ML20129A140
| ML20129A140 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 05/29/1985 |
| From: | Devincentis J PUBLIC SERVICE CO. OF NEW HAMPSHIRE |
| To: | Knighton G Office of Nuclear Reactor Regulation |
| References | |
| SBN-806, NUDOCS 8506040481 | |
| Download: ML20129A140 (21) | |
Text
_
'. MI May 29, 1985 SBN-806 Pub 5c Service of New Hampshire T.F. B7.1.1 New Hampshire Yankee Division United States Nuclear Regulatory Commission Washington, DC 20555 Attention:
Mr. George W. Knighton, Chief Licensing Branch No. 3 Division of Licensing
Reference:
(a) Construction Permits CPPR-135 and CPPR-136, Docket Nos.
50-443 and 50-444
Subject:
Seabrook Station Technical Specification Improvement Program
Dear Sir:
As discussed in our meeting with you on April 9,1985, we are developing a Seabrook Station Technical Specification Improvement Program. This Program, as described in the enclosure, is an effort to make the Seabrook Station Technical Specifications a more useful document for the control room operators. The Program involves both format changes and technical changes designed to improve the Technical Specifications administratively and enhance their safety impact.
The format changes consist of the removal of several tables and programs which are not used by the operators and which are located in other licensee-controlled documents. The technical changes are proposed in order to improve the impact of current Technical Specifications on plant safety while at the same time providing greater flexibility in operating the plant.
All of these changes are proposed within the context of the Standard Technical Specifications, modified to be Seabrook specific. It is acknowledged that further improvements could be made but these would involve additional time in creation of a new Standard and possibly rulemaking. This Program is designed to provide a substantial improvement to Technical Specifications while maintaining the Standard format.
Several detailed supporting analyses referenced in the Program are not complete but will be provided at the time of our submittal of the marked up Seabrook Technical Specifications. This Program of proposed changes is being sent prior to our Technical Specification submittal in order to not impact your review schedule. Please address any questions'or comments to Mr. Warren J. Itall at (603) 474-9574, extension 4046.
Very truly yours, L.l'0 J 8506040481 B50529 PDR ADOCK 05000443 John DeVincentis, Director A
PINI Engineering and Licensing
-JD/cb l
cc: Atomic Safety and Licensing Board Service List l
P.O. Box 300 Seabrook.NHO3874 Telephone (603)474-9521
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ENCLOSURE r
SEABROOK STATION TECHNICAL SPECIFICATION IMPROVEMENT PROGRAM s
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Table of' Contents Page I.
' INTRODUCTION...
2
..............=............
i -
LII.
FORMAT-CHANGES (Table 1 " Format Changes").............. 4
-III.
TECHNICAL CHANGES 4
J '
A.
General.....'...'..................-... 12 i
B.
System Importance.
....-.......-...-...... 13 i
- C.
High. Risk Importance Systems................... :- 15 '
D.
Low Risk Importance Systems (Table 2 " Technical Changes")...-. 15 i
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4 SEABROOK STATION TECHNICAL SPECIFICATION IMPROVEMENT PROGRAM (TSIP) 1.
Introduction The Seabrook Station Technical Spec'ification Improvement Program,
' described herein, is an effort to make the Seabrook Station Technical Specifications a more ~ useful. document for the control room operators..
The Program involves both' format changes and technical changes = designed to' improve the Technical. Specifications administratively and enhance their safety impact. The' format changes consist-of the removal of several tables and programs which are not used' by the operators and -
which are located in licensee-controlled documents. The technical changes are proposed in. order to. improve the impact of current Technical Specifications on plant safety while, at the same time providing r
. greater flexibility in operating the plant.
All of these changes are proposed within the context of the Standard Technical Specifications modified to be Seabrook specific.
It is
= acknowledged that further improvements could be made but these would involve additional time in creation of a new Standard and possibly rulemaking..This Program is designed'to provide a substantial
--improvement to Technical Specifications while maintaining the Standard
~
. format.
The basic philosophy used_in proposing changes to Technical Specifications-is that they should focus on those conditions'important
.to the' control. room operator. A number of current Technical Specifications involve conditions. that are not. controlled by the operator nor used by the operators in day-to-day operation. Also, these same Specifications are covered by licensee-controlled procedures and
- programs, which are subject to NRC inspection. In-addition, a number of current Technical Specifications can lead to unnecessary' plant shutdowns, test-induced plant transients and challenges to safety systems.
A number of general safety improvements were. considered for their
~
applicability to the current Seabrook Technical Specifications.. These-potential safety improvements are briefly described-below and are further referred to in Sections II and III:
1.
Action Statements a) Action statements may be modified to-prevent unnecessary shutdowns for-conditions only slightly "out of spec."
This would reduce the number of plant transients, unnecessary challenges to safety systems, and excessive equipment cycling.
1 b
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b) Equipment allowed outage times may be lengthened to provide more likely repair / restoration times. This would reduce the number of unnecessary plant shutdowns and allow repair and testing to be done with more planning.
c)
"One hour shutdown" action statements may be extended to allow sufficient time for an orderly plant shutdown. This would potentially reduce the number of plant transients and unnecessary challenges to safety systems.
2.
Test / Surveillance Requirements a) Test and surveillance f requencies may be reduced for equipment which.is highly reliable. This would decrease the unnecessary diversion of operator attention.
b) Test frequencies may be optimized. This would assure that the equipment is not out of service too often due to. test, does not wear out due to aver test, or does not remain undetected in a failed state too long.
c) Test and surveillance frequencies may be reduced for equipment in high radiation areas. This would reduce excessive, unwarranted radiation exposure to plant personnel.
d) Test frequencies may be reduced to reduce the likelihood of an inadvertant, test-induced plant transient. This would reduce the number of unnecessary plant transients and unnecessary challenges to safety systems.
e) The test type may be modified if it has a potential to degrade the equipment or if it requires placing the plant into a less safe configuration to perform a test.
f) Testing required af ter a f ailure may be modified to reduce the likelihood of damaging the redundant system. This would ensure that testing provides assurance that the system is available without causing an additional failure.
g) Testing and surveillance frequencies may be modified to more appropriately reflect the importance of the test.
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- II.- Format' Changes The programs and tables listed in Table 1 are proposed to be removed from-Technical Specifications and replaced with teferences to licensee-1 controlled documents which contain the same information. The bases for -
these changes include the following:
- (1).The programs and tables are notLused by control room operators in
. day-t o-day. ope ration..
(2),They-are not of'immediate safety importance.
.(3) ^They tend to clutter up the Technical Specifications,. making them
~
'less useable for operators.
(4) They'are contained in some other adminstrative1y controlled licensee program or. procedure, which is available for NRC inspection.=
1
- These changeslare proposed as format changes alone and it is intended.
that the activities that are indicated within the programs and tables will; be continued.
The' justifications indentified in Table 1. refer to either Seabrook-specific-documents or regulatory documents as applicable.
4.
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-A-r 1
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. TABLE 1
-FORMAT-CHANGESI TECH SPEC SYSTEM / COMPONENT MODE PROPOSED CHANGE-JUSTIFICATION:FOR CHANGE 4.2.1.1.a.2 Axial Flux Difference 1
-Eliminate this subsection
- Needless duplication of effort.-
of the Specification 3/4.2.2 Heat Flux Hot Channel 1
Rename -this Specification
- The current Specification' calls-Factor - F
' Planar Radial Peaking for the LOO on Fq to be met by q
Factor - FXY" Performing a surveillance on Fgy.
This change reflects actual practice and does not involve.
any technical ~ requirements.-
' 3/4.2.3 RCS Flow Rate and Nuclear 1
Eliminate RCS flow rate
- The current Specification has lost Enthalpyl Rise Hot Channel conditions its' significance with the elimina tion of rod bow-(R ).
Thus, this 2
Specification becomes just the FaH spec and RCS flow is covered under 3/4.2.5.
1 3.2.4 Quadrant Power Tilt Ratio 1
Simplify action statement:
- QPTR does not have a safety limit, )
it just indicates that something
~With QPTR determined to abnormal is happening in the core j
exceed 1.02, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> that warrants investigation.
and evary 7 days thereafter, Thus the action is to verify.that verify that FXY and FAH are the peaking factors are within within their limits by per-limits.
forming Surveillance 4.2.2.2-and 4.2.3.2 a
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-TABLE 1 FORMAT CHANGES.(cent'd)
TECll SPEC SYSTEM / COMPONENT MODE PROPOSED CHANGE JUSTIFICATION FOR CHANGE 4.2.4.2 Quadrant Power Tilt Ratio 1
Confirm indicateo QPTR
- QPTR is not expected.to change at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> significantly over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3/4.2.5 DNB Parameters 1
Add RCS flow limits and
- (See 3/4.2.3 Justification)
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Surveillance ' requirements formerly found in 3/4.2.3 3.3.3.2 Movable Incore Detectors Action: with less than the
- Unnecessary shutdown.
required detector thimbles but more than 50%, perform
- Current Specification would
.an evaluation of the reduced require an eventual plant shutdown.
number of detector ~ thimbles if the number of operable thimbles -
to determine what, if any, drop from 43 to 42.
increased unce'rtainty shall be applied to incore measure ment s.
4.3.3.2 Movable Incore Detectors ~
Eliminate Surveillance
- Current Surveillance has little to do with intent of the Speci.
fication;.also, the operability section covers detector normalization and plateau determination.
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- 7 TABLE 1
- FORMAT CHANGES '(c:At'd).
TECH SPEC SYSTEM / COMPONENT'
, MODE
. PROPOSED CHANGE JUSTIFICATION FOR CHANGE-
.3.3.l_
Reactor Trip System Various
- Remove Table 3.3-2 ~ "RTS
- Information contained in an I&C' Instrumentation
' Instrumentation Response Procedure.
Times"
- Engineer.will measure response time and report channel operability, to operators.
Operators will act according to-Table 3.3-1 of Tech Spec 3.3.1.
- Response time measurement made.
only every 18 months.-
3.3.2 ESFAS Instrumentation Various Remove. Table 3.3-5
- Information contained in an I&C
" Engineered Safety Features Procedure.
Response Times"
. Engineer will measure response -
time and report channel'
. operability to operators.
Operators will act according to.
Table 3.3-3 of Tech Spe. 3.3.2.
- Response time'measurementLmade only every 18 months.
- No action required for. response times out of spec.
3/4.3.4 Turbine Overspeed Protection 1,2,3
. Remove Specification;
- Information and. action statements Surveillance Frequency contained'in Maintenance. Procedure.
once per 92 days *
- Action:is inconsistent with low risk importance of system.
- Technical Change proposed. to be made in addition to the format change.
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TABLE 1
. FORMAT CHANGES (cont'd)
TECH SPEC SYSTEM / COMPONENT MODE PROPOSED CHANGE JUSTIFICATION FOR CHANGE 3/4.3.3.8 Loose Parts Detection System 1,2 Remove Specification
- Included in the Surveillance Program under the scope of the Test Control Manual.
3/4.4.5 Steam Generator Inspection 1,2,3,4 Remove Specification
- Action: Operability of S/G's is
. covered in Tech Specs 3.4.l~1, 3.4.1.2, and 3.4.1.3.
- Surveillance:
Included in the ISI/IST Program under the scope of the Test Control Manual.
4.4.9.1.2 Reactor Vessel Material All Remove Specification
- Surveillance:
Included in the Irradiation Surveillance ISI Program Manual-Specimens 3/4.4.10 RCS Structural Integrity All Remove Specification
- Included in the ISI/IST Program under the scope of the Test Control Manual.
3.6.1.2 Containment Leakage 5
Remove Table 3.6-1
- Information included in the Test Control Manual.
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3/4.6.1.6 Containment Structural 1,2,3,4 Remove Specification
- Surveillance under the scope of Integrity the Test Control Manual.
3/4.7.9 Fire Suppression Water All Remove Specification
- Surveillance under the scope of System the Test Control Manual and the Fire Protection Manual.
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. TABLE 1' FORMAT CHANGES (cont'd)'
TECH SPEC SYSTEM / COMPONENT MODE
~
PROPOSED CHANGE
> JUSTIFICATION FOR CHANGE-3/4.7.10 Fire Rated Assemblies.
All Remove Specification
- Surveillance under the scope of the Test Control Manual:and the Fire Protection Manual.
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-3/4.7.7 Snubbers 1,2,3,4 Remove Specification,
- Included in the ISI/IST Program 5,6 including Tables 3.7-3 a.
under the scope of the Test and b.
Control Manual.
- Engineer will perform visual inspection and functional test of snubbers.
If a snubber is I
inoperable and not repairable in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, operators will declare attached system inoper-f able and follow that action statement.
j 3.8.4.2 Containment Penetration 1,2,3,4 Remove Table 3.8-1
- Information contained in Conductor Overcurrent an Electrical Maintenance Protective Devices Procedure.
- Engineer will determine status of the protective devices and inform the operators. The operators will perform the action based on this information.
.10 TABLE l' FORMAT CHANGES (cent'd)
TECH SPEC
. SYSTEM / COMPONENT MODE PROPOSED CHANGE
.IUSTIFICATION FOR CHANGE 3.8.4.3-MOV Thermal Overload Various Remove Specification,
- This information and action Devices including Table 3.8-2 statement is contained in an Electrical Maintenance Procedure.
- Engineer will determine, status of device and report to operators. Operators will take appropriate action based on valves declared inoperable.
4.10.2.2 Special Text Exceptions -
1 Eliminate Surveillance
- Current Surveillance requires a Group Height, Insertion full core flux map.at least once and Power Distribution per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, which is of little Limits or no value during the Special i
Test which requires it.
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-l 3/4.11.1.1 Liquid Effluents' All Remove Table 4.11-1
- Releases are controlled ' by. the Concentration ODCM, which is an NRC-approved document based on 10 CFR 50 Appendix I and 10 CFR 20.
Surveillance frequencies are covered by Effluent Surveillance Program.
3/4.11.2.1 Gaseous Effluents Dose Rate All Remove Table 4.11-2
- Releases are ' controlled by the ODCM, which is aa NRC-approved document based on 10 CFR 50 Appendix I and.10 CFR 20.
- Surveillance' frequencies are covered by Effluent Surveillance Program.
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- 11 TABLE 1 FORMAT CHANGES (cent'd)
TECH SPEC SYSTEM / COMPONENT MODE PROPOSED CHANGE JUSTIFICATION FOR CHANGE.
3/4.12.1 Radiological Environmental All Remove Specification
- Program is included in the HP Monitoring Program procedures, Seabrook Station Environmental Report, and Test Control Manual.
3/4.12.2 Land Use Census All Remove Specification
- Program is included in the HP procedures, Seabrook Station Environmental Report, and Test Control Manual.
3/4.12.3 Interlaboratory Comparison All Remove Specification
- Program is included in the Yanke:
Program Atomic Environmental Laboratory Procedures.
6.2.3 Operational Engineering Remove all this section
- Information will be in FSAR Section except a summary paragraph 5 13.4.3.
6.5.1 Station Operation Review Remove all this section
- Information will be in FSAR Committee except a summary paragraph 5 13.4.1.
6.5.2 Nuclear Safety Audit and Remove all this section
- Information will be in FSAR Review Committee except a summary paragraph 5 13.4.2.
6.9 Reporting Requirements All Remove Specification
- Included in Station Reporting Manual.
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III. Technical Changes A.
General A number of technical changes to the Technical Specifications are proposed based on risk analysis and on engineering judgment. To justify these changes, the following steps were used and are documented herein:
- 1).The importance of systems and components addressed by Technical Specifications was determined by calculating risk importance measures and by use of engineering judgment. The risk importance measures were calculated by using methodology developed by Battelle Labs and applying it to the Seabrook Station Probabilistic Safety Assessment (SSPSA). The output of this analysis yielded an assignment of systems into HIGH and LOW categories of relative risk worth.*
- 2) For systems with a HIGH risk importance worth, the Technical Specifications were evaluated and modified as appropriate to assure that the Specification is optimized with regard to risk.**
- 3) For systems with a LOW risk importance worth, the Technical Specifications were examined for actions which consume operator time unnecessarily, have' potential for causing plant trip, result in undue radiation exposure to plant personnel, and generally are inappropriate to the safety significance of the action. The Technical Specifications were modified where such a change will reduce the complexity of the Specifications, reduce the potential for plant trip, and generally ease the operating restrictions of the Technical Specificationa while not drastically affecting the reliability of the system.
The evaluations described above were performed using insights from the SSPSA and the engineering judgment of operators and technical personnel. The SSPSA was used to determine the risk importance of systems in the Technical Specifications. The "importance" results were determined quantitatively; however, the results could have been determined for the most part from a qualitative inspection of the important sequences. The "importance" results are rigorous and do not depend on indvidual numerical calculations in the SSPSA.
- The detailed calculations of system importance will be submitted along with the submittal of the marked-up Seabrook Technical Specifications.
The determination of system importance which is summarized in Section III.B was based on engineering judgment and a qualitative consideration of the SSPSA. The detailed calculations will provide quantitative support for the system importance analysis but is not expected to affect the results.
- The detailed evaluation and optimization of the Technical Specifications for HIGH risk important systems will be submitted along with the submittal of the marked-up Seabrook Technical Specifications.
12
After determining risk ~importance, detailed analyses of the HIGH risk important'cystems were done with regard to Technical Specifications to ensure that the allowed outage times, surveillance / test types and frequencies, and maintenance activities
-are optimized. These analyses used the system analysis documented
~in the SSPSA as the basis for sensitivity calculations. The results oof this study indicate the important Technical Specification parameters and potential changes 'that can be made to maintain or improve the reliability of HIGH risk important systems and components.
For the rest lof 'the systems (the so-called LOW risk importance systems), engineering judgment was the primary tool to justify deviations from the Standard Technical Specification.
This assignment of systems into HIGH and LOW risk importance categories permitted-the detailed, quantitative risk optimization of Technical. Specifications to be targeted to those systems that are the most important with regard to plant risk. Technical changes to other systems were justified based on qualitative, engineering judgment. The results are described in the following sections.
B.
System Importance*
The risk importance of systems was determined by calculating risk importance factors and by using engineering judgment to assess the importance of systems to initiating events, containment failure and offsite release, and external events. The results are displayed below. The systems / components of HIGH risk importance are listed with a basis for their inclusicn.
The systems not listed below.are designated " LOW risk importance" systems. This designation does not imply that these systems are unimportant to plant safety or that they could be eliminated without affecting risk..Also the LOW risk importance systems are not necessarily at the same level of importance. Howevnt, for LOW risk importance systems, it is assumed that a change in the related Technical Specifications would have less effect on plant risk than for HIGH risk importance systems. Thus, minor changes to Technical Specifications for LOW risk importance systems can be justified without a detailed risk-sensitivity analysis.
-Systems designated "HIGH risk importance" are systems whose failures contribute to the most likely core melt or fission product release sequences.- Thus, a system appears on the HIGH importance list because of its critical safety importance and because it appears in a relatively high frequency core melt / release sequence.
- The detailed calculations of system importance will be submitted along with the marked-up Seabrook Technical Specifications.
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The following systems are judged to be "HIGH risk importance systems" with a brief justification for that designation. The details with regard to the basis for including systems in this' list will be' submitted later.
1.
A.C. Electric Power -
- Loss of all A.C. electric power (station blackout) is the most frequent category of core melt sequences. This involves loss of offsite power, failure of the diesels, and failure to recover electric power before core damage.
2.
This is important because of the dependency of diesels on service water. Failure of this system is a contributor to
" station blackout" sequences.
3.
Primary Component Cooling Water -
This is important because of the dependency of.high pressure injecticn pumps and RCP seal cooling on PCCW cooling. Failure of this system leads to small LOCA (leakage out the seals) with no makeup injection.
4.
Emergency'Feedwater -
l' This is important because of the importance of transient events, which usually demand EFW to respond. This system is also i
important in delaying the time to core melt in " station blackout" sequences, which affects likelihood of recovery of emergency power.
5.
Containment Isolation -
This is important to offsite releases because of the analyzed 4-failure modes of containment.
If the containment is. intact at time of core melt, the most likely failure would be late overpressure failure which occurs one'to three days after the accident. For this late failure, settling of the fission products plus complete evacuation contribute to make this-release and the resultant health effects relatively small.
Thus, to get a significant source term release, the containment must be open at time of core melt, i.e.,
failure of containment isolation, and more specifically, failure of the purge valves to close if open.
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s C.
High Risk Importance Systems High Risk importance systems listad in Section III.B have been analyzed with respect to their Technical Specifications to ensure that the e' lowed outage-times, test / surveillance times and types, and maintenance activities are optimized. The detailed analysis and the proposed changes coming out of the analysis will be submitted later along with the marked-up Seabrook Technical Specifications.
D.
Low Risk Importance Systems-(see Table 2)
The proposed changes to Tech Specs for low risk importance systems are given in Table 2.
The justifications for changes listed are based on consideration of the potential safety improvements I
described in Section I.
1 15
116' TABLE 2 TECHNICAL CHANGES
c TECH SPEC SYSTEM / COMPONENT MODE PROPOSED CHANGE' JUSTIFICATION FOR CHANGEJ
- 2.1.1 Safety Limits -
1,2 Be in HOT STANDBY
- One hour shuidown greatly Reactor Core within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> increased the' likelihood of-human' error causing a plant trip or other less safe plant conditions.
2.1.2 Safety. Limits -
1,2,3,4,5 Be in HOT STANDBY
- One hour shutdown greatly.
RCS pressure with RCS pressure increases the likelihood ofL within its. limit human error causing a' plant within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> trip or other less safe plant conditions.-
3.4.1.1 Reactor Coolant Loops 1,2 Be in at least-HOT STANDBY
- One hour shutdown greatly and Coolant Circulation within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> increases the likelihood of human error' causing a-plant trip or other less safe pl~ ant conditions.
4.4.1.1 Rea<: tor Coolant Loops 1,2 Eliminate surveillance
- Surveillance not necessary, requirement condition is obvious to operators by alarms and indicators, unnecessary diversion of operators.
3.4.3a Pressurizer Heaters 1,2,3 With-either Groups A'or B
- 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allowed outage time inoperable, restore too short based on low safety
-inoperable group to Operable
' significance of heaters.-
status within 7 days or...
- Only Groups A and'B are used in
. Safe Shutdown analysis.
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' TABLE 2' TECHNICAL CHANGES (cont'd)
TECH SPEC SYSTEM / COMPONENT MODE-PROPOSED CHANGE JUSTIFICATION FOR CHANGE' 4.4.3.2 Pressurizer Heaters 1,2,3 Tested once every 18 months
- 92 day surveillance interval too short, heaters don't change drastically in power over.18 months.
- Test is time consuming, involves personnel hazard.
3.4.6.2 Reactor Coolant System 1,2,3,4 Remove Specification
- Specification not necessary; RCS and Pressure Isolation Valves subsections referring to pressure isolation valve 4.4.6.2.2 RCS. Pressure Isolation
. leakage would be discovered Valves, including Action when measuring identified or Statement C and Table 3.4-1 unidentified leakage and thus would require action if these limits were exceeded.
- Testing of these valves will be performed in accordance with the ISI Program.
3.5.1.la Accumulators 1,2,3 Restore inoperable
- I hour allowed outage time too accumulator (except as a short to accomplish most likely-result of. closed isolation repairs.
valves) to operable within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
- Accumulators are needed in accident analysis for large LOCA which is a relatively unimportant risk contributor.
18 TABLE 2 TECHNICAL CHANGES (cant'd)
TECH SPEC SYSTEM / COMPONENT MODE PROPOSED CHANGE JUSTIFICATION FOR CHANGE 4.5.1.1 Accumulators 1,2,3 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> surveillance
- 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> surveillance'too interval frequent, occupies operator time and attention unnecessarily; operator will respond to alarms.
3.5.2 ECCS Subsystems 1,2,3 Restore inoperable
- 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allowed outage time too subsystem to operable short to accomplish most likely within 7 days repairs, requiring unnecessary shutdowns.
- Longer allowed outage times do not affect ECCS system reliability.
4.5.2a ECCS Subsystems 1,2,3 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> surveillance
- 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> surveillance too interval for valve frequent, occupies operator time alignment and attention unnecessarily.
3.5.5 RWST 1,2,3,4 2nd Level Specification:
- Most likely inoperable if volume or boron conditions are slightly out of concentration is out of spec level or concentration, spec by no more than 10%,
which have minimal ef fect on restore to operable within ability of RWST to function, 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
requiring unnecessary shutdowns.
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'19 TABLE 2 TECHNICAL CHANGES (c:nt'd)
TECH SPEC SYSTEM / COMPONENT MODE PROPOSED CHANGE JUSTIFICATION FOR CHANGE 3.6.2.1 Containment Building Spray 1,2,3,4 Restore inoperable system
- 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allowed outage time too to operable within 7 days short to accomplish most likely repairs, requiring unnecessary shutdowns.
- Longer allowed outage time does not affect ECCS system reliability.
3.6.2.2 Spray Additive System 1,2,3,4 Restore inoperable system
- System of very low importance, to operable within 31 days compared to CBS, action state-ment not appropriate to risk significance of system outage.
- Risk from plant shutdown is much greater than the risk from having SAS inoperable.
4.6.2.2a Spray Additive System 1,2,3,4 Verify valve alignment at
- System of very low risk least once per 6 months importance.
- Valve alignment very unlikely to change in 6 month period.
4.6.4.1 Hydrogen Monitors 1,2 Channel check once per 7
- 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> channel check is not days consistent with importance of Hydrogen Monitors as reflected by the 30 day allowed outage time.
..w
20 TABLE 2 TECHNICAL CHANGES (cont'd)
TECH SPEC SYSTEM / COMPONENT MODE PROPOSED CHANGE JUSTIFICATION FOR CHANGE 4.6.4.2a Hydrogen Recombiners 1,2 System functional test
- Unnecessary cycling of equipment.
every 18 months 3.6.5.1 Containment Enclosure 1,2,3,4
' Restore integrity within
- 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> allo <ed outage time is Building Integrity 7 days too short, not appropriate to to the low risk significance of enclosure building.
- Enclosure building not important in reducing offsite releases from a core damage / melt accident.
3.6.5.3 Containment Enclosure 1,2,3,4 Restore integrity within
- (See 3.6.5.1 Justification
.above)
Building Structural 7 days Integrity 4.7.6a Control Room Makeup Air All Eliminate verifying control
- Unnecessary diversion of System room temperature is less operators, such a condition will than 120*F become obvious long before the temperature approaches 120*F.
.