ML18096A085
| ML18096A085 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 06/20/1991 |
| From: | Miltenberger S Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML18096A086 | List: |
| References | |
| GL-90-06, GL-90-6, LCR-91-06, LCR-91-6, NLR-N91095, NUDOCS 9106260285 | |
| Download: ML18096A085 (19) | |
Text
Public Service Electric and Gas Company Steven E. Miltenberger Public Service Electric and Gas Company P.O. Box 236, Hancocks Bridge, NJ 08038 609-339-1100 Vice President and Chief Nuclear Officer JUN 2 o 1991 NLR-N91095 LCR 91-06 United States Nuclear Regulatory Commission Document Control Desk Washington, D.C.
20555 Gentlemen:
REQUEST FOR LICENSE AMENDMENT INCORPORATION OF GENERIC LETTER 90-06 REQUIREMENTS SALEM GENERATING STATION UNITS 1 AND 2 FACILITY OPERATING LICENSE NOS. DPR-70 AND DPR-75 DOCKET NOS. 50-272 AND 50-311 In accordance with the requirements of 10 CFR 50.90, Public Service Electric and Gas Company (PSE&G) hereby submits a request for amendment of Facility Operating Licenses DPR-70 (Unit 1) and DPR-75 (Unit 2) of the Salem Generating Station.
Pursuant to the requirements of 10 CFR 50.91(b) (1), a copy of this submittal has been sent to the state of New Jersey as indicated below.
The proposed change would incorporate the requirements of NRC Generic Letter 90-06 pertaining to Power Operated Relief Valve and Block Valve operability and low-temperature overpressure protection for light-water reactors. of this submittal provides detailed description and justification for the proposed changes and Attachment 2 contains marked up Technical Specifications reflecting the proposed changes.
PSE&G believes that this amendment request submittal includes adequate technical justification to conclude that a detailed specialist review should not be required and can therefore be classified as Category 2.
Upon NRC approval, please issue a License Amendment which will be effective upon issuance and shall be implemented within 60 days of issuance.
9106260285 910620 PDR ADOCK 05000272 P
~
Document Control Desk NLR-N91095 JUN 2 o 1991 Should you have any questions on this submittal, please do not hesitate to contact us.
Thank you.
Sincerely, Attachments (2)
C Mr. J. c. Stone Licensing-Project Manager -
Salem Mr. T. Johnson Senior Resident Inspector Mr. W. T. Russell Administrator - Region I Mr. Kent Tosch Chief -
New Jersey Department of Environmental Protection Division of Environmental Quality Bureau of Nuclear Engineering CN 415 Trenton, NJ 08625
Ref:
NLR-N91095 LCR 91-06 STATE OF NEW JERSEY SS.
COUNTY OF SALEM
- s. E. Miltenberger, being duly sworn according to law deposes and says:
I am Vice President and Chief Electric and Gas Company, and forth on our letter dated Salem Generating station, are information and belief.
My Commission expires on Nuclear Officer of Public Service as such, I find the matters set
, concerning the true to the best of my knowledge, ELIZABETH J. KIDD Notary Public of New Jersey My Commission Expires April 25, 1995
ATTACHMENT 1 REQUEST FOR LICENSE *AMENDMENT INCORPORATION OF GENERIC LETTER 90-06 REQUIREMENTS SALEM GENERATING STATION UNITS 1 AND 2 FACILITY OPERATING LICENSE NOS. DPR-70 AND DPR-75 DOCKET NOS. 50-272 AND 50-311
I.
Description of Change SALEM LCR 91-06 NLR-N91095 This amendment change request proposes that existing Technical Specification 3.1.2.3, "CHARGING PUMP -
SHUTDOWN", be administratively revised such that:
A surveillance is added to verify that all safety injection pumps and centrifugal charging pumps are inoperable except for the single pump required to be operable.
This guidance is currently provided in technical specification 3.5.3 but is being restated here for clarity.
A note is added to specify that a maximum of one safety injection pump or one centrifugal charging pump shall be operable in Mode 5 or Mode 6 when the head is on the reactor vessel.
This guidance is currently provided in technical specification 3.5.3 but is being restated here for clarity.
This amendment change request proposes that existing Technical Specification 3.1.2.4, "CHARGING PUMPS -
OPERATING", be administratively revised such that:
A surveillance is added to verify that all safety injection pumps and centrifugal charging pumps are inoperable except for the single pump required to be operable while in Mode 4 and the temperature of any cold leg is less than or equal to 312°F.
This guidance is currently provided in technical specification 3.5.3 but is being restated here for clarity.
A note is added to specify that a maximum of one safety injection pump or one centrifugal charging pump shall be operable in Mode 4 when the temperature of any cold leg is less than or equal to 312°F.
This guidance is currently provided in technical specification 3.5.3 but is being restated here for clarity.
This amendment change request proposes that existing Technical Specification 3/4.4.3 for DPR-70 (Unit 1) and 3/4.4.5 for DPR-75 (Unit 2), "RELIEF VALVES", be revised such that:
With one or both Power Operated Relief Valves (PORVs) inoperable due to excessive leakage, continued plant operation shall be permitted only if the associated Block Valve(s) is closed with power maintained.
PAGE 1 OF 10 ATTACHMENT 1
SALEM LCR 91-06 NLR-N91095 With one PORV inoperable for reasons other than excessive leakage, continued plant operation shall be permitted only if the associated Block Valve is closed and de-energized within one hour and the affected PORV is returned to operable status within 7 days.
With both PORVs inoperable for reasons other than excessive leakage, continued plant operation shall be permitted only if at least one PORV can be restored to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
With one or both Block Valve(s) inoperable, either the valve(s) shall be restored to operable status or the associated PORVs shall be placed in manual control within one hour.
If both Block Valves are inoperable, at least one shall be restored to operable status within the next 24
.hours.
Any remaining inoperable Block Valve shall be restored to operable status within 7 days.
Each PORV shall be demonstrated operable on an 18 month test interval by: 1) operating the valve through one complete cycle of. travel during Modes 3.or 4, 2) operating solenoid, control, and check valves associated with the PORV accumulators through one complete cycle of travel, and 3) performing a channel calib~ation of. the actuation instrumentation.
- Each Block Valve shall be demonstra.ted operable on a 92 day test interval by operating the valve through one complete cycle of travel.
This amendment change request proposes that existing Technical Specification 3.4.9.3 for DPR-70 and 3.4.10.3 for DPR-75, "OVERPRESSURE PROTECTION SYSTEMS", be revised such that:
With one PORV inoperable in Mode 4, the inoperable PORV shall be restored to operable status within 7 days or the reactor coolant system shall be vented within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
With one PORV inoperable in Modes 5 or 6, the inoperable PORV shall be restored to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or the reactor coolant system shall be vented within a total of 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br />.
The reference to the specific ASME valve category is deleted from the surveillance requirements for consistency (applicable to DPR-70 only).
PAGE 2 OF 10 ATTACHMENT 1
SALEM LCR 91-06 NLR-N91095 Finally, this amendment change request proposes that existing Technical Specification 3.5.3, "ECCS SUBSYSTEMS - Tave< 350°F",
be administratively revised such that:
The applicability of the note pertaining to the LCO and surveillance 4.5.3.1 is reworded for clarity.
II.
Reason for Proposed Change PSE&G is submitting this amendment request in partial fulfillment of the requirements set forth in NRC Generic Letter (GL) 90-06 dated June 25, 1990.
GL 90-06 provided the staff's positions on the resolution of Generic Issue 70 (GI-70), "Power-Operated Relief Valve and Block Valve Reliability", and Generic Issue 94 (GI-94), "Additional Low-Temperature Overpressure Protection for Light-Water Reactors".
GI-70 involved the evaluation of the reliability of power-operated relief valves (PORVs) and block valves and their safety significance in PWR plants.
There are various plants at which neither the PORVs nor block valves have traditionally been classified as safety-related (this does not apply to Salem Units 1 or 2 where the PORV and block valves are classified as Safety-Related and environmentally qualified).
Consequently, the Technical Specifications governing PORVs on most operating PWRs including the Salem units, which deal with closing the block valves and removing power, were developed to allow continued plant operation with degraded PORVs, but did not consider the need for PORVs to perform any safety related functions.
Following the TMI-2 accident, the NRC initiated an effort to evaluate the role of PORVs in performing certain safety-related functions.
Consequently, the NRC determined that over a period of time, the role of PORVs has changed such that PORVs are now relied upon by various plants to perform one, or more, of the following safety-related functions:
- 1.
Mitigation of a design-basis steam generator tube rupture accident,
- 2.
Low-temperature overpressure protection of the reactor vessel during startup and shutdown, and/or,
- 3.
Plant cooldown in compliance with Branch Technical Position RSB 5-l*to*sRP 5.4.7, "Residual Heat Removal (RHR) System".
PAGE 3 OF 10 ATTACHMENT 1
SALEM LCR 91-06 NLR-N91095 Based on these findings, it was determined that the safety classification of PORVs and block valves should be reconsidered.
The NRC subsequently issued a list of actions in GL 90-06 to be taken by PWR plants to increase the reliability of PORVs and block valves and provide assurance that they* will function as required.
One of the required actions delineated in GL 90-06 is to modify the existing technical specifications for the PORVs and block valves in Modes 1, 2, and 3.
GI-94 arose as a result of continuing low-temperature overpressure events and the unavailability of Low-Temperature Overpressure Protection (LTOP) channels.
PORVs are relied upon by most Westinghouse PWRs, Salem included, to provide low-temperature overpressure protection.
Based on the NRC evaluation of the LTOP system unavailability, it was concluded that additional restrictions are warranted and that existing technical specifications should be modified for the affected plants.
The technical specification changes encompassed by this amendment request are those that have resulted from the resolutions of GI-70 and GI-94 and which have been delineated and justified by the NRC in GL 90-06.
Where differences exist between the modified technical specifications provided in GL 90-06 and those contained in this *amendment request, full justification has been provided.
III. Justific~tion for the Proposed Change Specifications 3.1.2.3, "CHARGING PUMP -
SHUTDOWN". and 3.1.2.4, "CHARGING PUMP -
OPERATING".
The changes that are proposed for specifications 3.1.2.3 and 3.1.2.4 are in response to the requirements delineated and justified in GL 90-06, Enclosure B, pages B-10 and B-11.
These requirements provide Low Temperature Overpressure Protection (LTOP) and apply when: 1) the plant is in Mode 4 and any RCS cold leg temperature is less than or equal to 312°F (plant specific value), 2) Mode 5, and 3) Mode 6 when the head is on the reactor vessel.
LTOP is provided by allowing a maximum of one safety injection pump or one centrifugal charging pump to be operable thereby limiting the largest potential mass addition to the RCS to within the capability of the LTOP system.
This guidance is currently contained in specification 3.5.3, "ECCS -
T ave < 350°F".
However, since specification 3.5.3 is only applicable in Mode 4, it was felt prudent.to restate the guidance in specification 3.1.2.3, applicable in Modes 5 and 6, and specification 3.1.2.4, applicable in Modes 1, 2, 3, and 4, for improved clarity.
PAGE 4 OF 10 ATTACHMENT 1
SALEM LCR 91-06 NLR-N91095 As these changes do not delineate any new or differing guidance and are proposed for clarity only, they are considered administrative in nature.
Specification 3/4.4.3 CDPR-70, Unit ll and 3/4.4.5 CDPR-75, Unit
- 2), "RELIEF VALVES 11
- The changes that are proposed for specification 3/4.4.3 (DPR-70, Unit 1) and 3/4.4.5 (DPR-75, Unit 2) are in response to the requirements delineated in GL 90-06, Enclosure A, pages A-4, A-5, and A-7 and justified on pages A-8 through A-10.
These requirements are based on improving the reliability of PORVs and block valves.
The modified technical specifications contained on pages A-4, A-5, and A-7 of Enclosure A to GL 90-06 have been fully incorporated into this amendment request with the following exceptions:
The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Allowed Outage Time (AOT) specified in Action b has been extended to 7 days.
The one hour AOT specified in Action c has been extended to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and one hour AOTs specified in Action d have been extended to 7 days and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> respectively.
The surveillance requirement to test the emergency power supply for the *PORVs and block valves has been deleted.
As committed to the NRC by PSE&G in letter NLR-N90234 (December 21, 1990), a Probabilistic Risk Assessment (PRA) of the technical specifications recommended in GL 90-06 and as modified by PSE&G, by extending AOTs as described above, has been completed (reference SCI-91-0204 dated April 4, 1991).
This PRA report is available for NRC review upon request.
The following discusion is based on the results of the study.
The three action statements were analyzed individually and then combined to determine the total effect on Core Damage Frequency (CDF)
- It was conservatively assumed for this analysis that all PORV failures were due to causes other than excessive leakage.
This assumption tended to magnify the effects of the proposed technical specifications thereby providing a measure of the maximum change that is expected to occur to the CDF.
PAGE 5 OF 10 ATTACHMENT 1
SALEM LCR 91-06 NLR-N91095 Historical data used for this PRA is based on componenet outage durations over a 6 year time period taken from the control room operator's logs.
For Unit 1, this resulted in average outage durations of 704.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />s/outage for the PORVs and 213 hour0.00247 days <br />0.0592 hours <br />3.521825e-4 weeks <br />8.10465e-5 months <br />s/outage for the block valves:
For Unit 2 the values were 6.06 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />s/outage for the PORVs and 36.06 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />s/outage for the block valves.
Additionally, in order account for the failure of a second PORV or block valve, a failure frequency was devleloped based on a testing interval of 18 months (13140 hours).
- Again, these assumptions are conservative and over estimate the impact of the proposed AOTs, but provide a limit to their maximum effect on CDF.
The first action statement (refer to Table 1) proposed by PSE&G defined an extended 7 day AOT for maintenance if one PORV is declared inoperable for causes other than seat leakage.
For Salem Unit 1, the 7 day AOT increased PORV availability, decreasing the CDF from 5.7856E-05 events per year to 5.6351E-05 events per year.
Conversely, for Salem Unit 2, the AOT decreased PORV availability, increasing the CDF from 4.0685E-05 to 4.1359E-05 events per year.
The second action statement (refer to Table 2) defined an extended 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> AOT for the first PORV when two are declared inoperable.
This AOT had a much smaller effect on the CDF for both units.
For Salem Unit 1, the CDF was decreased from 5.7864E...;05 to 5.7775E-05 events per year and for Salem Unit 2, the CDF was decreased from 4.0695E-05 to 4.0681E-05 events per year.
Both changes are less than 0.1%.
The final action statement (refer to Table 3) combined the previously defined extended AOTs and applied them to the block valves.
For Salem Unit 1, the CDF was decreased from 5.7856E-05 to 5.7680E-05 events per year and for Salem Unit 2, the CDF was increased from 4.0685E-05 to 4.1509E-05 events per year.
With respect to the individual analysis of the AOTs recommended by GL 90-06 it was found that the same trends in decreasing the Unit 1 CDF and increasing the Unit 2 CDF were shown.
- However, because the AOTs were of shorter.duration than those proposed by PSE&G, the decrease in Unit 1 CDF was marginally greater and the increase in Unit 2 CDF was marginally.smaller.
Based on the results of the individual analysis, the overall effect of the combined action statements understandably differed for each Salem unit (refer to Table 4).
For Salem Unit 1, the overall CDF decreased from 5.7856E-05 to 5.6112E-05 events per year or 3%.
However, for Salem Unit 2, the overall.CDF increased from 4.0685E-05 to 4.2192E-05 events per year or 3.7%.
PAGE 6 OF 10 ATTACHMENT 1
SALEM LCR 91-06 NLR-N91095 The reason the extended AOTs resulted in a decrease of the CDF for Unit 1 is because outages, on an average, have historically had longer durations than the proposed AOTs.
Therefore, for Unit 1, the proposed AOTs will increase the availability of the PORVs and block valves.
Conversely, the outage durations have historically been shorter than the proposed AOTs at Unit 2.
Therefore, assuming that the full AOTs are utilized, the availability of the PORVs and block valves will decrease, thus increasing the CDF.
With respect to the combined action statements containing the AOTs recommended by GL 90-06, a similar decrease in CDF for Salem Unit 1 and increase in CDF for Salem Unit 2 is calculated.
The same causal factors apply to the results of this calculation as applied to the calculation associated with the AOTs proposed by PSE&G.
For Unit 1, the combined action statements containing the AOTs recommended by GL 90-06 decreased CDF from 5.7856E-05 to 5.5509E-05 events per year.
When compared to the AOTs proposed by PSE&G, this represents an additional decrease of only 6.03E-07 events per year or 1.1%.
For Unit 2, the GL 90-06 AOTs caused an increase in CDF from 4.0685E-05 to 4.1180E-05.
This increase is only 1.0lE-06 events per year less than that caused by the PSE&G proposed AOTs, a difference of 2.5%.
These results indicate that when compared to the AOTs proposed by PSE&G, the AOTs recommended by GL 90-06 offer only a small benefit for the restrictive operational limits they impose.
The analysis primarily indicate that the technical specification changes proposed by PSE&G would improve the overall safety and performance of Salem Unit 1.
With respect to Unit 2, the analysis indicate that the proposed AOTs would have an insignificantly negative effect only if the full AOTs are utilized.
However, when the following factors are taken in to account the extended AOTs are not expected to have any adverse effect on CDF or plant performance:
- 1.
Operational experience indicates that much shorter outage durations occur than those that.would be permitted by either GL 90-06 or the extended AOTs proposed by PSE&G.
- 2.
A high degree of conservatism has been built into the study by assuming all PORV failures to be caused by other than excessive leakage.
- 3.
The magnitude of changes in CDF indicated by this study are insignificant enough to be bounded by the accuracy of the input data.
PAGE 7 OF 10 ATTACHMENT 1
1-SALEM LCR 91-06_
NLR-N91095 Finally, the analysis indicate that the more restrictive AOTs recommended by GL 90-06 offer a very marginal increase in plant safety and performance.
Therefore, PSE&G feels that the AOTs proposed in this amendment request are justified.
The surveillance requirement to test the emergency power supply for the PORVs and block valves has not been incorporated into this amendment request because these valves receive power-from safety-related, diesel-backed busses.
The operability of these busses is verified by surveillance requirements pertaining to electrical power systems (Technical Specification 3.8.1.1).
Specification 3.4.9.3 CDPR-70, Unit 1) and 3.4.10.3 CDPR-75. Unit 2), "OVERPRESSURE PROTECTION SYSTEMS".
The changes that are proposed for specifications 3.4.9.3 (DPR-70, Unit 1) and 3.4.10.3 (DPR-75, Unit 2) are in response to the requirements delineated in GL 90-06, Enclosure B, pages B-6 and B-7 and justified on pages B-8 and B-9.
These requirements seek to instill improved administrative restrictions on the Low Temperature overpressure Protection (LTOP) System thereby improving availability when the potential for an overpressure event is the highest and especially during water-solid conditions.
Salem Units 1 and 2 currently have technical specifications pertaining to PORVs used for LTOP; therefore, the only changes that have been proposed for these sections are to restrict the applicability of Action a to Mode 4 and to incorporate Action b from the modified technical specifications delineated in GL 90-06, Enclosure B, pages B-6 and B-7.
This is in compliance with the instructions given on page B-8 of the GL.
One additional administrative change has been proposed for the Unit 1 technical specification.
The reference to the specific ASME valve category has been deleted from sueveillance requirement 4.4.9.3.1 (DPR-70).
This type of information is contained in the In-Service -Test Program and.is not normally included in technical specifications.
Specification 3.5.3, "ECCS SUBSYSTEMS -
T ave < 350°F 11
- Changes are proposed for specification 3.5.3 in order to mar~
clearly specify the applicability of Surveillance 4.5.3.1. and the note pertaining to the LCO.
The surveillance and note apply when: 1) the plant is in Mode 4 and any RCS cold leg temperature is less than or equal to 312°F (plant specific value), 2) Mode 5, and 3) Mode 6 when the head is on the reactor vessel.
PAGE 8 OF 10 ATTACHMENT 1
SALEM LCR 91-06 NLR-N91095 Since this specification is applicable only in Mode 4, the guidance provided by the surveillance and note has been restated in specification 3.1.2.3, applicable in Modes 5 and 6, and specification 3.1.2.4, applicable in Modes 1, 2, 3, and 4.
As these changes do not delineate any new or differing guidance and are proposed for clarity only, they are considered administrative in nature.
PAGE 9 OF 10 ATTACHMENT 1
SALEM LCR 91-06 NLR-N91095 IV.
Significant Hazards Consideration PSE&G has, pursuant to amendment to deterinine hazards consideration.
Salem Units 1 and 2 in 10 CFR 50.92, reviewed the proposed whether our request involves a significant We have determined that operation of accordance with the proposed change:
- 1.
Will not involve a significant increase in the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated.
Neither PORV nor block valve operability is assumed in any of the events analyzed in the Updated Final Safety Analysis Report (UFSAR).
Therefore, the proposed amendment does not involve a physical or procedural change to any structure, component, or system that significantly affects accident/malfunction probabilities or consequences previously evaluated in the UFSAR.
- 2.
Will not create the possibility of a new or different kind of accident from any previously evaluated.
The proposed amendment does not involve any physical changes to plant structures, components, or systems.
- 3.
Will not involve a significant reduction in a margin of safety.
As described in the Justifications section of this submittal, PRA results and operational experience indicate that this amendment will not have any significant impact on core damage frequency and will therefore not adversely impact any margin of safety.
- v.
Conclusions Based on the above discussions and those presented in the Justification Section, it has been determined that the proposed Technical Specification revisions do not involve a significant increase in the probability or consequences of an accident over previous evaluations, create the possibility of a new or different kind of accident, or involve a significant reduction in a margin of safety.
Therefore, the requested license amendment does not involve a significant hazards consideration.
PAGE 10 OF 10 ATTACHMENT 1
SALEM LCR 91-06 NLR-N91095 REACTOR COOLANT SYSTEM 3/4.4.4 RELIEF VALVES As specified in GL 90-06 Action b. With one PORV inoperable due to causes other than excessive seat leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV to OPERABLE status or close its associated block valve and remove power from the block valve; restore the PORV to OPERABLE status within the following 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
As proposed by PSE&G The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> AOT will be extended to 7 days.
SOURCE GL 90-06 PSE&G HISTORICAL SOURCE GL 90-06 PSE&G HISTORICAL SALEM UNIT 1 ACTION b EFFECT ON CDF AOT (HRS)
CDF 72 168 704.5 5.6084E-05 5.6351E-05 5.7856E-05 SALEM UNIT 2 ACTION b EFFECT ON CDF AOT (HRS)
CDF 72 4.0958E-05 168 4.1359E-05 6.06
- 4. 0685K-05
- NOTE:
Delta CDF referenced to historical value.
TABLE 1 ATTACHMENT 1..
- DELTA CDF*
1.77E-06
- 1. 5iE-06 DELTA CDF*
+ 2.73E-07
+ 6.74E-07
SALEM LCR 91-06 NLR-N91095 REACTOR COOLANT SYSTEM 3/4.4.4 RELIEF VALVES As specified in GL 90-06 Action c. With both PORVs inoperable due to causes other than excessive seat leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore at least one PORV to OPERABLE status or close its associated block valve and remove power from the block valve and be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
As proposed by PSE&G The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> AOT will be extended to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SOURCE GL 90-06 PSE&G HISTORICAL SOURCE GL 90-06 PSE&G HISTORICAL SALEM UNIT 1 ACTION c EFFECT ON CDF AOT (HRS)
CDF 1, 72 5.7762E...;05 24, 168 5.7775E-05 704.5, 13140 5.7864E-05 SALEM UNIT 2 ACTION c EFFECT ON CDF AOT (HRS)
CDF 1, 72 4.0677E-05 24, 168 4.0681E-05 6.06, 13140 4.0695E-05
- NOTE:
Delta CDF referenced to historical value.
TABLE 2 ATTACHMENT 1 DELTA CDF*
1.02E-07 8.9E-08 DELTA CDF*
- 1. 80E-08
- 1. 40E-08
SALEM LCR 91-06 NLR-N91095 REACTOR COOLANT SYSTEM 3/4.4.4 RELIEF VALVES As specified in GL 90-06 Action d With one or both block valves inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restore the block valve(s) to OPERABLE status or place its associated PORV(s) in manual control.
Restore at least one block valve to OPERABLE status within the next hour if both block valves are inoperable; restore any remaining block valves to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUT DOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
As proposed by PSE&G SOURCE GL 90-06 PSE&G The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> AOT will be extended to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> AOT will be extended to 7 days.
SALEM UNIT 1 ACTION d** EFFECT ON CDF AOT (HRS)
CDF DELTA CDF*
1, 72 5.7358E-05 4.98E-07 24, 168 5.7680E-05
CDF DELTA CDF*
GL 90-06 1, 72 4.0902E-05
+ 2.17E-07 PSE&G 24, 168 4.1509E-05
+ 8.24E-07 HISTORICAL 36.06, 13140 4.0685E-05
- NOTE:
Delta CDF referenced to historical value.
Action d of GL 90-06 has been subdivided into actions d and e in the proposed technical specifications for clarity.
TABLE 3 ATTACHMENT 1
I SOURCE GL 90-06 PSE&G SALEM LCR 91-06 NLR-N91095 REACTOR COOLANT SYSTEM 3/4.4.4 RELIEF VALVES SALEM UNIT 1 ACTIONS b, c, AND d** OVERALL EFFECT ON CDF AOT (HRS)
CDF DELTA CDF*
1, 72 5.5509E-05 2.35E-06 24, 168 5.6112E-05 1.74E-06 HISTORICAL AS MODELED 5.7856E-05 SALEM UNIT 2 ACTIONS b, c, AND d** OVERALL EFFECT ON CDF SOURCE AOT (HRS)
CDF DELTA CDF*
GL 90-06 1, 72 4.1180E-05
+ 4.95E-07 PSE&G 24, 168 4.2192E-05
+ 1.51E-06 HISTORICAL AS MODELED 4.0685E-05
- NOTE:
- NOTE:
Delta CDF referenced to historical value.
Action d of GL 90-06 has been subdivided into actions d and e in the proposed technical specifications for clarity.
TABLE 4 ATTACHMENT 1
ATTACHMENT 2 REQUEST FOR LICENSE AMENDMENT INCORPORATION OF GENERIC LETTER 90-06 REQUIREMENTS SALEM GENERATING STATION UNITS 1 AND 2 FACILITY OPERATING LICENSE NOS. DPR-70 AND DPR-75 DOCKET NOS. 50-272 AND 50-311