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 Surveillance Requirements for Fort St. Vrain Nuclear Generating Station," Of' ice of Nuclear Reactor Regulatory, U.S. Nuclear Regulatory C:mmission, 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, Of fice of Nuclear Reactor Regulat' ion, U.S. Nuclear Regulatory Commission, Fall 1981. l
- 9. D. Warembourg letter to J. T. Collins, " Fort St. Vrain Plant Protective System Technical Specifications," public Service Company of l
Colorado, P-34078, 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 Regulatory Commission, April 6,1987. I 1
~
- 11. H. L. Brey letter to Jose Calvo, " Final Draf t Upgrade Technical l 1
Specification Sections 3/4.8, Dated November 30, 1985," Public Service Company of Colorado, P-87272, August 24, 1987.
- 12. O. R. Lee letter to H. N. Berkow, " Upgraded Technical Specifications,"
Public Service Company of Colorado, P-85448, November 27, 1985. I 47 l
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