ML20138A320
| ML20138A320 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 04/18/1997 |
| From: | CENTERIOR ENERGY |
| To: | |
| Shared Package | |
| ML20138A317 | List: |
| References | |
| NUDOCS 9704280041 | |
| Download: ML20138A320 (25) | |
Text
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Doexst Number 50-346 License Numbar NPF-3 Serial Number 2441 Enclosure Page 1 APPLICATION FOR AMENDMENT TO FACILITY OPERATING LICENSE NUMBER NPF-3 DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 Attached are the requested changes to the Davis-Besse Nuclear Power Station, Unit Number 1 Facility Operating License Number NPF-3.
Also included is the Safety Assessment and Significant Hazards Consideration.
The proposed changes (submitted under cover letter Serial Number 2441) concern:
Appendix A, Technical Specifications (TS).
3/4.3.2.1 Safety Features Actuation System Instrumentation 3/4.5.2 Emergency Core Cooling Systems - ECCS Subsystems -
Tavg,>,280*F Appendix A, Technical Specification Bases:
3/4.3.1 and Reactor Protection System and Safety System l
3/4.3.2 Instrumentation 3/4.5.2 and ECCS Subsystems 3/4.5.3 By:
Wood [VicePreside'nt - Nuclear J.
K.
9704280041 970410 PDR ADOCK 05000346 P
PDR Sworn to and subscribed before me this 18th day of April,1997.
D No Wry Pu ic State of Ohio J. STRAUSS Notary Public. State of Ohio My Commimon Empires 3/22/98
Docket Number 50-346 License Number NPF-3 Serial Number 2441 Enclosure Page 2 The following information is provided to support issuance of the requested changes to the Davis-Besse Nuclear Power Station (DBNPS), Unit Number 1 Operating License Number NPF-3, Appendix A, Technical Specifications (TS): TS 3/4.3.2.1, Safety Features Actuation System Instrumentation, TS 3/4.5.2, Emergency Core Cooling Systems - ECCS Subsystems - Tavg 2 280*F, and associated Bases.
A.
Time Required to Implement: This change is to be implemented concurrent with related changes to be proposed by separate license amendment applications, prior to the commencement of the Eleventh Refueling Outage (11RFO). The 11RFO is pr6sently scheduled to commence in April, 1998.
B.
Reason for Change (License Amendment Request Number 96-0014):
The proposed revisions would modify presently specified 18 month surveillance frequencies in certain TS Surveillance Requirements contained in the above-mentioned TS Sections to new specified frequencies of once each Refueling Interval, or once each 24 months, based on the DBNPS Instrument Drift Study.
These changes are in accordance with the NRC guidance provided by Generic Letter 91-04,
" Changes in Technical Specification Surveillance Intervals to Accommodate a 24-Month Fuel Cycle," dated April 2, 1991, and will support conversion of the DBNPS from an 18 month to a 24 month fuel cycle.
Setpoint revisions are required by the results of the DBNPS Instrument Drift Study. These proposed revisions are based on NUREG-1430, Revision 1,
" Standard Technical Specifications, Babcock and Wilcox Plants," dated April 1995.
C.
Safety Assessment and Significant Hazards Consideration: See Attachment A.
=
l Docket Number 50-346 i
Licensa Numbar NPF-3 Serial Number 2441 Attachment A l
1 l
SAFETY ASSESSMENT AND SIGNIFICANT HAZARDS CONSIDERATION FOR
{
LICENSE AMENDMENT REQUEST NUMBER 96-0014 (137 pages follow) i
)
LAR 96-0014 Page 1 SAFETY ASSESSMENT AND SIGNIFICANT HAZARD CONSIDERATION FOR LICENSE AMENDMENT REQUEST NO. LAR 96-0014 TITLE:
Proposed Modification to the Davis-Besse Nuclear Power Station (DBNPS),
Unit Number 1 Operating License Number NPF-3, Appendix A, Technical Specifications, to Revise Technical Specifications Regarding Safety Features Actuation System Instrumentation, and Emergency Core Cooling Systems - ECCS Subsystems - Tavg 2 280
- F, for Conversion to a 24 Month Fuel Cycle.
DESCRIPTION:
The Davis-Besse Nuclear Power Station (DBNPS) Unit Number 1 is converting from an 18 month to a 24 month fuel cycle. This conversion will allow i
the DBNPS to operate at full power for a longer period of time between refueling outages.
In order to support this conversion, it is necessary that the DBNPS Operating License NPF-3, Appendix A, Technical Specifications be amended to change the 18 month interval surveillance requirements to 24 month interval surveillance requirements.
In addition, the continued application of Technical Specification (TS) 4.0.2, which allows surveillance intervals to be increased up to 25%
on a non-routine basis, will allow a 24 month surveillance interval to be extended up to 30 months.
License Amendment Request (LAR) Number 96-0014 addresses only a portion of the scope of changes required for the 24 month cycle conversion.
Additional required Technical-Specification changes have been submitted j
under separate but related license amendment applications. Associated changes to the DBNPS Updated Safety Analysis Report (USAR), including the Chapter 15 Accident Analysis, are being evaluated under the 10 CFR 50.59 process.
In accordance with 10 CFR 50.59, should this evaluation determine that an unreviewed safety question exists, the USAR changes would be submitted for NRC approval under the license amendment application process.
Related License Amendment Reauests It is noted that there are other Instrumentation TS Surveillance Requirements, presently on an 10 month frequency under current TS requirements, that are not included in this Safety Assessment and Significant Hazards Consideration, and other related License Amendment Requests (LAR) which are associated with the DBNPS conversion to a 24 month fuel cycle. These related LAR's are as follows:
LAR 96-0014 Page 2
- 1. License Amendment Request 95-0027 (DBNPS letter Serial Number 2405 dated December II, 1996) proposes a revision to TS Definition Table 1.2 redefining Notation "R" from "At least once per 18 months" to "At least once per 24 months" and defining a new Notation "E" as "At least once per 18 months."
There are several TS Table 4.3-2 revisions proposed to Surveillance Requirement intervals by this LAR (96-0014) in which the interval is defined by the "R" notation and for which it is proposed to apply the new definition of the "R"
notation discussed in LAR 95-0027.
Except for the TS Table 4.3-2 revisions proposed by this LAR (96-0014), all other TS Table 4.3-2 18 month Surveillance Requirement intervals currently defined by the "R" notation will remain at 18 months and are proposed by LAR 95-0027 to be designated by the new "E" notation.
- 2. Also, it is noted that whereas this amendment application includes a proposed change to Surveillance Requirement (SR) 4.5.2.d, there are several additional license amendment applications submitted to the NRC, LAR 95-0019, 95-0022, and 95-0027, which also affect SR 4.5.2.d.
Therefore, Toledo Edison is also requesting that amendments for LAR 95-0019, 95-0022, 95-0024, 95-0027, and this amendment application, LAR 96-0014, be issued together by the NRC.
Prucosed Revisions to Surveillapee Reauirement Inte rvals The NRC guidance provided by Generic Letter (GL) 91-04, " Changes in Technical Specification Surveillance Intervals to Accommodate a 24-Month Fuel Cycle," dated April 2, 1991, was utilized in the preparation of this Safety Assessment and Significant Hazards Consideration.
Consistent with this guidance, the phrase "At least once per 18 months" is proposed to be replaced with "At least once each REFUELING INTERVAL."
The term
" REFUELING INTERVAL" is defined by TS Definition 1.42 as "a period of time s 730 days."
It is proposed that a new definition for the "R"
notation be applied in various TS tables as delineated below. The "R"
notation is presently defirad by TS Definition Table 1.2 as a frequency of "At least once per 18 munLhs."
License Amendment Request 95-0027 proposes that the "R" notation be defined as "At least once per 24 months."
In accordance with the guidance provided by GL 91-04 an instrument drift study was performed for those instruments that have a surveillance requirement for calibration for which a plant shutdown is required or desired to perform the calibration.
This Safety Assessment and Significant Hazards Consideration (SASHC) proposes a revision to the frequency of several such Surveillance Requirements and addresses the associated instrument drift. These
LAR 96-0014 Page 3 Surveillance Requirements and the specific proposed interval revision l
(i.e., REFUELING INTERVAL or "R" notation) are described in the enclosures to this SASHC, and include:
- 1. 4.3.2.1.1 - Table 4.3-2, Safety Features Actuation System Instrumentation Surveillance Requirements, Functional Unit 1.d, I
RCS Pressure - Low, Channel Calibration.
l 2.
4.3.2.1.1 - Table 4.3-2, Safety Features Actuation System j
Instrumentation Surveillance Requirements, Functional Unit 1.e, RCS Pressure - Low-Low, Channel Calibration.
3.
4.3.2.1.1 - Table 4.3-2, Safety Features Actuation System l
Instrumentation Surveillance Requirements, Punctienal Unit 5.a, Decay Heat Isolation Valve Interlock, Channel Calavration.
i 4
4.3.2.1.1 - Table 4.3-2, Safety Features Actuation System Instrumentation Surveillance Requirements, Functional Unit 5.b, 4
Pressure Heater Interlock, Channel Calibration.
l
- 5. 4.5.2.d.1.a - Decay Heat Isolation Valve and Pressurizer Heater 4
Interlocks, Channel Functional Test (referenced by TS Table 1
4.3-2, Functional Units 5.a and 5.b).
I I
6.
4.5.2.d.1.b - Decay Heat Isolation Valve Interlock, Channel 3
Functional Test (referenced by TS Table 4.3-2, Functional Unit 5.a).
3 Proposed Setooint Revisions Supportina the Instrument Qrift Study I
The following revisions to the Technical Specifications are proposed as a result of the instrument drif t study analysis required by GL 91-04 and j
the subsequent revisions to safety analyses.
It should also be noted j
that application of the Allowable Value to the Channel Functional Test
]
only and deletion of the Trip Setpoint from the Technical Specifications reflects the DBNPS's desire to change the assumptions of specific i
setpoint analyses for conformance with NUREG-1430, Revision 1,
" Standard i
Technical Specifications, Babcock and Wilcox Plants," and is not solely
]
driven by the results of the instrument drift study analysis.
i 1.
Technical Specification 3/4.3.2.1 Table 3.3-4, Safety Features 4
}
Actuation System Instrumentation Trip Setpoints, Allowable Value i
for Instrument String Functional Unit d, RCS Pressure - Low, is proposed for revision from 2 1615.75 psig to 2 1576.2 psig. This Allowable value is designated as applicable to the Channel Functional Test by application of a new footnote "##" to TS Table 1
3.3-4, to read as follows:
- Allowable Value for CHANNEL FUNCTIONAL TEST l
The Trip Setpoint associated with this Allowable Value is proposed i
for deletion and is proposed for designation as "N.A.",
which means "not applicable."
4 4
LAR 96-0014 Page 4
- 2. Technical Specification 3/4.3.2.1 Table 3.3-4, Safety Features Actuation System Instrumentation Trip Setpoints, Allowable Value for Instrument String Functional Unit e, RCS Pressure -
Low-Low, is proposed for revision from 2 415.75 psig to 2 441.42 psig. This Allcwable value is designateu as applicable to the Channel Functional Test by application of the new footnote "##" to TS Table 3.3-4, as described above.
The Trip Setpoint associated with this Allowable Value is proposed for deletion and is proposed for designation as "N.A."
In addition, related to this change, footnote "**" to TS Table 3.3-3, Safety Features Actuation System Instrumentation, which applies to Instrument String Functional Unit 1.e, RCS Pressure Low-Low, is proposed for revision. This footnote presently allows the trip function to be bypassed in Mode 3 (Hot Standby) with RCS pressure below 600 psig, and specifies that the bypass shall be automatically removed when RCS pressure exceeds 600 psig. The proposed change would revise the 600 psig value to 660 psig for both the bypass permissive and the reset.
- 3. Technical Specification 3/4.3.2.1 Table 3.3-4, Safety Features Actuation System Instrumentation Trip Setpoints, Allowable value for Interlock Channel Functional Unit a, Decay Heat Isolation valve and Pressurizer Heater, is proposed for revision from < 443 psig to < 328 psig. This Allowable Value is designated as applicable to the Channel Functional Test by application of the new footnote "##" to TS Table 3.3-4, as described above.
Footnote "*" to TS Table 3.3-4, which presently reads
" Referenced to the centerline of DH11 and DH12," is proposed to be revised to read " Referenced to the RCS Pressure instrument tap."
The Trip Setpoint associated with this Allowable value is proposed for deletion and is proposed for designation as "N.A."
Other Proposed Revisions 1.
The TS 3.3.2.1 Limiting Condition for Operation (LCO) and Action Statement 3.3.2.1.a are proposed for revision to reflect the proposed changes to the SFAS Trip Setpoints and Allowable Values.
Limiting Condition for Operation (LCO) 3.3.2.1 is proposed to be changed to read as follows:
(.
b LAR 96-0014 Page 5 1
The Safety Features Actuation System (SEAS) functional units shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip setpoint column of Table 3.3-4, with the exception of Instrument Strings Functional Units d and e and Interlock Channels Functional Unit a which shall be set consistent with the Allowable Value column of Table 3.3-4, and with RESPONSE TIMES as shown in Table 3.3-5.
Action Statement 3.3.2.1.a is proposed to be changed to read as follows:
With a SFAS functional unit trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3-4, declare the functional unit inoperable and apply the applicable ACTION requirement of Table 3.3-3, until the functional unit is restored to OPERABLE status with the trip setpoint adjusted consistent with Table 3.3-4.
j
- 2. Technical Specification 3/4.3.2.1 Table 3.3-3, Safety Features
)
Actuation System Instrumentation, Actions 23 and 14, which 4
apply to Interlock Channel Functional Units 5.a, Decay Heat Isolation Valve, and 5.b, Pressurizer Heaters, respectively, are proposed for revision to reflect the above-mentioned i
proposed revision to TS Table 3.3-4, Allowable value for Interlock Channels Functional Unit a.
TS Table 3.3-3 Action 13 is proposed to read as follows:
- a. With less than the Minimum Units OPERABLE and indicated j
reactor coolant pressure 2 328 psig, both Decay Heat I
Isolation Valves (DH11 and DH12) shall be verified
- closed, l
- b. With less than the Minimum Units OPERABLE and indicated reactor coolant pressure < 328 psig operation may continue; however, the functional unit shall be OPERABLE prior to increasing indicated reactor coolant pressure above 328 psig.
4 TS Table 3.3-3 Action 14 is proposed to read as follows:
With less than the Minimum Units OPERABLE and indicated reactor coolant pressure < 328 psig, operation may continue; however, the functional unit shall be OPERABLE
]'
prior to increasing indicated reactor coolant pressure above 328 psig, or the inoperable functional unit shall be placed in the tripped state.
d
.I T
LAR 96-0014 page 6
- 3. Surveillance Requirement 4.5.2.d.1 is also proposed for revision to reflect the above-mentioned proposed revision to i
TS Table 3.3-4, Safety Features Actuation System j
Instrumentation Trip Setpoints, Allowable Value for Interlock Channels Functional Unit a, i
Surveillance Requirement 4.5.2.d.1 is proposed to read as follows:
i
- 1. Verifying that the erlocks:
a) Close DH-11 and km 12 and deenergize the pressurizer heaters, if either DH-11 or DH-12 is open and a simulated reactor coolint system pressure which is greater than the Allowable Value (<328 psig) is applied. The interlock to clobe DH-11 and/or DH-12 is not required if the valve le closed and 400 V AC power is disconnected from its motor operators, b) Prevent the opening of DH-11 and DH-12 when a simulated or actual reactor coolant system pressure which is greater than the Allowable value (<328 psig) is applied.
4.
An administrative change to TS Bases 3/4.3.1 and 3/4.3.2, Reactor Protection System and Safety System Instrumentation, is proposed to reflect the above-mentioned changes to TS 3/4.3.2.1.
A new insert is proposed to be added following the current third paragraph, to read as follows:
For the RPS, SEAS Table 3.3-4 Functional Unit Instrument Strings d and e and Interlock Channel a, and SFRCS Table 3.3-12 Punctional Unit 2:
only the Alloweble value is specified for each Function.
Nominal trip setpoints are specified in the setpoint analysis. The nominal trip setpoints are selected to ensure the setpoints measured by CHANNEL FU;1CTIONAL TESTS do not exceed the Allowable value if the bistable is performing as required. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable value, is acceptable provided that operation and testing are consistent with the assumptions of the specific setpoint calculations.
Each Allowable value specified is more conservative than the analytical limit assumed in the safety analysis to account for instrument uncertainties appropriate to the trip parameter.
These uncertainties are defined in the specific setpoint analysis.
i
LAR 96-0014 Page 7 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Setpoints must be found within the specified Allowable Values. Any setpoint adjustment shall be consistent with the assumptions of the current specific setpoint analysis.
A CHANNEL CALIBRATION is a complete check of the instrument channel, including the sensor. The test verifies that the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift to ensure that the instrument channel remains operational between successive tests. CHANNEL CALIBRATION shall find that measurement errors and bistable setpoint errors are within the assumptions of the setpoint analysis. CRANNEL CALIBRATIONS must be performed consistent with the assumptions of the setpoint analysis.
The frequency is justified by the assumption of an 18 or 24 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis.
It is noted that with the exception of the introductory phrase, this insert is identical to the Bases change proposed by LAR 95-0024.
Also, as noted in the cover letter to this SASHC, Toledo Edison is requesting that LAR 95-0024 and this amendment application, LAR 96-0014, be issued together by the NRC.
The proposed Bases change for LAR 95-0024 referred to SFRCS Table 4.3-11 Functional Unit 1.b (Steam Generator Level
- Low).
It is proposed to revise this reference to SFRCS Table 3.3-12, Functional Unit 2 (Steam Generator Level - Low).
This reference is more appropriate since Allowable Values are actually provided in Table 3.3-12, where applicable.
Combining the proposed Bases changes, and taking into account the above mentioned administrative changes to the SFRCS Table reference, the introductory phrase would read:
For the RPS, SFAS Table 3.3-4 Functional Unit Instrument Strings d and e and Interlock Channel a, and SFRCS Table 3.3-12 Functional Unit 2:
5.
An administrative change to TS Bases 3/4.5.2 and 3/4.5.3, ECCS Subsystems, is proposed to add a brief discussion of the Decay Heat Isolation Valve and Pressurizer Heater Interlocks. This change is associated with the above-mentioned changes to SR 4.5.2.d.1.
The new insert would read as follows:
1 LAR 96-0014 Page 8 l
The Decay Heat Isolation Valve and Pressurizer Hu?mr Interlock setpoint is based on preventing over-l pressurization of the Decay Heat Removal System nom..sa suction line piping. The value stated is the RCS pressure
)
at the sensing instrument's tap.
It has been adjusted to reflect the elevation diffelence between the sensor's j
3 location and the pipe of concern.
Marked-uo Operatino License Paces 4
a Each of the proposed revisions are shown on the attached marked-up l
Operating License pages. The marked-up page showing the proposed change
{
to TS Bases 3/4.3.1 and 3/4.3.2 reflects the combined insert for LAR
{
95-0024 and LAR 96-0014, as discussed above.
SYSTEMS, COMPONENTS, AND ACTIVITIES AFFECTED:
Systems Affected The system affected by the proposed revisions is the Safety Features Actuation System, including the decay heat isolation valve and pressurizer heater interlocks. Surveillance requirements associated with I
these interlocks are also specified in the Emergency Core Cooling System Technical Specifications, which are also proposed for revision.
Components Affected The components affected by the proposed revisions can be discussed in a generic fashion since the instrument uncertainty is applied to the instrument strings in a generic manner with minimal regard to the specific instrument string involved.
I The instrument strings involved with this LAR and LAR 95-0024 can be categorized into three general types:
- 1. A process sensing switch with a digital (contact state) output.
2.
Process transmitter and indication.
- 3. A process transmitter, signal conditioning circuitry, bistables, a controller, and/or indication.
A. list of all instrument strings and corresponding Technical Specifications is provided in the " Instrument Drift Data Analysis Methodology and Assumptions" for DBNPS (Attachment 1) on pages 14 and 15 of 19.
A summary of the Instrument Drift Study Analysis, as performed on each of the affected strings specifically for this LAR, is contained in Section 3 of the Enclosures to this LAR.
i LAR 96-0014 i
Page 9 Activities Affected The basic activity affected by these proposed revisions is the performance of certain surveillance tests on a 24 month frequency instead of an 18 month frequency. The scope of the effect on the surveillance tests, as made necessary by the proposed revisions, involves either an extension in the interval of performance or an extension of the interval of performance with the application of a modified setpoint, due to a revised Allowable Value. Note that each of the individual surveillance tests affected have been individually identified as a part of the Instrument Drift Study.
1 FUNCTIONS OF THE AFFECTED SYSTEMS, COMPONENTS, AND ACTIVITIES:
Punctions of the Systems Affected The functions of the systems affected by the proposed revisions are discussed in item 2.B of the Enclotures to this LAR.
l Functions of the components Affected
)
The Functions of the Components Affected by the revisions proposed in this LAR and LAR 95-0024 can be discussed in generic fashion since the application of instrument uncertainty is applied to instrument funct' ions in a generic manner with minimal regard to the specific instrument string involved. The functions of the instrument strings involved can again be categorized according to the same three types listed in the " Components Affected" section:
- 1. For the process sensing switch string, the switch assembly i
senses the process (e.g., pressure or level). When the process value goes above or below a prescribed setpoint, the switch provides a digital (i.e.,
contact state) output to
}
indication, logic circuitry, or equipment actuation, i
2.
For the transmitter with an indication string, the transmitter senses the process (e.g.,
pressure, temperature, radiation, or differential pressure) and generates a voltage or current signal proportional to the process. This signal is then conditioned (e.g.,
converted, scaled, or isolated) for further use.
The conditioned signal can then be provided as process indication.
3.
For the string consisting of trecsmitter, signal conditioning circuitry, bistable, controller, and/or it.dication, the transmitter senses the process (e.g., precsure, temperature, radiation, or differential pressure) and generates a voltage or current signal proportional to the process. This signal is then conditioned (e.g.,
converted, scaled, or isolated) for
d j
LAR 96-0014 Page 10 1
further use.
The conditioned signal can then be applied to a bistable, which when above or below a prescribed setpoint, provides a digital (i.e., contact state) output to various logic circuitry or actuation equipment, or can be applied to a controller.
The functions of each of the instrument strings applicable to this LAR are discussed in more detail in the Enclosures.
Functions of the Activities Affected The TS surveillance testing verifien compliance with specific TS criteria, such as Allowable values. The frequency of the testing and the specific acceptance criteria based on the Technical Specifications are 2
- affected, i.
INSTRUMENT DRIFT ANALYSIS:
Note: Throughout this document the term " drift" refers to the difference between as-found calibration data and the corresponding as-left data from the previous calibration. The only notable exceptions to this are:
item I
15b of the " Instrument Drift Data Analysis Methodology and Assumptions" for DBNPS (Attachment 1); anywhere that drift refers to an uncertainty term in a setpoint or instrument string error calculation; and when used in reference to vendor specifications.
For these exceptions, drift is used as defined in ANSI /ISA - S51.1, " Process Instrumentation i
Terminology."
Generic Letter 91-04, " Changes in Technical Specification Surveillance Intervals to Support a 24-Month Fuel Cycle," dated April 2,
- 1991, identifies several f.ssues to be addressed in justifying increased surveillance intervals to accommodate a 24-month fuel cycle:
1.
" Confirm that instrument drift as determined by as-found and as-left calibration data from surveillance and maintenance j
records has not, except on rare occasions, exceeded acceptable limits for a calibration interval."
Calibration data was reviewed for those surveillan, proposed for revision to a 24 month interval. As-found and x*
left data has not exceeded acceptable limits except on rare oct. ions.
The Enclosures to this LAR discuss notable occasions whert seceptable limits were exceeded.
2.
" Confirm that the values of drift for each instrument type (make, model, and range) and application have been determined with a i
high probability and a high degree of confidence.
Provide a summary of the methodology and assumptions used to determine the rate of instrument drift with time based upon historical plant calibration data."
LAR 96-0014 Page 11 Drift values were determined for each instrument string. Basic statistics calculated for each drift value included a 95/95%
tolerance factor and a 95/95% tolerance interval.
Potential data outliers were identified in accordance with ANSI / ASTM E178-1994,
" Standard Practice for Dealing With Outlying Observations."
Normality of the drift data was verified by the W or D' test in accordance with ANSI N15.15-1974, " Assessment of the Assumption of Normality (Employing Individual Observed Values). " When the test results indicated that the assumption of normality should be rejected a histogram was developed to verify that the drift data was bounded by a normal distribution.
Time dependency was determined for each instrument string by application of various methods described in Step 13 of the
" Instrument Drift Data Analysis Methodology and Assumptions" for DBNPS (Attachment 1).
If the results of the evaluation were incenclusive then some time dependency was assumed to exist.
3.
"Confinn that the magnitude of instrument drif t has been determined with a high probability and a high degree of confidence for a bounding calibration interval of 30 months for each instrument type (make, model number, and range) and application that performs a safety function.
Provide a list of channels by TS section that identifies these instrument applications."
The projected 95/95% tolerance interval for a 30 month interval was determined as described in Step 14 of the " Instrument Drift Data Analysis Methodology and Assumptions" for DBNPS (Attachment 1).
If the drift was determined to be time-independent then the previously calculated 95/95% tolerance interval was applied to the 30 month interval.
If the drift was determined to exhibit time-dependency then the 30 month interval drift was extrapolated for each individual drift data point with a calibration interval less than 30 months.
The list of affected instrument strings is provided in the
" Instrument Drift Data Analysis Methodology and Assumptions" for DBNPS (Attachment 1) on pages 14 and 15 of 19.
Note that the list also includes instrument strings addressed by LAR 95-0024.
4.
" Confirm that a comparison of the projected instrument drift errors has been made with the values of drift used in the setpoint analysis.
If this results in revised setpoints to accommodate larger drift errors, provide proposed TS changes to update trip setpoints.
If the drift errors result in a revised safety analysis to support existing setpoints, provide a summary of the updated analysis conclusions to confirm that the safety limits and safety analysis assumptions are not exceeded."
Page 12 For instrument strings that perform an automatic protective function, an evaluation of the calculations which established the existing setpoint was performed.
Step 15 of the " Instrument Drif t Data Analysis Methodology and Assumpt ions" for DBNPS (Attachment 1) describes the methods used to perform these evaluations.
Setpoint changes are proposed for the Allowable Values for TS 3/4.3.2.1, Safety Features Actuation System (SFAS)
Instrumentation, TS Table 3.3-4, Instrument String Functional Unit d (RCS Pressure - Low), Instrument String Functional Unit e (RCS Pressure - Low-Low), and Interlock Channel Functional Unit a (Decay Heat Isolation Valve and Pressurizer Heater). Associated with the proposed change for the RCS Pressure Low-Low Allowable Value, changes were also necessary for the RCS Pressure Low-Low bypass permissive and reset values specified in TS Table 3.3-3.
5.
" Confirm that the projected instrument errors caused by drift are acceptable for control of plant parameters to effect a safe I
shutdown with the associated instrumentation."
As discussed in the above response to question 4, Step 15 of the
" Instrument Drift Data Analysis Methodology and Assumptions" for DBNPS (Attachment 1) describes the methods used to perform these evaluations.
However, as discussed in the Enclosures, the instrument strings applicable to this LAR do not control plant parameters in an analog fashion, but rather provide protective action signals to either initiate operation of actuated equipment or deenergize equipment. Hence, this question is not applicable.
6.
" Confirm that all conditions and assumptions of the setpoint and safety analyses have been checked and are appropriately reflected in the acceptance criteria of plant surveillance procedures for channel checks, channel functional tests, and channel calibrations."
The applicable surveillance and periodic test procedures were reviewed to verify that they appropriately reflect all applicable conditions and assumptions of the setpoint and safety analysis.
Where field setpoints, Allowable Values, and calibration methods are changing, the conditions and assumptions of the appropriate setpoint analyses will be verified to be incorporated in revised and/or new surveillance test procedures during implementation of this LAR.
7.
" Provide a summary description of the program for monitoring and assessing the effects of increased calibration surveillance intervals on instrument drift and its effect on safety."
Appropriate DBNPS surveillance procedures will be revised to require that for any instrument string which had its calibration interval extended from 18 to 24 months, a Potential Condition Adverse Quality Report (PCAQR) be initiated if as-found data for
LAR 96-0014 Page 13 any component or string exceeds its expected value limit, as determined consistent with the drift study methodology.
Following each refueling outage an evaluation will be performed of each PCAQR initiated during that outage to verify that instrument drifts occurring over the increased calibration interval are consistent with the 95/95 porcent tolerance intervals established by the drift study and any applicable setpoint and safety analyses. The requirement for performing the evaluation will be incorporated into a revision to the appropriate DBNPS procedures.
The enclosures to this SASHC discuss the results of the instrument drift I
study.
EFFECTS ON SAFETY:
Proposed Revisions to Surveillance Requirement Intervals The enclosures to thir SASHC describe the effect on safety due to increasing certain surveillance test intervals from 18 to 24 months and the continued application of TS 4.0.2 (which allows surveillance intervals to be increased up to 25% on a non-routine basis). As required by Generic Letter 91-04, an instrument drift study analysis was conducted for each affected instrument string. Where projected instrument drift errors exceeded the allowance for instrument drift that was used to establish setpoints, either a new setpoint has been established or the calculation's exce.ss margin was reduced, or both.
In addition, the licensing basis was reviewed for each proposed revision to ensure it was not invalidated, and applicable surveillance data and maintenance history reviews were performed.
Based on the results of the instrument drift study analysis and j
applicable surveillance data and maintenance history reviews, it is concluded that there is no adverse effect on nuclear safety due to increasing the surveillance test intervals from 18 to 24 months and the continued application of TS 4.0.2.
In addition, the licensing basis remains valid.
Manufacturer or vendor maintenance information for the affected components is considered in the DBNPS Preventive Maintenance (PM)
Program. The PM Program is being evalue+ cd as a separate activity in support of the conversion from an 18 month to a 24 month fuel cycle.
Changes will be made, as necessary, in the PM Program to facilitate a 24 month fuel cycle.
Proposed Setpoint Revisions Supportino the Instrument Drift Study
- 1. The 30 Month Drift Study found that the SFAS RCS Pressure -
Low instrument string uncertainty would be greater than the uncertainty used in calculating the current Technical specification Allowable value. The additional uncertainty
LAR 96-0014 Page 14 1
would reduce +.he meagin, during a normal plant shutdown, f
between receipt of the SFAS Pressure - Low trip block permissive signal and the trip setpoint. To maintain a l
suitable operating margin, the potential for lowering the SFAS Pressure - Low trip setpoint was investigated. By lowering the Technical Specification Allowable Value, the field setpoints could be set to preserve the operating margin.
A review has determined that of the many accident analyses which use the SFAS Pressure - Low trip, three analyses were based on an analytical value of 1585 psia, and the remaining analyses were based on an analytical value of 1515 psia or less. It was concluded that a new analytical value of 1515 psia should form the basis of the revised Allowable value. The three accident analyses impacted by a reduction of the analytical value to 1515 psia are the 0.04 square foot Small Break Loss-of-Coolant Accident (SBLOCA) analysis, the letdown line break analysis, and the Steam Generator Tube Rupture (SGTR) analysis. Each is discussed below.
- a. SBLOCA Analysis The 0.04 square foot SBLOCA analysis was originally performed to address NRC questions concerning NUREG-0737,
" Clarification of TMI Action Plan Requirements."
The.
effect of lowering the trip to 1515 psia is a slight delay in the actuation of high pressure injection (HPI) and a slightly lower core collapsed liquid level. Based on a previous evaluation, an additional 7.7 seconds are required for the RCS to reach 1515 psia. After accounting for the mass the HPI delivered during the 7.7 seconds, the reduction in core liquid level was determined to be 1.33 inches. This amounts to a reduction in the minimum core collapsed liquid level from 10.07 feet to 9.96 feet.
The previous evaluation determined that the collapsed liquid level must fall below 8.5 feet before core uncovery is predicted and peak clad temperature (PCT) excursions occur. Therefore the analytical trip value of 1515 psia will not yield PCT excursions and still meets the accident analysis acceptance criteria.
- b. Letdown Line Break Analysis i
l The letdown line break event described in USAR Section 15.4.5, " Break in Instrument Lines or Lines From Primary System That Penetrate Containment," was revised in 1988 to incorporate a reduction in the Reactor Protection System (RPS) RCS Pressure - Low trip from 1985 psia to 1900 psig.
This analysis has since been revised to isolate the line break 15 seconds after the SFAS Pressure - Low trip.
Therefore, a reduction in the SFAS trip setpoint results in delaying isolation, causing an additional mass and energy release from the break.
It is noted that the isolation delay is only applicable to the USAR Chapter 15 accident
1 analysis and associated radiological consequences j
(discussed below) and it does not affect the USAR Chapter 3 High Energy Line Break (HELB) analysis since the letdown cooler outlet temperature switches are credited for l
mitigation of this event. The additional depressurization i
time required to reach 1515 psia versus 1585 psia was i
i determined to be 45.51 seconds. The additional mass and
(
energy releases associated with the time delay were j
determined to be 7,104 lbm and 3,054,720 BTU, respectively.
Using the information described above, a radiation dose calculation was performed to determine the increase in doses at the exclusion area boundary (EAB) and the low population zone (LPZ) due to the proposed change. The following table compares the results of the old dose calculation performed in 1988 and the new radiation doses l
associated with an SFAS RCS Pressure - Low trip setpoint of i
1515 psia.
Exclusion Area Low Population I
Boundarv Zone (Old)
(New)
(Old)
(New) l Thyroid (Rem) 3.52 4.83 0.18 0.25 l
Whole Body (Rem) 0.03 0.04 0.002 0.002 l
As the table shows, the doses are higher than those l
l previously reported, due to the longer blowdown time, i
l However, these results still satisfy the acceptance criteria of NRC Standard Review Plan (SRP) Section 15.6.2,
" Radiological Consequences of the Failure of Small Lines i
Carrying Primary Coolant Outside Containment," which l
requires that doses do not exceed a small fraction (10%) of l
10 CFR 100 guideline values, that is, 2.5 Rem and 30 Rem for the whole-body and thyroid doses, respectively.
l
- c. SGTR Analysis l
The Steam Generator Tube Rupture Analysis reported in USAR l
Section 15.4.2, " Steam Generator Tube Rupture," used an i
analytical limit abova the proposed value of 1515 psia.
l The analysis assumed that all the reactor coolant that i
transferred to the secondary side of the Steam Generator had a constant activity concentration. No credit was taken for the dilution of the RCS by the water injected from the Borated Water Storage Tank as the accident progressed.
Consequently, the amount of radioactive material available j
to contribute to the offsite dose is unaffected by the point at which HPI is initiated. Delaying the initiation of High Pressure Injection does not increase the time to depressurize the RCS to below the setpoint of the lowest Main Steam Safety Valve. Consequently, the time to isolate the faulted Steam Generator and the resulting radioactive
}
release is unaffected by a revised SFAS RCS Pressure Low analytical value of 1515 psia.
1 l
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.m___m____..-
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Page 16 1.
Using 1515 psia as the analytical setpoint, the TS Allowable Value was determined to be 1576.2 psig using the methodology l
of ISA S67.04, Part I-1994, "Setpoints for Nuclear i
j Safety-Related Instrumentation."
i l
Based on the above, this proposted setpoint revision will have no adverse effect on plant safety.
h Applicability of the new Allowable Value to the Channel Functional Test only and the proposed deletion of the Trip Setpoint is consistent with NUREG-1430, Revision 1,
" Standard Technical Specifications, Babcock and Wilcox Plants," dated l
April, 1995.
The calculated Trip Setpoint value will be
}
maintained in DBNPS design basis documentation.
Therefore, j
there will be no adverse effect on nuclear safety.
l l
)
- 2. The SEAS RCS Pressure Low-Low Trip is based on an analytical value of 370 psig. Using this value, the TS Allowable Value
]-
is 441.42 psig, calculated using the methodology of ISA 4
S67.04, Part I-1994, "Setpoints for Nuclear Safety-Related l
j Instrumentation." Since the proposed Allowable Value ensures the Low-Low trip analytical value is preserved, the change ham l
l no adverse effect on nuclear safety.
.I' Associated with this Allowable value change, the RCS Pressure l
Low-Low bypass permissive and reset values, as specified in l
footnote "**" to TS Table 3.3-3, Safety Features Actuation j
d System Instrumentation, are proposed to be changed from j
600 psig to 660 psig.
This bypass allows depressurization of l
3 the RCS, in Mode 3, without initiating the RCS Pressure l
Low-Low trip?' Since the Allowable value is being char.ged, the j
bypass values need to be changed in order to maintain an j
approximately equivalent opportunity, during controlled RCS depressurization, for the operator to initiate the bypass l
prior to reaching the trip. This change will have no adverse effect on plant safety.
Applicability of the new Allowable Value to the Channel Functional Test only and the proposed deletion of the Trip Setpoint is consisteat with NUREG-1430, Revision 1, " Standard
[
Technical Specifications, Babcock and Wilcox Plants," dated April, 1995.
The calculated Trip Setpoint value will be maintained in DBNPS design basis documentation. Therefore, there will be no adverse effect on nuclear safety.
- 3. An evaluation of the SFAS Decay Heat Isolation Valve and Pressurizer Heater Incerlock trip setpoint revealed that the current Technical Specification value is not appropriate for the design basis of the interlock. The current value does not prevent over-pressurization of the Decay Heat Removal System
[
normal suction line p ping. This condition was documented via the plant corrective action program. The proposed Allowable I
LAR 96-0014 Page 17 Value, 328 psig, includes the correct design basis.
It also accounts for the 30 Month Projected Drift value. The curront Allowable value is referenced to the centerline elevation of valves DH-11 and DH-12, however the proposed Allowable Value is referenced to the RCS Pressure instrument tap, which allows the operators to more easily monitor proximity to the trip.
This change provides an operational improvement. Overall, the proposed changes will have no adverse effect on nuclear safety.
Applicability of the new Allowable value to the Channel
)
Functional Test only and the proposed deletion of the Trip Setpoint is consistent with NUREG-1430, Revision 1,
" Standard Technical Specifications, Babcock and Wilcox Plants," dated i
April, 1995.
The calculated Trip Setpoint value will be maintained in DBNPS design basis documentation. Therefore, there will be no adverse effect on nuclear safety.
Other ProDosed Revisions l
1.
The proposed revisions to the TS 3.3.2.1 Limiting Condition for Operation (ICO) and Action Statement 3.3.2. lea are associated with the proposed revisions to the Allowable values for SFAS RCS Pressure - Low, SFAS RCS Pressure - Low-Low, and 4
Decay Heat Isolation Valve and Pressurizer Heater Interlocks, j
and are consistent with NUREG-1430, Revision 1,
" Standard Technical Specifications, Babcock and Wilcox Plants," dated
)
April, 1995.
Therefore, as discussed above, these changes will have no adverse effect on nuclear safety.
3
- 2. The proposed revisions to TS 3/4.3.2.1 Table 3.3-3 Actions 13 and 14 reflect the proposed changes to the TS Table 3.3-4 Allowable value for Interlock Channel Functional Unit a, and will have no adverse effect on nuclear safety.
- 3. The proposed revisions to Surveillance Requirement 4.5.2.d.1 reflect the change to the TS Table 3.3-4 Allowable value for Interlock Channel Functional Unit a.
These proposed revisions are consistent with the proposed revisions to the associated Allowable Value and the proposed deletion of the Trip Setpoint, and therefore will have no adverse effect on nuclear safety.
- 4. The proposed revision to TS Bases 3/4.3.1 and 3/4.3.2 is an administrative change related to the other proposed changes, and will have no adverse effect on nuclear safety.
S. The proposed revision to TS Bases 3/4.5.2 and 3/4.5.3 is an administrative change related to the other proposed changes, and will have no adverse effect on nuclear safety.
l i
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Page 18 i
SIGNIFICANT HAZARDS CONSIDERATION:
The Nuclear Regulatory Commission has provided standards in 10CFR50.92(c) for determining whether a significant hazard exists due to a proposed amendment to an Operating License for a facility. A proposed amendment involves no significant hazards consideration if operation of the facility in accordance with the proposed changes would: (1) Not involve a significant increase in the probability or consequences of an accident previously evaluated; (2) Not create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Not involve a significant reduction in a margin of safety. The Davis-Besse Nuclear Power Station has reviewed the proposed changes and determined that a significant hazards consideration does not exist because operation of the Davis-Besse Nuclear Power Station, Unit No.
1, in accordance with these changes would:
l la.
Not involve a significant increase in the probability of an accident l
4 previously evaluated because the initiation of such accidents are j
not affected by the proposed revisions to increase the surveillance j
test intervals from 18 to 24 months for the subject Technical j
Specifications (TS): TS 3/4.3.2.1, Safety Features Actuation System Instrumentation, and TS 3/4.5.2, Emergency Core Cooling Systems -
ECCS Subsystems - Tavg 2 280*F.
Initiating conditions and assumptions remain as previously analyzed for accidents in the DBNPS Updated Safety Analysis Report.
Results of the instrument drift study analysis and review of historical 18 month surveillance data and applicable maintenance 3
records support an increase in the surveillance test intervals from j
j 18 to 24 months (and up to 30 months on a non-routine basis) because: the projected instrument errors caused by drift are bounded by the existing setpoint analysis or a new analysis has been performed incorporating a more conservative setpoint; and no potential for a significant increase in a failure rate of a system or component was identified during surveillance data and applicable maintenance records reviews.
1 These proposed revisions are consistent with the NRC guidance on evaluating and proposing such revisions as provided in Generic Letter 91-04, " Changes in Technical Specification Surveillance Intervals to Accommodate a 24-Month Fuel Cycle," dated April 2, 1991.
J The proposed revisions to Allowable values for Safety Features Actuation System (SFAS) Reactor Coolant System (RCS) Pressure - Low, RCS Pressure - Low-Low, RCS Pressure - Low-Low bypass permissive, and Decay Heat Isolation Valve and Pressurizer Heater Interlocks have no bearing on the probability of the initiation of an accident previously evaluated.
The application of the Allowable value to only the Channel Functional Test and not the Channel Calibration, the proposed deletion of the Trip Setpoints, the proposed revision of the
. -. -. -. - - -. _ _. -. - -. -.-. _.- -.-.--~ - ~.. -
LAR 96-0014 Page 19 TS 3.3.2.1 Limiting Condition for Operation (LCO) and Action Statement 3.3.2.1.a, and the proposed revisions to Actions 13 and 14 l
of TS Table 3.3-3, are associated with the proposed revision of the Allowable Values for Safety Features Actuation System RCS Pressure -
i Low, RCS Pressure - Low-Low, and Decay Heat Isolation Valve and Pressurizer Heater Interlocks, and are consistent with NUREG-1430, Revision 1,
" Standard Technical Specifications, Babcock and Wilcox Plants," dated April, 1995.
The proposed revisions have no bearing on the probability of the initiation of an accident previously l
evaluated.
l The proposed changes to TS Bases 3/4.3.1 and 3/4.3.2, " Reactor l
Protection System and Safety System Instrumentation," and TS Bases 3/4.5.2 and 3/4.5.2, "ECCS Subsystems," are administrative changes i
associated with the other proposed changes, and do not affect l
previously analyzed accidents.
l l
lb.
Not involve a significant increase in the consequences of an accident previously evaluated because the slight increase in doses due to a letdown line break event as a result of the proposed change t
to the SFAS RCS Pressure - Low Allowable Value still satisfy the NRC Standard Review Plan Section 15.6.2 acceptance criteria that doses do not exceed a small fraction (10%) of the 10 CFR 100 guideline j
values.
The remaining proposed changes to Allowable values, and the other changes proposed by this License Amendment Request do not increase the radiological consequences of previously analyzed accidents because the source term, containment isolation, or radiological releases are not being changed by the proposed revisions.
2.
Not create the possibility of a new or different kind of accident j
from any accident previously evaluated, for the reasons discussed below.
j i
No changes are being proposed to the type of testing currently being performed, only to the length of the surveillance test interval.
Results of the instrument drift study analysis and review of historical 18 month surveillance data and maintenance records support an increase in the surveillance test intervals from 18 to 24 months (and up to 30 months on a non-routine basis) because: the projected instrument errors caused by drift are bounded by the existing setpoint analysis or a new analysis has been performed i
incorporating a more conservative setpoint; and no potential for a significant increase in a failure rate of a system or component was i
identified during surveillance data and applicable maintenance records reviews.
i i
The proposed revisions to Allowable Values for Safety Features Actuation System RCS Pressure - Low, RCS Pressure - Low-Low, RCS Pressure Low-Low bypass permissive, and Decay Heat Isolation Valve
]
l and Pressurizer Heater Interlocks, do not alter the type of any j
testing currently being performed.
j i
l
,-~
__m._
s i
Page 20 The application of the Allowable value to only the Channel
{
Functional Test and not the Channel Calibration, the proposed
(
deletion of the Trip Setpoints, revision of the TS 3.3.2.1 Limiting i
Condition for Operation (LCO) and Action Statement 3.3.2.1.a, and l
the proposed revisions to Actions 13 and 14 of TS Table 3.3-3, are
}
associated with the proposed revision to the Allowable Values for i
Safety Features Actuation System RCS Pressure - Low, RCS Pressure -
I Low-Low, RCS Pressure Low-Low bypass permissive, and Decay Heat l
Isolation valve and Pressurizer Heater Interlocks, and are consistent with NUREG-1430, Revision 1,
" Standard Technical Specifications, Babcock and Wilcox Plants," dated April, 1995.
The proposed revisions do not alter the type of testing currently being performed.
The proposed changes to TS Bases 3/4.3.1 and 3/4.3.2, " Reactor i
Protection System and Safety System Instrumentation," and TS Bases 3/4.5.2 and 3/4.5.3, "ECCS Subsystems," are administrative changes associated with the other proposed changes, and do not alter any I
testing currently being performed.
3.
Not involve a significant reduction in a margin of safety. The results of the instrument drift study analysis and review of historical le month surveillance data and applicable maintenance records' support an increase in the surveillance test intervals from
[
18 to 24 months (and up to 30 months on a non-routine basis)
I because: the projected instrument errors caused by drift are bounded by the existing setpoint analysis or a new analysis has been I
performed incorporating a more conservative setpoint; and no potential for a significant increase in a failure rate of a system or component was identified during surveillance data and applicable maintenance records reviews. Existing system and component redundancy is not affected by these proposed changes.
There are no new or significant changes to the initial conditions contributing to accident severity or consequences, consequently I
there are no significant reductions in a margin of safety.
CONCLUSIONS:
On the basis of the above, the Davis-Besse Nuclear Power Station has l
determined that the License Amendment Request does not involve a l
significant hazards consideration.
e I
i l
_l
e l
LAR 96-0014 Page 21 ATTAfNK1ruTg :
i Attached are the proposed marked-up changes to the Operating License.
Enclosed are summaries of the licensing basis and responses to the Generic Letter 91-04, Enclosure 2 issues for each of the Technical Specification Surveillance Requirements proposed for revision as listed
)
below: : SR 4.3.2.1.1 Table 4.3-2 Functional Units 1.d and 1.e : SR 4.3.2.1.1 Table 4.3-2 Functional Units 5.a and 5.b, and SR 4.5.2.d.1 Also attached is a copy of the " Instrument Drift Data Analysis Methodology and Assumptions" for DBNPS. (Attachment 1), which includes a listing of each instrument string and type evaluated in the drift study.
Also attached, as an example, is a copy of the complete instrument drift study for Reactor Protection System (RPS), Reactor Coolant (RC) Flow j
(Attachment 2), and a summary of each of th<
applicable instrument drift study analyses (Attachment 3).
1 I
l J
i
LAR 96-0014 Page 22
REFERENCES:
1.
Davis-Besse Nuclear Power Station (DBNPS) Unit No.
1, Operating License NPF-3, Appendix A, Technical Specifications, through Amendment 214.
2.
Davis-Besse Nuclear Power Station Updated Safety Analysis Report, through Revision 19.
3.
Generic Letter 91-04, " Changes in Technical Specification Surveillance Intervals to Accommodate a 24-Month Fuel Cycle," dated April 2, 1991.
4.
10 CFR 50.59, " Changes, Tests, and Experiments."
5.
NUREG-1430, Revision 1,
" Standard Technical Specifications, Babcock and Wilcox Plants," dated April, 1995.
6.
LAR 95-0019, Toledo Edison Serial Number 2383, dated September 17, 1996.
7.
LAR 95-0022, Toledo Edison Serial Number 2393, dated January 20, 1997.
8.
LAR 95-0024, Toledo Edison Serial Number 2428, dated January 30, 1997.
9.
LAR 95-0027, Toledo Edison Serial Number 2405, dated December 11, 1996.
10.
ISA 567.04, Part I-1994, "Setpoints for Nuclear Safety-Related Instrumentation."
I l
11.
ANSI / ASTM E178-1994, " Standard Practice for Dealing With Outlying i
Observations."
12.
ANSI N15.15-1974, " Assessment of the Assumption of Normality (Employing Individual Observed Values). "
1