LR-N16-0003, License Amendment Request to Amend the Accident Monitoring Instrumentation Technical Specifications: Difference between revisions

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{{#Wiki_filter:PSEG Nuclear LLC P.O.Box 236, Hancoclw Bridge, New Jersey 08038-0236 PSEG NudearLLC NOV 17 2016.
{{#Wiki_filter:PSEG Nuclear LLC P.O.Box 236, Hancoclw Bridge, New Jersey 08038-0236 PSEG NudearLLC NOV 17 2016.
10 CFR 50.90 LR-N16-0003 LAR S16-01 U . S . Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Salem Generating Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-70 and DPR-75 NRC Docket Nos. 50-272 and 50-311 S ubject:        Li cense Amendment Request to Amend the Accident Monitoring Instrumentation Technical Specifications I n accordance with the provisions of 10 CFR 50.90, PSEG N uclear LLC (PSEG) is submitting a request for an amendment to the Technical Specifications (TS) for Salem Generating Station (Salem) Units 1 and 2.
10 CFR 50.90 LR-N16-0003 LAR S16-01 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001  
The proposed amendment revises the TS 3/4.3.3.7, Accident Monitoring I nstrumentation.
 
Specifically, this change modifies the l ist of instruments required to be operable based on implementation of Regulatory Guide (RG) 1.97 for Salem, and revises the allowed outage times and Actions for inoperable channels to be consistent with NUREG-1431, Revision 4, "Standard Technical Specifications - Westinghouse Plants." provides an evaluation supporting the proposed changes. Attachment 2 provides the existing TS pages marked up to show the proposed changes. Attachment 3 provides existing TS Bases pages marked up to show the proposed changes and are being provided for information only.
==Subject:==
PSEG requests approval of this license amendment request (LAR) in accordance with standard NRC approval process and schedule. Once approved, the amendment wil l be implemented within 60 days from the date of issuance.
Salem Generating Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-70 and DPR-75 NRC Docket Nos. 50-272 and 50-311 License Amendment Request to Amend the Accident Monitoring Instrumentation Technical Specifications In accordance with the provisions of 10 CFR 50.90, PSEG Nuclear LLC (PSEG) is submitting a request for an amendment to the Technical Specifications (TS) for Salem Generating Station (Salem) Units 1 and 2.
I n accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated State of New Jersey Official.
The proposed amendment revises the TS 3/4.3.3.7, Accident Monitoring Instrumentation.
Specifically, this change modifies the list of instruments required to be operable based on implementation of Regulatory Guide (RG) 1.97 for Salem, and revises the allowed outage times and Actions for inoperable channels to be consistent with NUREG-1431, Revision 4, "Standard Technical Specifications - Westinghouse Plants." provides an evaluation supporting the proposed changes. Attachment 2 provides the existing TS pages marked up to show the proposed changes. Attachment 3 provides existing TS Bases pages marked up to show the proposed changes and are being provided for information only.
PSEG requests approval of this license amendment request (LAR) in accordance with standard NRC approval process and schedule. Once approved, the amendment will be implemented within 60 days from the date of issuance.
In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated State of New Jersey Official.
There are no regulatory commitments contained in this letter.
There are no regulatory commitments contained in this letter.
If you have any questions or require additional information, please contact Ms. Tanya Timberman at 856-339-1426.
If you have any questions or require additional information, please contact Ms. Tanya Timberman at 856-339-1426.  


LR-N16-0003                                                               10 CFR 50.90 Page 2 I declare under penalty of perjury that the foregoing is true and correct.
LR-N16-0003 Page 2 I declare under penalty of perjury that the foregoing is true and correct.
Executed on   __   t_t _/--.,.\ _7_/_1 :=--
Executed on __ t_t _/--.,.\\ _7_/_1 :=--6 (Date)
6 ____
(Date)
Respectfully, Charles V. McFeaters Site Vice President Salem Generating Station Attachments:
Respectfully, Charles V. McFeaters Site Vice President Salem Generating Station Attachments:
: 1. Evaluation of Proposed Changes
: 1. Evaluation of Proposed Changes
: 2. Mark-up of Proposed Technical Specification Pages 3 . Mark-up of Proposed Technical Specification Bases Pages cc:     Mr. D. Dorman, Administrator, Region I , NRC Ms. C. Parker, Project Manager, NRC NRC Senior Resident Inspector, Salem Mr. P. Mulligan, Chief, NJBNE PSEG Corporate Commitment Tracking Coordinator Salem Comm itment Tracking Coordinator
: 2. Mark-up of Proposed Technical Specification Pages
: 3. Mark-up of Proposed Technical Specification Bases Pages cc:
Mr. D. Dorman, Administrator, Region I, NRC Ms. C. Parker, Project Manager, NRC NRC Senior Resident Inspector, Salem Mr. P. Mulligan, Chief, NJBNE PSEG Corporate Commitment Tracking Coordinator Salem Commitment Tracking Coordinator 10 CFR 50.90
 
LR-N16-0003 LAR 516-01 Evaluation of Proposed Changes Table of Contents


LR-N16-0003                                                                                                                                                      LAR 516-01 Attachment 1 Evaluation of Proposed Changes Table of Contents
==1.0 DESCRIPTION==
................................................................................................................ 1  


==1.0   DESCRIPTION==
==2.0 PROPOSED CHANGE==
      .................................... . .................................... . . .....................................                                        1 2.0    PROPOS E D CHANGE            ..........................                 . .. . ........... . . ................. . ....... . ..... . ................. ........               1
.................................................................................................... 1  


==3.0   BACKGROUND==
==3.0 BACKGROUND==
        ............ . .................................................................. . . . . . . . . . . . .. ..................   .                         2 4.0   TECHN ICAL ANAL YSIS             ..............................................                             . ................................... . ...............           4 4.1 Methodology     ......................................... . . . .......................... . . . . . ........                         . .. . ........... . ......... 4 4.2 Technical Evaluation           ..... . ................................................... .......................................                                       5 4.3   Individual I nstrument Evaluations                           . . . ....................             . ......................................... . . . . . . . . . . . 5 4.4 LCO Times and Separate Condition Entry                                                 . . . . ........ . ............................... . . . . . ............         14 5.0    REGU LATORY ANALYSI S                . . . . . . .... . . . . . .. . . . . . .. . . . . ............ ........ . . .. . . .......... ....... . . ..................          16 5.1 No S ignificant Hazards Consideration                                        ....................................... . ........... . . . ......... . . . .                16 5 .2 Applicable Regulatory Requirements/Criteria                                                    . .......... ....... .............. . . . . . . . . . . . . ............. 18 6.0    ENVI RONMENTAL CONS I DERATION .                                      ............... . . . . ..................... .............................. . .                      19
............................................................................................................... 2 4.0 TECHNICAL ANAL YSIS  
.................................................................................................. 4 4.1 Methodology............................................................................................................ 4 4.2 Technical Evaluation  
................................................................................................ 5 4.3 Individual Instrument Evaluations............................................................................ 5 4.4 LCO Times and Separate Condition Entry............................................................. 14  


==7.0   REFERENCES==
==5.0 REGULATORY ANALYSIS==
        ................. . ............................................ ...... ................... . . . . ...................                                     19
........................................................................................... 16 5.1 No Significant Hazards Consideration................................................................... 16 5.2 Applicable Regulatory Requirements/Criteria
......................................................... 18


LR-N16-0003                                                                                  LAR S16-01
==6.0 ENVIRONMENTAL CONSIDERATION==
......................................................................... 19


==1.0         DESCRIPTION==
==7.0 REFERENCES==
.............................................................................................................. 19


The proposed amendment revises the TS 3/4. 3 . 3 . 7, Accident Monitoring I nstrumentation.
LR-N16-0003
Specifically, this change m odifies the list of instruments required to be operable based on implementation of Regu latory Guide (RG) 1 .97 for Salem, and revises the allowed outage times and Actions for inoperable channels to be consistent with N U REG-1 43 1 , Revision 4, "Standard Technical Specifications - Westinghouse Plants."
C hanges are proposed to Technical Specification (TS) Table 3.3-1 1 , to Actions 1 , 2, 4 and 6, and to TS Section 6.9.4. In addition, consistent with NUREG-1 431 a note is added to Section 3 . 3 . 3. 7 allowing separate condition entry for each function (i.e., affected instrument).


==2.0         PROPOSED CHANGE==
==1.0 DESCRIPTION==
LAR S16-01 The proposed amendment revises the TS 3/4.3.3.7, Accident Monitoring Instrumentation.
Specifically, this change modifies the list of instruments required to be operable based on implementation of Regulatory Guide (RG) 1.97 for Salem, and revises the allowed outage times and Actions for inoperable channels to be consistent with NUREG-1 431, Revision 4, "Standard Technical Specifications - Westinghouse Plants."
Changes are proposed to Technical Specification (TS) Table 3.3-1 1, to Actions 1, 2, 4 and 6, and to TS Section 6.9.4. In addition, consistent with NUREG-1 431 a note is added to Section 3.3.3. 7 allowing separate condition entry for each function (i.e., affected instrument).


This License Amendment Request revises Salem Units 1 and 2 TS Table 3 .3-1 1 and Table 4. 3-1 1 , Accident Monitoring Instrumentation as follows:
==2.0 PROPOSED CHANGE==
Instruments removed from TS Table 3 . 3-1 1 and TS Table 4.3-1 1 because they are not regulatory guide ( RG) 1 .97 Type A, or Category 1 :
This License Amendment Request revises Salem Units 1 and 2 TS Table 3.3-1 1 and Table 4.3-1 1, Accident Monitoring Instrumentation as follows:
* Reactor Coolant System Sub-Cooling Margin Monitor*
Instruments removed from TS Table 3.3-1 1 and TS Table 4.3-1 1 because they are not regulatory guide (RG) 1.97 Type A, or Category 1 :
* PORV Position Indicator (Unit 2 only)*
Reactor Coolant System Sub-Cooling Margin Monitor*
* PORV Block Valve Position I ndicator*
PORV Position Indicator (Unit 2 only)*
* Pressurizer Safety Valve Position Indicator*
PORV Block Valve Position Indicator*
* Containm ent Pressure (Narrow Range)
Pressurizer Safety Valve Position Indicator*
            *The associated footnotes are also removed.
Containment Pressure (Narrow Range)  
Instruments added to TS Table 3 .3-1 1 and TS Table 4. 3-1 1 because they are RG 1 .97 Type A, or Category 1 instruments that are not currently in the tables:
*The associated footnotes are also removed.
* Wide Range Neutron Flux Monitors
Instruments added to TS Table 3.3-1 1 and TS Table 4.3-1 1 because they are RG 1.97 Type A, or Category 1 instruments that are not currently in the tables:
* Auxiliary Feed Water (AFW) Storage Tank (Condensate Storage Tank) Water Level The following TS Table 3 . 3-1 1 Actions are revised:
Wide Range Neutron Flux Monitors Auxiliary Feed Water (AFW) Storage Tank (Condensate Storage Tank) Water Level The following TS Table 3.3-1 1 Actions are revised:
Action 1 With the num ber of O P E RABLE accident monitoring channels less than the Required Num ber of Channels shown in Table 3 . 3-1 1 , restore the inoperable channel(s) to O P ERABLE status within 30 days, or submit a special report in accordance with Specification 6.9.4.
Action 1 With the number of OPERABLE accident monitoring channels less than the Required Number of Channels shown in Table 3.3-1 1, restore the inoperable channel(s) to OPERABLE status within 30 days, or submit a special report in accordance with Specification 6.9.4.
Action 2 With the number of O P E RABLE accident monitoring channels less than the Minimum N u m ber o f Channels shown in Table 3. 3-1 1 , restore t he inoperable channel(s) to O P E RABLE status within 7 days or be in HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
Action 2 With the number of OPERABLE accident monitoring channels less than the Minimum Number of Channels shown in Table 3.3-1 1, restore the inoperable channel(s) to OPERABLE status within 7 days or be in HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
1
1  


LR-N1 6-0003                                                                                     LAR 51 6-01 Attachment 1 Action 4 With the number of OPERABLE channels one less than the Required N umber of channels shown in Table 3. 3-11, operation may proceed provided that an OPERABLE Steam Generator Wide Range level channel is available as an alternate means of indication for the Steam Generator with no OPERABLE Auxiliary Feedwater Flow Rate channel: OTHERWISE, restore the inoperable channel to OPERABLE status within 30 days, or submit a special report in accordance with Specification 6.9.4.
LR-N16-0003 Action 4 LAR 516-01 With the number of OPERABLE channels one less than the Required Number of channels shown in Table 3.3-11, operation may proceed provided that an OPERABLE Steam Generator Wide Range level channel is available as an alternate means of indication for the Steam Generator with no OPERABLE Auxiliary Feedwater Flow Rate channel: OTHERWISE, restore the inoperable channel to OPERABLE status within 30 days, or submit a special report in accordance with Specification 6.9.4.
Action 6 With the number of OPERABLE channels less than the Minimum Number of channels shown in Table 3.3-11, restore the inoperable channel(s) to OPERABLE status within 7 days, or be in HOT STANDBY within the next 6 hours and in HOT S HUTDOWN within the following 6 hours.
Action 6 With the number of OPERABLE channels less than the Minimum Number of channels shown in Table 3.3-11, restore the inoperable channel(s) to OPERABLE status within 7 days, or be in HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
TS 3.3.3.7 is revised to add a note which allows for a Separate Condition entry for each instrument.
TS 3.3.3.7 is revised to add a note which allows for a Separate Condition entry for each instrument.
The surveillance frequencies for Line Items 22 and 23 that are being added to TS Table 4.3-11 will be contained in the Surveillance Frequency Control Program with the exception of the Channel Functional Test which is not applicable for these instruments. Salem has implemented TSTF Change Traveler TSTF-425, Revision 3 ( Reference 11) as approved by the N RC with TS Amendments 299 and 282 (Reference 12} for Units 1 and 2 respectively. The initial surveillance frequency will be 31-days for the CHAN NEL CHECK and 18-months for the CHANNEL CAL I B RATION consistent with NUREG-1431, Standard Technical Specifications for Westinghouse Plants.
The surveillance frequencies for Line Items 22 and 23 that are being added to TS Table 4.3-11 will be contained in the Surveillance Frequency Control Program with the exception of the Channel Functional Test which is not applicable for these instruments. Salem has implemented TSTF Change Traveler TSTF-425, Revision 3 (Reference 11) as approved by the NRC with TS Amendments 299 and 282 (Reference 12} for Units 1 and 2 respectively. The initial surveillance frequency will be 31-days for the CHANNEL CHECK and 18-months for the CHANNEL CALIBRATION consistent with NUREG-1431, Standard Technical Specifications for Westinghouse Plants.
The criteria for relocation of a surveillance frequency to a licensee controlled program in accordance with TSTF-425 were reviewed. These surveil lance frequencies are periodic surveillances that: 1) do not reference other approved programs for the specified interval, 2) are not event driven, 3) do not have a time component based on event occurrence, and 4) are not related to a specific condition for performance. Therefore the periodic surveillance frequencies are within the scope of TSTF-425 for location in the licensee controlled Surveillance Frequency Control Program.
The criteria for relocation of a surveillance frequency to a licensee controlled program in accordance with TSTF-425 were reviewed. These surveillance frequencies are periodic surveillances that: 1) do not reference other approved programs for the specified interval, 2) are not event driven, 3) do not have a time component based on event occurrence, and 4) are not related to a specific condition for performance. Therefore the periodic surveillance frequencies are within the scope of TSTF-425 for location in the licensee controlled Surveillance Frequency Control Program.
Finally, TS 6.9.4, Special Reports, is revised to include Actions 1 and 4 of TS Table 3 .3-11.
Finally, TS 6.9.4, Special Reports, is revised to include Actions 1 and 4 of TS Table 3.3-11.  
 
==3.0      BACKGROUND==


I n August 2015, pursuant to 10 CFR 50.90, PSEG requested an amendment to the Salem facility operating license to remove the Pressurizer Power Operated Relief Valve (PORV) position indication from the Unit 1 Technical Specifications 3/4.3.3. 7, Accident Monitoring Instrumentation. The change was requested and approved on an emergency basis as permitted by 10 CFR 50.91(a)(5), and was the result of a 10 CFR Part 21 evaluation associated with the Namco limit switches. Per LAR S15-04, 1 due to the emergency nature of the request, the LAR was only for the specific line item for Salem Unit 1.
==3.0 BACKGROUND==
In August 2015, pursuant to 10 CFR 50.90, PSEG requested an amendment to the Salem facility operating license to remove the Pressurizer Power Operated Relief Valve (PORV) position indication from the Unit 1 Technical Specifications 3/4.3.3. 7, Accident Monitoring Instrumentation. The change was requested and approved on an emergency basis as permitted by 10 CFR 50.91(a)(5), and was the result of a 10 CFR Part 21 evaluation associated with the Namco limit switches. Per LAR S15-04, 1 due to the emergency nature of the request, the LAR was only for the specific line item for Salem Unit 1.
1 PSEG Emergency License Amendment Request to Remove Pressurizer Power Operated Relief Valve {PORV)
1 PSEG Emergency License Amendment Request to Remove Pressurizer Power Operated Relief Valve {PORV)
Position Indication Instrumentation from the Accident Monitoring Instrumentation Technical Specifications, dated August 31, 2015 (accession number ML15243A491).
Position Indication Instrumentation from the Accident Monitoring Instrumentation Technical Specifications, dated August 31, 2015 (accession number ML15243A491).
2
2  


LR-N16-00 0 3                                                                                          LAR 816-01 PSEG is submitting this LAR to revise Unit 1 and Unit 2 TS Tables 3 . 3-11 and 4.3-1 1, Accident Monitoring Instrumentation. The purpose is to align the remainder of the TS tables with the requirements of Regulatory Guide ( RG) 1 .97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Fol lowing an Accident",
LR-N16-0003 LAR 816-01 PSEG is submitting this LAR to revise Unit 1 and Unit 2 TS Tables 3.3-11 and 4.3-1 1, Accident Monitoring Instrumentation. The purpose is to align the remainder of the TS tables with the requirements of Regulatory Guide (RG) 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident",
Revision 22 dated December 1980 and the N UREG 1431, Revision 4, Standard Technical Specifications (STS).
Revision 22 dated December 1980 and the NUREG 1431, Revision 4, Standard Technical Specifications (STS).
N U REG-0737, "Clarification of TMI Action Plan Requirements" (Reference 1 3), included requirements for accident m onitoring instrumentation to be listed in plant Technical Specifications. Subsequently, the N RC Final Policy Statement on Technical Specifications I m provements for N uclear Power Reactors, which was later codified by changes to 10 CFR 5 0 . 36, provid ed a specific set of objective criteria as guidance for determining which regulatory requirements and operating restrictions should be included In Technical Specifications:
NUREG-0737, "Clarification of TMI Action Plan Requirements" (Reference 1 3), included requirements for accident monitoring instrumentation to be listed in plant Technical Specifications. Subsequently, the NRC Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, which was later codified by changes to 10 CFR 50.36, provided a specific set of objective criteria as guidance for determining which regulatory requirements and operating restrictions should be included In Technical Specifications:
(A) Criterion 1. Installed instru mentation that is used to detect, and ind icate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
(A) Criterion 1. Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
( B) Criterion 2. A process variable, d esign feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission prod uct barrier.
(B) Criterion 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
(C) Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to m itigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
(C) Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
(D) Criterion 4. A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.
(D) Criterion 4. A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.
RG 1 .97 describes a method acceptable to the NRC for complying with the regulations to provide instrum entation to monitor plant variables and systems during and following an accident in a light-water-cooled nuclear power plant. Variables and Categories are defined as follows:
RG 1.97 describes a method acceptable to the NRC for complying with the regulations to provide instrumentation to monitor plant variables and systems during and following an accident in a light-water-cooled nuclear power plant. Variables and Categories are defined as follows:
Type A Variables are those variables to be monitored that provide the primary information required to permit the control room operator to take specific manually controlled actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for design basis accident events.
Type A Variables are those variables to be monitored that provide the primary information required to permit the control room operator to take specific manually controlled actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for design basis accident events.
Type B Variables are those variables that provide information to ind icate whether plant safety functions are being accomplished . Plant safety functions are (1) reactivity control, (2) core cooling, (3) maintaining reactor coolant system integrity, and (4) maintaining containment integrity (includ ing radioactive effl uent control).
Type B Variables are those variables that provide information to indicate whether plant safety functions are being accomplished. Plant safety functions are (1) reactivity control, (2) core cooling, (3) maintaining reactor coolant system integrity, and (4) maintaining containment integrity (including radioactive effluent control).
2 Salem was designed prior t o the issuance o f R G 1.97. The UFSAR was subsequently updated t o demonstrate compliance with the intent of RG 1.97, Revision 2. In various site-specific evaluations of PAM instruments, Revision 3 was used by Salem as more clearly presenting guidance for PWR variables while at the same time being essentially equivalent to Revision 2 guidance. The current RG 1.97 (Revision 4) states that it is primarily intended for new reactors and that previous versions of the RG remain in effect for licensees of current operating reactors.
2 Salem was designed prior to the issuance of RG 1.97. The UFSAR was subsequently updated to demonstrate compliance with the intent of RG 1.97, Revision 2. In various site-specific evaluations of PAM instruments, Revision 3 was used by Salem as more clearly presenting guidance for PWR variables while at the same time being essentially equivalent to Revision 2 guidance. The current RG 1.97 (Revision 4) states that it is primarily intended for new reactors and that previous versions of the RG remain in effect for licensees of current operating reactors.
Revision 2 remains the Salem licensing basis for RG 1.97 compliance.
Revision 2 remains the Salem licensing basis for RG 1.97 compliance.
3
3  


LR-N16-0003                                                                                 LAR 516-01 Attachment 1 Type C Variables are those variables that provide information to indicate the potential for being breached or the actual breach of the barriers to fission product releases. The barriers are (1) fuel cladding, (2) primary coolant pressure boundary, and (3) containment.
LR-N16-0003 LAR 516-01 Type C Variables are those variables that provide information to indicate the potential for being breached or the actual breach of the barriers to fission product releases. The barriers are (1) fuel cladding, (2) primary coolant pressure boundary, and (3) containment.
Type D Variables are those variables that provide information to indicate the operation of individual safety systems and other systems important to safety. These variables are to help the operator make appropriate decisions in using the individual systems important to safety in mitigating the consequences of an accident.
Type D Variables are those variables that provide information to indicate the operation of individual safety systems and other systems important to safety. These variables are to help the operator make appropriate decisions in using the individual systems important to safety in mitigating the consequences of an accident.
Type E Variables are those variables to be monitored as required for use in determining the magnitude of the release of radioactive materials and continually assessing such releases.
Type E Variables are those variables to be monitored as required for use in determining the magnitude of the release of radioactive materials and continually assessing such releases.
Category 1, i n general, provides for full qualification, redundancy, and continuous real time display and requires onsite (standby) power.
Category 1, in general, provides for full qualification, redundancy, and continuous real time display and requires onsite (standby) power.
Category 2, i n general, provides for qualification but is less stringent in that it does not (of itself) i nclude seismic qualification, redundancy, or continuous display and requires only a high-reliability power source (not necessarily standby power).
Category 2, in general, provides for qualification but is less stringent in that it does not (of itself) include seismic qualification, redundancy, or continuous display and requires only a high-reliability power source (not necessarily standby power).
Category 3 is the least stringent. It provides for high-quality commercial-grade equipment that requires only offsite power.
Category 3 is the least stringent. It provides for high-quality commercial-grade equipment that requires only offsite power.
The current TS contains instrumentation that does not meet any of the 10 CFR 50.36(c)(2)(ii) screening criteria for i nclusion in TS. It also contains TS allowed outage times (7 days for less than required number of OPERABLE channels, 48 hours for less than minimum number of OPERABLE channels) that are overly restrictive and are not commensurate with the safety significance of the i nstruments.
The current TS contains instrumentation that does not meet any of the 10 CFR 50.36(c)(2)(ii) screening criteria for inclusion in TS. It also contains TS allowed outage times (7 days for less than required number of OPERABLE channels, 48 hours for less than minimum number of OPERABLE channels) that are overly restrictive and are not commensurate with the safety significance of the instruments.  


==4.0     TECHNICAL ANALYSIS==
==4.0 TECHNICAL ANALYSIS==
4.1 Methodology The accident monitoring instrumentation for the Salem Unit 1 and Unit 2 TS Tables 3.3-11 and 4.3-11 was assessed as follows:
The following documents were reviewed and compared with each other, specifically to identify post-accident monitoring (PAM) instruments as Type A or Category 1:
Document Reference Regulatory Guide 1.97 1
Salem UFSAR 2
Salem TS Tables 3.3-11/4.3-11 3
Standard Westinghouse TS 4
Engineering Evaluation 5
S-C-GSC-CE E-04 70 4


4.1     Methodology The accident monitoring i nstrumentation for the Salem Unit 1 and Unit 2 TS Tables 3.3-11 and 4.3-11 was assessed as follows:
LR-N16-0003 LAR 516-01 Based on the comparisons, instruments that should be removed because they are not Type A variable or not Category 1 were identified.
* The following documents were reviewed and compared with each other, specifically to identify post-accident monitoring (PAM) instruments as Type A or Category 1:
Based on the comparisons, instruments that should be added to the TS because they are Type A variables, or non-Type A, but Category 1 were identified.
Document                              Reference Regulatory Guide 1.97                     1 Salem UFSAR                                2 Salem TS Tables 3.3-11/4.3-11              3 Standard Westinghouse TS                  4 Engineering Evaluation                     5 S-C-GSC-CE E-04 70 4
Type A Variables are defined as those variables to be monitored that provide the primary information required to permit the control room operator to take specific manually controlled actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for design basis accident events.
Category 1, in general, provides for full q ualification, redundancy, and continuous real-time display and requires onsite (standby) power.
The intent is to align the instruments in the Salem Unit 1 and U nit 2 TS Tables 3.3-1 1 and 4.3-1 1 to be consistent with the scope of the N UREG-1 431 reviewer's note which states that the TS should include all RG 1.97 Type A instruments and all RG 1.97, Category 1, non-Type A instruments in accordance with the Unit's RG 1.97 Safety Evaluation Report. The instruments proposed for removal from the TS tables are not being removed from the plant and will continue to satisfy their RG 1.97 requirements.
4.2 Technical Evaluation Each of the instruments listed in Table 1 below, were evaluated and determined to be removed or added to the Salem Unit 1 and U nit 2 TS Tables 3.3-1 1 and 4.3-1 1.
Table 1 Instrument Reactor Coolant System Sub-Cooling Margin Monitor PORV Position Indicator PORV Block Valve Position Indicator Pressurizer Safety Valve Position I ndicator Containment Pressure (Narrow Range)
Wide Range Neutron Flux Monitors AFW (Condensate) Storage Tank Water Level 4.3 Individual I nstrument Evaluations Reactor Coolant System Sub-Cooling Margin Monitor Salem Unit Action U 1 and U2 Remove Unit 2 only:;
Remove U 1 and U2 Remove U 1 and U2 Remove U 1 and U2 Remove U 1 and U2 Add U1 and U2 Add The reactor coolant system (RCS) sub-cooling margin monitor (SMM ) is not listed in the STS. It should be noted, that this instrument is categorized in Table 3 of RG 1.97 as a Category 2 variable.
For the Salem Generating Station, Eng ineering Evaluation S-C-GCS-CEE-0470 states that the sub-cooling margi n monitor is a Type B Category 2 variable in accordance with the requirements of RG 1.97. It is currently listed in U FSAR Table 7.5-3, as a Type A variable, Category 1. As discussed below, degrees of subcooling should be removed from U FSAR Table 7.5-3 because it does not meet the Type A or Category 1 criteria and should be classified as Type B, Category 2.
3 The PORV position indicator was removed from Salem Unit 1 by TS Amendment 310 5


LR-N16-0003                                                                                         LAR 516-01
LR-N16-0003 LAR 516-01 Type B Variables are defined as those variables that provide information to indicate whether plant safety functions are being accomplished. Plant safety functions are (1) reactivity control, (2) core cooling, (3) maintaining reactor coolant system integrity, and
* Based on the comparisons, instruments that s h o u ld be removed because they are not Type A variable or not Category 1 were identified .
( 4) maintaining containment integrity (including radioactive effluent control).
* Based on the com parisons, i n struments that shou ld be added to the TS because they are Type A variables, or non-Type A, but Category 1 were identified .
Type A Variables are defi ned as those variables to be m o nitored that provide the primary i nformation req u ired to permit the control room operator to take specific manually controlled actio n s for which no automatic control is provided and that are req uired for safety systems to accomplish their safety fu n ctions for desi g n basis accident events .
Category 1 , in general, provides for full q ualificati o n , red u ndancy, and conti nuous real-time d i splay and req uires o n s ite (standby) power.
The i ntent is to align the i n stru ments in the Salem U n it 1 and U n it 2 TS Tables 3 . 3 - 1 1 and 4 . 3 - 1 1 to be consistent with the scope of the N UREG - 1 431 reviewer's note which states that the TS should i n cl ude all RG 1 . 97 Type A instruments and all RG 1 . 97, Category 1 , n o n-Type A instruments i n accordance with the U n it's RG 1 . 97 Safety Evaluation Report. The i n struments proposed for removal from the TS tables are n ot bei n g removed from the plant and will conti nue to satisfy their RG 1 . 97 req uirements.
4.2          Technical Eval uatio n Each o f the i n struments listed i n Table 1 below, were evaluated a n d determ i ned t o b e removed or added to the Salem U n it 1 and U n it 2 TS Tables 3 . 3- 1 1 and 4.3-1 1 .
Table 1 Instrument                                                              Salem Unit        Action Reactor Coolant System Sub-Cooling Marg i n M o nitor                    U 1 and U 2      Remove PORV Position Indicator                                                  U n it 2 o n ly  Remove PORV Block Valve Position I n d icator                                    U 1 and U 2      Remove Pressurizer Safety Valve Position I ndicator                              U 1 and U 2      Remove Contai n m ent Pressure (Narrow Range)                                    U 1 and U 2      Remove Wide Range Neutron Flux Mon itors                                        U 1 and U 2      Add AFW (Condensate) Storage Tank Water Level                                U1 and U2        Add 4.3          I n d ivid ual I nstrument Evaluatio n s Reactor Coolant System S ub-Coo l i ng Margin M o n itor The reactor coolant system ( RCS) s ub-cooling marg i n m o n itor (SM M ) is not listed i n the STS . It sho u l d be noted , that this i n strument is categorized i n Table 3 of RG 1 . 97 as a Category 2 variable.
For the Salem Generati ng Statio n , Eng i neeri ng Eval uatio n S-C-GCS-CEE-0470 states that the s ub-co o l i n g marg i n monitor is a Type B Category 2 variable i n accordance with the req u irements of RG 1 .97. It is currently l isted i n U FSAR Table 7.5-3, as a Type A variable, Categ ory 1 . As d iscussed below, degrees of subcooling should be removed from U FSAR Table
: 7. 5-3 because it does not meet the Type A or Category 1 criteria and should be classified as Type B , Category 2.
3 The    PORV    position indicator was removed from Salem Unit 1 by TS Amendment 310 5
 
LR-N16-0003                                                                                LAR 516-01 Attachment 1 Type B Variables are defined as those variables that provide information to indicate whether plant safety functions are being accomplished. Plant safety functions are (1 )
reactivity control, (2) core cooling, (3) maintaining reactor coolant system integrity, and
( 4) maintain ing containment integrity (including radioactive effluent control).
Category 2, in general, provides for qualification but is less stringent in that it does not (of itself) include seismic qualification, redundancy, or continuous display and requires only a high-reliability power source (not necessarily standby power).
Category 2, in general, provides for qualification but is less stringent in that it does not (of itself) include seismic qualification, redundancy, or continuous display and requires only a high-reliability power source (not necessarily standby power).
The RCS sub-cooling margin indication provides information to the operators related to satisfying one of the Safety I njection (SI) termination criteria following a steam line break or Steam Generator Tube Rupture (SGTR) accident. The inputs to the RCS sub-cooling margin monitor are the core exit thermocouples for RCS temperature and the wide range RCS pressure indication for RCS pressure.
The RCS sub-cooling margin indication provides information to the operators related to satisfying one of the Safety Injection (SI) termination criteria following a steam line break or Steam Generator Tube Rupture (SGTR) accident. The inputs to the RCS sub-cooling margin monitor are the core exit thermocouples for RCS temperature and the wide range RCS pressure indication for RCS pressure.
The RCS SMM does not detect or indicate a significant abnormal degradation of the reactor coolant pressure boundary, as required by Criterion 1 . This is consistent with the NRC Final Policy Statement, which provided that Criterion 1 is intended to ensure that those instruments specifically installed to detect excessive reactor coolant system leakage be i ncluded in the TS.
The RCS SMM does not detect or indicate a significant abnormal degradation of the reactor coolant pressure boundary, as required by Criterion 1. This is consistent with the NRC Final Policy Statement, which provided that Criterion 1 is intended to ensure that those instruments specifically installed to detect excessive reactor coolant system leakage be included in the TS.
The instrumentation that satisfies Criterion 1 is contained in Salem Unit 1 TS 3.4.6. 1 and Unit 2 TS 3.4.7.1 "Reactor Coolant System Leakage, Leakage Detection Systems."
The instrumentation that satisfies Criterion 1 is contained in Salem Unit 1 TS 3.4.6.1 and Unit 2 TS 3.4. 7.1 "Reactor Coolant System Leakage, Leakage Detection Systems."
RCS sub-cooling instrumentation is not a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis considered in Criterion 2.
RCS sub-cooling instrumentation is not a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis considered in Criterion 2.
The RCS S M M is not part of a primary success path as indicated in Criterion 3.
The RCS SMM is not part of a primary success path as indicated in Criterion 3.
The loss of the RCS SMM instrumentation has no effect on the probabilistic safety assessment, and has not been shown to be significant to health and safety as considered i n Criterion 4.
The loss of the RCS SMM instrumentation has no effect on the probabilistic safety assessment, and has not been shown to be significant to health and safety as considered in Criterion 4.
The RCS sub-cooling margin indication provides information to indicate whether the core cooling safety function is being accomplished, and is therefore considered a Type B variable.
The RCS sub-cooling margin indication provides information to indicate whether the core cooling safety function is being accomplished, and is therefore considered a Type B variable.
The RCS sub-cooling indication is a backup to the core exit thermocouples and RCS pressure, and is therefore considered Category 2. I n support of this LAR, the UFSAR (Tables 7.5-1 , 7.5-3, and 7.5-4) and associated design documentation, as appropriate, are being revised to reflect that the subcooli ng margin monitor is not a Type A variable. PSEG has determined these documentation changes can be implemented in accordance with 1 0 CFR 50.59. The Salem UFSAR will be revised prior to implementation of this proposed TS change.
The RCS sub-cooling indication is a backup to the core exit thermocouples and RCS pressure, and is therefore considered Category 2. In support of this LAR, the UFSAR (Tables 7.5-1, 7.5-3, and 7.5-4) and associated design documentation, as appropriate, are being revised to reflect that the subcooling margin monitor is not a Type A variable. PSEG has determined these documentation changes can be implemented in accordance with 1 0 CFR 50.59. The Salem UFSAR will be revised prior to implementation of this proposed TS change.
The RCS sub-cooling margin indication does not meet any of the four screening criteria of the NRC Final Policy Statement in 58 Federal Reg ister 391 32 (58FR391 32), which provides criteria to determine which items are required to be included in the TS as LCOs. This conclusion is supported by the absence of operability and surveillance requirements for the sub-cooling margin ind ication in the improved standard TS presented in NUREG-1 431. Accordingly, removal of the RCS sub-cooling margin indication from the TS conforms to the STS.
The RCS sub-cooling margin indication does not meet any of the four screening criteria of the NRC Final Policy Statement in 58 Federal Register 391 32 (58FR391 32), which provides criteria to determine which items are required to be included in the TS as LCOs. This conclusion is supported by the absence of operability and surveillance requirements for the sub-cooling margin indication in the improved standard TS presented in NUREG-1 431. Accordingly, removal of the RCS sub-cooling margin indication from the TS conforms to the STS.
Therefore, the reactor coolant system sub-cooling marg in monitor is removed from the Salem Unit 1 and Unit 2 TS Tables 3.3- 1 1 and 4.3-11 .
Therefore, the reactor coolant system sub-cooling margin monitor is removed from the Salem Unit 1 and Unit 2 TS Tables 3.3-1 1 and 4.3-11.
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LR-N16-0003                                                                             LAR 516-01 PORV Position Indication The reactor coolant system is protected against over-pressurization by control and protective ci rcuits such as the pressurizer pressure high reactor trip and by the PORVs connected to the top of the pressurizer. The PORVs provide a means for pressure relief. Each PORV is equipped with two limit switches to provide Open and Closed indication (i.e. lights) in the control room.
LR-N16-0003 PORV Position Indication LAR 516-01 The reactor coolant system is protected against over-pressurization by control and protective circuits such as the pressurizer pressure high reactor trip and by the PORVs connected to the top of the pressurizer. The PORVs provide a means for pressure relief. Each PORV is equipped with two limit switches to provide Open and Closed indication (i.e. lights) in the control room.
The PORV limit switch position i ndicators provide information to the control room operators related to the position of the pressurizer PORV's. The Desig n Basis Accident (DBA) analysis of an inadvertent opening of the PORV does not rely on operator diagnosis and closure of the PORV or block valve, the D BA analysis assum es that automatic safety injection actuation will provide adequate protection . At Salem, this event is bounded by the more limiting inadvertent opening of a code safety valve as described i n Salem UFSAR Section 1 5.2. 12.1 , Accidental Depressurization of the Reactor Coolant System . At Salem, no credit for diagnosis and re closure of a PORV, block valve, or code valve is assum ed in the DBA analysis.
The PORV limit switch position indicators provide information to the control room operators related to the position of the pressurizer PORV's. The Design Basis Accident (DBA) analysis of an inadvertent opening of the PORV does not rely on operator diagnosis and closure of the PORV or block valve, the DBA analysis assumes that automatic safety injection actuation will provide adequate protection. At Salem, this event is bounded by the more limiting inadvertent opening of a code safety valve as described in Salem UFSAR Section 15.2.12.1, Accidental Depressurization of the Reactor Coolant System. At Salem, no credit for diagnosis and re closure of a PORV, block valve, or code valve is assumed in the DBA analysis.
The Salem UFSAR identifies "Primary System Safety Relief Valve Positions (including PORV and code valves)" as Type D, Category 2 Variables, consistent with RG 1.97 (UFSAR Section 7.5). Type D variables are those variables that provide information to indicate the operation of individual safety system s and other systems important to safety. Category 2 instruments are desig ned to less stringent qualifications that do not require seismic qualification, redundancy, or continuous d isplay, and require only a hig h reliability power source, not necessarily standby power. Removing the PORV position indication from the TS conforms with the NRC position on application of the screening criteria to post-accident monitoring instrumentation.
The Salem UFSAR identifies "Primary System Safety Relief Valve Positions (including PORV and code valves)" as Type D, Category 2 Variables, consistent with RG 1.97 (UFSAR Section 7.5). Type D variables are those variables that provide information to indicate the operation of individual safety systems and other systems important to safety. Category 2 instruments are designed to less stringent qualifications that do not require seismic qualification, redundancy, or continuous display, and require only a high reliability power source, not necessarily standby power. Removing the PORV position indication from the TS conforms with the NRC position on application of the screening criteria to post-accident monitoring instrumentation.
The PORVs themselves are part of the primary success path in the UFSAR accident analysis because they are assumed to actuate to m itigate a DBA and therefore meet Criterion 3 of the NRC Final Policy Statement. For example, they are credited in Salem UFSAR Section 1 5.2. 1 4, Spurious Operation of the Safety Injection System at Power. The operability of the PORVs is therefore required by TS 3.4.3, " Relief Valves." However, PORV position indication does not detect or indicate a sig nificant abnormal degradation of the reactor coolant pressure boundary, as required by Criterion 1 . P O RV position indication is not a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis considered in Criterion 2.
The PORVs themselves are part of the primary success path in the UFSAR accident analysis because they are assumed to actuate to mitigate a DBA and therefore meet Criterion 3 of the NRC Final Policy Statement. For example, they are credited in Salem UFSAR Section 15.2. 1 4, Spurious Operation of the Safety Injection System at Power. The operability of the PORVs is therefore required by TS 3.4.3, "Relief Valves." However, PORV position indication does not detect or indicate a significant abnormal degradation of the reactor coolant pressure boundary, as required by Criterion 1. PORV position indication is not a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis considered in Criterion 2.
While the function of the PORVs themselves is part of the primary success path in the UFSAR, PORV position indication is not part of the primary success path. UFSAR accident analysis assumes that the PORVs open as desig ned to reduce reactor pressure and no operator action based on PORV position ind ication is required. Therefore, PORV position indication is not part of the primary success path as i nd icated in Criterion 3.
While the function of the PORVs themselves is part of the primary success path in the UFSAR, PORV position indication is not part of the primary success path. UFSAR accident analysis assumes that the PORVs open as designed to reduce reactor pressure and no operator action based on PORV position indication is required. Therefore, PORV position indication is not part of the primary success path as indicated in Criterion 3.
The loss of PORV position indication instrum entation has no effect on the probabilistic safety assessment, and has not been shown to be significant to health and safety as considered in Criterion 4.
The loss of PORV position indication instrumentation has no effect on the probabilistic safety assessment, and has not been shown to be significant to health and safety as considered in Criterion 4.
PORV position indication instrumentation does not meet any of the four screening criteria of the N RC Final Policy Statement in 58 Federal Register 391 32 (58FR39132), which provides criteria to d etermine which items are required to be included in the TS as LCOs. This conclusion is supported by the absence of operability and surveillance requirements for the PORV position indication instrumentation in the improved standard TS presented in NUREG-1 431 .
PORV position indication instrumentation does not meet any of the four screening criteria of the NRC Final Policy Statement in 58 Federal Register 391 32 (58FR39132), which provides criteria to determine which items are required to be included in the TS as LCOs. This conclusion is supported by the absence of operability and surveillance requirements for the PORV position indication instrumentation in the improved standard TS presented in NUREG-1 431.
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LR-N16-000 3                                                                              LAR 516-01 Accordingly, removal of PORVs from TS conforms to the STS. The PORV position indicator was previously removed from Salem Unit 1 through TS Amendment 31 0 .
LR-N16-0003 LAR 516-01 Accordingly, removal of PORVs from TS conforms to the STS. The PORV position indicator was previously removed from Salem Unit 1 through TS Amendment 31 0.
Therefore, the PORV position indication is removed from the Salem Unit 2 T S Tables 3.3-1 1 and 4. 3-11 .
Therefore, the PORV position indication is removed from the Salem Unit 2 TS Tables 3.3-1 1 and 4.3-11.
PORV B lock Valve Position Indication The PORVs can be isolated by the PORV block valves which are connected in line with the PORVs. The PORV block valves provide closure redundancy to the PORVs, to isolate PORVs with excessive seat leakage or that stick open. Use of the PORV block valves in the event of a PORV malfunction can prevent a severe depressurization of the RCS with potential for uncovering of the reactor core.
PORV Block Valve Position Indication The PORVs can be isolated by the PORV block valves which are connected in line with the PORVs. The PORV block valves provide closure redundancy to the PORVs, to isolate PORVs with excessive seat leakage or that stick open. Use of the PORV block valves in the event of a PORV malfunction can prevent a severe depressurization of the RCS with potential for uncovering of the reactor core.
The PORV block valve limit switch position indication provides information to the control room operators on the position of the pressurizer PORV block valves. It could be used to d iagnose the availability of the pressurizer PORV's for use in depressurizing the RCS or to indicate the isolation of a stuck open PORV ( LOCA) at lower RCS temperatures. The PORV block valve l i m it switch position indication does not provide an indication for operator actions for which no automatic control is provided or impact the response of the PORVs to a design basis accident.
The PORV block valve limit switch position indication provides information to the control room operators on the position of the pressurizer PORV block valves. It could be used to diagnose the availability of the pressurizer PORV's for use in depressurizing the RCS or to indicate the isolation of a stuck open PORV (LOCA) at lower RCS temperatures. The PORV block valve limit switch position indication does not provide an indication for operator actions for which no automatic control is provided or impact the response of the PORVs to a design basis accident.
Furthermore, this instrumentation is not needed for m anual operator action necessary for safety system s to accomplish their safety function for the design-basis events.
Furthermore, this instrumentation is not needed for manual operator action necessary for safety systems to accomplish their safety function for the design-basis events.
The PORVs themselves are part of the primary success path in the U FSAR accident analysis because they are assumed to actuate to mitigate a Design Basis Accident ( DBA) and therefore m eet Criterion 3 of the NRC Final Policy Statement. In order to provide this function, the PORV block valves m ust be open or capable of being manually opened. The operability of the PORV block valves is therefore required by Unit 1 TS 3. 4.3 and Unit 2 TS 3.4. 5, "Relief Valves."
The PORVs themselves are part of the primary success path in the U FSAR accident analysis because they are assumed to actuate to mitigate a Design Basis Accident (DBA) and therefore meet Criterion 3 of the NRC Final Policy Statement. In order to provide this function, the PORV block valves m ust be open or capable of being manually opened. The operability of the PORV block valves is therefore required by Unit 1 TS 3.4.3 and Unit 2 TS 3.4.5, "Relief Valves."
However, PORV block valve position indication does not detect or indicate a significant abnormal degradation of the reactor coolant pressure boundary, as required by Criterion 1 . This is consistent with the NRC Final Policy Statement, which provided that Criterion 1 is intended to ensure that those instruments specifically installed to detect excessive reactor coolant system leakage be included in the TS. Criterion 1 is not to be interpreted to include instru mentation installed to identify the source of actual leakage, for example valve position indicators. The instrumentation that satisfies Criterion 1 is contained in Salem Unit 1 TS 3.4.6 . 1 and Unit 2 TS 3.4. 7. 1 "Reactor Coolant System Leakage, Leakage Detection Systems. "
However, PORV block valve position indication does not detect or indicate a significant abnormal degradation of the reactor coolant pressure boundary, as required by Criterion 1. This is consistent with the NRC Final Policy Statement, which provided that Criterion 1 is intended to ensure that those instruments specifically installed to detect excessive reactor coolant system leakage be included in the TS. Criterion 1 is not to be interpreted to include instrumentation installed to identify the source of actual leakage, for example valve position indicators. The instrumentation that satisfies Criterion 1 is contained in Salem Unit 1 TS 3.4.6.1 and Unit 2 TS 3.4. 7.1 "Reactor Coolant System Leakage, Leakage Detection Systems."
PORV block valve position indication is not a process variable, design feature, or operating restriction that is an initial condition of a D BA or transient analysis considered in Criterion 2.
PORV block valve position indication is not a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis considered in Criterion 2.
While the function of the PORVs themselves is part of the primary success path in the U FSAR, the PORV block valve position indication is not part of the primary success path. U FSAR accident analysis assu mes that the PORVs open as designed to reduce reactor pressure. In the event a PORV block valve is closed to isolate PORV valve seat leakage as allowed by the technical specifications, the emergency operating proced ures direct the operators to manually open the PORV and associated block valve. This action does not rely on the PORV block valve position indication. Therefore, PORV block valve position indication is not part of the primary success path as indicated in Criterion 3.
While the function of the PORVs themselves is part of the primary success path in the UFSAR, the PORV block valve position indication is not part of the primary success path. U FSAR accident analysis assumes that the PORVs open as designed to reduce reactor pressure. In the event a PORV block valve is closed to isolate PORV valve seat leakage as allowed by the technical specifications, the emergency operating procedures direct the operators to manually open the PORV and associated block valve. This action does not rely on the PORV block valve position indication. Therefore, PORV block valve position indication is not part of the primary success path as indicated in Criterion 3.
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LR-N16-0 003                                                                                        LAR 516-0 1 Attachment 1 The loss of PORV block valve position indication instrumentation has no effect on the probabilistic safety assessment, and has not been shown to be significant to health and safety as considered i n Criterion 4.
LR-N16-0003 LAR 516-0 1 The loss of PORV block valve position indication instrumentation has no effect on the probabilistic safety assessment, and has not been shown to be significant to health and safety as considered in Criterion 4.
PORV block valve position indication does not meet any of the four screening criteria of the NRC Final Policy Statement in 58 Federal Reg ister 391 32 (58FR391 32), which provides criteria to determine which items are requi red to be included in the TS as LCOs. This conclusion is supported by the absence of operability and surveillance requirements for the PORV block valve position indication in the improved standard TS presented in NU REG-1 431 . Accordingly, removal of the PORV block valve position indication from the TS conforms to the STS.
PORV block valve position indication does not meet any of the four screening criteria of the NRC Final Policy Statement in 58 Federal Register 391 32 (58FR391 32), which provides criteria to determine which items are required to be included in the TS as LCOs. This conclusion is supported by the absence of operability and surveillance requirements for the PORV block valve position indication in the improved standard TS presented in NU REG-1 431. Accordingly, removal of the PORV block valve position indication from the TS conforms to the STS.
Thus, since the PORV block valve limit switch positions provide information to indicate the status of the pressurizer PORV block valves which are used to isolate the PORVs in the event of excessive PORV leakage, they are a Type D4 variable. For the same reasons as the PORV position indications, the block valves position indicators do not meet Criterion 1 , 2, 3 or 4 of the NRC Final Policy Statement and an LCO is not required concerning the PORV block valve position indicators. Because the PORV block valve position indicator is not classified as a Type A variable, or a non-Type A but Category 1 variable, the PORV block valve position indication is not required to be in the Salem U nit 1 and Unit 2 TS Tables 3.3-1 1 and 4.3-1 1 .
Thus, since the PORV block valve limit switch positions provide information to indicate the status of the pressurizer PORV block valves which are used to isolate the PORVs in the event of excessive PORV leakage, they are a Type D4 variable. For the same reasons as the PORV position indications, the block valves position indicators do not meet Criterion 1, 2, 3 or 4 of the NRC Final Policy Statement and an LCO is not required concerning the PORV block valve position indicators. Because the PORV block valve position indicator is not classified as a Type A variable, or a non-Type A but Category 1 variable, the PORV block valve position indication is not required to be in the Salem Unit 1 and Unit 2 TS Tables 3.3-1 1 and 4.3-1 1.
Therefore, the PORV block valve position indication is removed from the Salem Unit 1 and Unit 2 TS Tables 3.3-1 1 and 4.3-1 1 .
Therefore, the PORV block valve position indication is removed from the Salem Unit 1 and Unit 2 TS Tables 3.3-1 1 and 4.3-1 1.
Pressurizer Safety Valve Position I ndication The Pressurizer Safety Valve Position I ndication provides information to the control room operators on the position of the pressurizer safety valves. It could be used to d iagnose high RCS pressures or a stuck open safety valve (i.e. LOCA) at lower RCS pressures. They are both considered Type D variable, Category 2.
Pressurizer Safety Valve Position Indication The Pressurizer Safety Valve Position Indication provides information to the control room operators on the position of the pressurizer safety valves. It could be used to diagnose high RCS pressures or a stuck open safety valve (i.e. LOCA) at lower RCS pressures. They are both considered Type D variable, Category 2.
The Pressurizer safety relief valves themselves are part of the primary success path in the UFSAR accident analysis because they are assumed to actuate to m itigate a Design Basis Accident (DBA) and therefore meet Criterion 3 of the NRC Final Policy Statement. The operabil ity of the Pressurizer safety valves is therefore required by U nit 1 TS 3.4.2 and Unit 2 TS 3.4.2 & 3.4.3, "Safety Valves." However, the Pressurizer safety valve position indication does not detect or indicate a significant abnormal degradation of the reactor coolant pressure boundary, as required by Criterion 1 . This is consistent with the NRC Final Policy Statement, which provided that Criterion 1 is intended to ensure that those instruments specifically installed to detect excessive reactor coolant system leakage be included in the TS. Criterion 1 is not to be interpreted to include instrumentation installed to identify the source of actual leakage, for example valve position i ndicators. The instrumentation that satisfies Criterion 1 is contained in Salem Unit 1 TS 3.4.6.1 and Unit 2 TS 3.4. 7.1 "Reactor Coolant System Leakage, Leakage Detection Systems."
The Pressurizer safety relief valves themselves are part of the primary success path in the UFSAR accident analysis because they are assumed to actuate to mitigate a Design Basis Accident (DBA) and therefore meet Criterion 3 of the NRC Final Policy Statement. The operability of the Pressurizer safety valves is therefore required by Unit 1 TS 3.4.2 and Unit 2 TS 3.4.2 & 3.4.3, "Safety Valves." However, the Pressurizer safety valve position indication does not detect or indicate a significant abnormal degradation of the reactor coolant pressure boundary, as required by Criterion 1. This is consistent with the NRC Final Policy Statement, which provided that Criterion 1 is intended to ensure that those instruments specifically installed to detect excessive reactor coolant system leakage be included in the TS. Criterion 1 is not to be interpreted to include instrumentation installed to identify the source of actual leakage, for example valve position indicators. The instrumentation that satisfies Criterion 1 is contained in Salem Unit 1 TS 3.4.6.1 and Unit 2 TS 3.4. 7.1 "Reactor Coolant System Leakage, Leakage Detection Systems."
Pressurizer safety relief valve position i ndication is not a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis considered in Criterion 2.
Pressurizer safety relief valve position indication is not a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis considered in Criterion 2.
4 UFSAR Section 7.5 describes implementation of RG 1.97 at Salem. Type D variables are defined in RG 1.97, as variables that provide information to indicate the operation of individual safety systems and other systems important to safety.
4 UFSAR Section 7.5 describes implementation of RG 1.97 at Salem. Type D variables are defined in RG 1.97, as variables that provide information to indicate the operation of individual safety systems and other systems important to safety.
9
9  


LR-N16-0003                                                                               LAR 516-01 Attachment 1 While the function of the Pressurizer safety relief valves themselves is part of the primary success path in the UFSAR, the valve position indication is not part of the primary success path.
LR-N16-0003 LAR 516-01 While the function of the Pressurizer safety relief valves themselves is part of the primary success path in the UFSAR, the valve position indication is not part of the primary success path.
UFSAR accident analysis assumes that the Pressurizer safety relief valves open as designed to reduce reactor pressure and no operator action based on valve position indication is required.
UFSAR accident analysis assumes that the Pressurizer safety relief valves open as designed to reduce reactor pressure and no operator action based on valve position indication is required.
Therefore, the Pressurizer safety valve position indication is not part of the primary success path as indicated in Criterion 3.
Therefore, the Pressurizer safety valve position indication is not part of the primary success path as indicated in Criterion 3.
The loss of Pressurizer safety valve position indication instrumentation has no effect on the probabilistic safety assessment, and has not been shown to be significant to health and safety as considered in Criterion 4.
The loss of Pressurizer safety valve position indication instrumentation has no effect on the probabilistic safety assessment, and has not been shown to be significant to health and safety as considered in Criterion 4.
Pressurizer Safety Valve position indication does not meet any of the four screening criteria of the NRC Final Policy Statement in 58 Federal Register 391 32 (58FR391 32), which provides criteria to determine which items are required to be included in the TS as LCOs. This conclusion is supported by the absence of operability and surveillance requirements for the Pressurizer Safety Valve position indication in the improved standard TS presented in NUREG-1 431 . Accordingly, removal of the Pressurizer Safety Valve position indication from the TS conforms to the STS.
Pressurizer Safety Valve position indication does not meet any of the four screening criteria of the NRC Final Policy Statement in 58 Federal Register 391 32 (58FR391 32), which provides criteria to determine which items are required to be included in the TS as LCOs. This conclusion is supported by the absence of operability and surveillance requirements for the Pressurizer Safety Valve position indication in the improved standard TS presented in NUREG-1 431. Accordingly, removal of the Pressurizer Safety Valve position indication from the TS conforms to the STS.
Since the position indication for these valves does not provide an indication for operator actions for which no automatic control is provided, it does not satisfy Criterion 1 , 2, 3 or 4 of the NRC Final Policy Statement for the same reasons as discussed for the PORVs above.
Since the position indication for these valves does not provide an indication for operator actions for which no automatic control is provided, it does not satisfy Criterion 1, 2, 3 or 4 of the NRC Final Policy Statement for the same reasons as discussed for the PORVs above.
The Pressurizer Safety Valve indication is not included in the STS. They are not Category 1 in RG 1 .97 nor are they Type A or Category 1 within the Salem UFSAR or Engineering Evaluation S-C-GCS-CEE-0470, thus, the Pressurizer Safety Valve Position Indication is not required to be in the Salem Unit 1 and Unit 2 TS Tables 3.3- 1 1 and 4.3-1 1 .
The Pressurizer Safety Valve indication is not included in the STS. They are not Category 1 in RG 1.97 nor are they Type A or Category 1 within the Salem UFSAR or Engineering Evaluation S-C-GCS-CEE-0470, thus, the Pressurizer Safety Valve Position Indication is not required to be in the Salem Unit 1 and Unit 2 TS Tables 3.3-1 1 and 4.3-1 1.
Therefore, the Pressurizer Safety Valve Position I ndication is proposed to be removed from the Salem Unit 1 and Unit 2 TS Tables 3.3-1 1 and 4.3-1 1 .
Therefore, the Pressurizer Safety Valve Position Indication is proposed to be removed from the Salem Unit 1 and Unit 2 TS Tables 3.3-1 1 and 4.3-1 1.
Containment Pressure (Narrow Range)
Containment Pressure (Narrow Range)
The containment pressure indication provides information for assessing an inadequate containment cooling condition and for determining the potential challenge to the containment pressure retaining integrity. The wide range containment pressure instrumentation, which is also listed in Salem Unit 1 and U nit 2 TS Table 3.3-1 1 (Line Item #1 6), provides an adequate range and sensitivity for this purpose. Only the wide range instrumentation is used in the emergency operating procedures (EOPs) to define the potential for a challenge to containment integrity due to over-pressurization .
The containment pressure indication provides information for assessing an inadequate containment cooling condition and for determining the potential challenge to the containment pressure retaining integrity. The wide range containment pressure instrumentation, which is also listed in Salem Unit 1 and Unit 2 TS Table 3.3-1 1 (Line Item #16), provides an adequate range and sensitivity for this purpose. Only the wide range instrumentation is used in the emergency operating procedures (EOPs) to define the potential for a challenge to containment integrity due to over-pressurization.
If containment heat removal systems are functioning properly, no challenge to containment integrity should occur due to containment pressure. Containment pressure instrumentation is used in the severe accident management guidelines (SAMG) to indicate a possible containment integrity challenge and to initiate the assessment of containment venting strategies. It is also used as an indicator of the potential loss of a fission product barrier in the emergency plan.
If containment heat removal systems are functioning properly, no challenge to containment integrity should occur due to containment pressure. Containment pressure instrumentation is used in the severe accident management guidelines (SAMG) to indicate a possible containment integrity challenge and to initiate the assessment of containment venting strategies. It is also used as an indicator of the potential loss of a fission product barrier in the emergency plan.
Containment pressure is a key indicator in the declaration of a General Emergency level and the potential need for offsite radiological protection actions. Therefore, the containment wide range instrumentation (by itself) satisfies Criterion 4 of 10 CFR 50. 36(c)(2)(ii).
Containment pressure is a key indicator in the declaration of a General Emergency level and the potential need for offsite radiological protection actions. Therefore, the containment wide range instrumentation (by itself) satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).
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LR-N16-0003                                                                             LAR 516-01 The STS requires only the Containment Pressure (Wide Range) instrumentation be included in the PAM TS. Per Salem UFSAR 7. 5-3, the m easurement of Containm ent Pressure is a Type A, Category 1 variable; however, no distinction is m ad e between the use of wide range or narrow range instrum ents.
LR-N16-0003 LAR 516-01 The STS requires only the Containment Pressure (Wide Range) instrumentation be included in the PAM TS. Per Salem UFSAR 7.5-3, the measurement of Containment Pressure is a Type A, Category 1 variable; however, no distinction is made between the use of wide range or narrow range instruments.
Since the redundancy requirements and other requirements of RG 1.97 are m et with the Containment Pressure (Wide Range) channels, the Containment Pressure (Narrow Range) channels are not required to be in the Salem Unit 1 and Unit 2 TS Tables 3. 3-11 and 4. 3-11.
Since the redundancy requirements and other requirements of RG 1.97 are met with the Containment Pressure (Wide Range) channels, the Containment Pressure (Narrow Range) channels are not required to be in the Salem Unit 1 and Unit 2 TS Tables 3.3-11 and 4.3-11.
In support of this LAR, the UFSAR (Tables 7. 5-1, 7.5-3, and 7. 5-4, as appropriate) and Item 11 of S-C-GCS-CEE-0470 will be revised to reflect that the narrow range containment pressure instrum entation is not a Type A variable. PSEG has determ ined these documentation changes can be implemented in accordance with 10 CFR 50. 59. The Salem UFSAR wi ll be revised prior to i mplementation of this proposed TS change.
In support of this LAR, the UFSAR (Tables 7.5-1, 7.5-3, and 7.5-4, as appropriate) and Item 11 of S-C-GCS-CEE-0470 will be revised to reflect that the narrow range containment pressure instrumentation is not a Type A variable. PSEG has determined these documentation changes can be implemented in accordance with 10 CFR 50.59. The Salem UFSAR will be revised prior to implementation of this proposed TS change.
The containment pressure narrow instrument channels themselves are part of the primary success path in the UFSAR accident analysis because they are assum ed to actuate to mitigate a Design Basis Accident ( D BA) and therefore m eet Criterion 3 of the NRC Final Policy Statem ent. The operability of the containm ent pressure narrow range channels is therefore required by Unit 1 and 2 TS 3.3.2.1, "Engineered Safety Feature Actuation System Instrum entation." However, containment pressure narrow range indication d oes not detect or indicate a significant abnormal degradation of the reactor coolant pressure boundary, as required by Criterion 1. This is consistent with the NRC Final Policy Statem ent, which provided that Criterion 1 is intended to ensure that those instruments specifically installed to detect excessive reactor coolant system leakage be included in the TS. Criterion 1 is not to be interpreted to include instrumentation installed to identify the source of actual leakage, for example valve position ind icators.
The containment pressure narrow instrument channels themselves are part of the primary success path in the UFSAR accident analysis because they are assumed to actuate to mitigate a Design Basis Accident (DBA) and therefore meet Criterion 3 of the NRC Final Policy Statement. The operability of the containment pressure narrow range channels is therefore required by Unit 1 and 2 TS 3.3.2.1, "Engineered Safety Feature Actuation System Instrumentation." However, containment pressure narrow range indication does not detect or indicate a significant abnormal degradation of the reactor coolant pressure boundary, as required by Criterion 1. This is consistent with the NRC Final Policy Statement, which provided that Criterion 1 is intended to ensure that those instruments specifically installed to detect excessive reactor coolant system leakage be included in the TS. Criterion 1 is not to be interpreted to include instrumentation installed to identify the source of actual leakage, for example valve position indicators.
Containment pressure narrow range ind ication is not a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis considered in Criterion 2.
Containment pressure narrow range indication is not a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis considered in Criterion 2.
While the function of the containm ent pressure narrow range channels themselves is part of the primary success path in the UFSAR, the pressure indication is not part of the primary success path. UFSAR accident analyses assum e autom atic actuation of the safety injection and containment spray system based on containment narrow range pressures and no operator actions based on containment pressure narrow range indication are required. Therefore, the containment pressure narrow range indication is not part of the prim ary success path as indicated in Criterion 3.
While the function of the containment pressure narrow range channels themselves is part of the primary success path in the UFSAR, the pressure indication is not part of the primary success path. UFSAR accident analyses assume automatic actuation of the safety injection and containment spray system based on containment narrow range pressures and no operator actions based on containment pressure narrow range indication are required. Therefore, the containment pressure narrow range indication is not part of the primary success path as indicated in Criterion 3.
The Containment Pressure narrow range ind ication has no effect on the probabilistic safety assessment, and has not been shown to be significant to health and safety as considered in Criterion 4.
The Containment Pressure narrow range indication has no effect on the probabilistic safety assessment, and has not been shown to be significant to health and safety as considered in Criterion 4.
Containment Pressure (Narrow Range) indication does not meet any of the four screening criteria of the NRC Final Policy Statement in 58 Federal Reg ister 39132 (58FR39132), which provides criteria to determine which items are required to be included in the TS as LCOs. This conclusion is supported by the absence of operability and surveillance requirements for the Containment Pressure ( Narrow Range) indication in the i mproved standard TS presented in 11
Containment Pressure (Narrow Range) indication does not meet any of the four screening criteria of the NRC Final Policy Statement in 58 Federal Register 39132 (58FR39132), which provides criteria to determine which items are required to be included in the TS as LCOs. This conclusion is supported by the absence of operability and surveillance requirements for the Containment Pressure (Narrow Range) indication in the improved standard TS presented in 11  


LR-N16-0 0 0 3                                                                                      LAR 516-01 NUREG-1431. Accord ing ly, removal of Contain ment Pressure (Narrow Range) indication from the TS conforms to the STS.
LR-N16-0003 LAR 516-01 NUREG-1431. Accordingly, removal of Containment Pressure (Narrow Range) indication from the TS conforms to the STS.
Therefore, the Containment Pressure ( Narrow Range) is removed from the Salem Unit 1 and Unit 2 TS Tables 3.3- 1 1 and 4.3-11 .
Therefore, the Containment Pressure (Narrow Range) is removed from the Salem Unit 1 and Unit 2 TS Tables 3.3-1 1 and 4.3-11.
Wide Range Neutron Flux Monitors (Gam ma-Metrics Post-Accident Neutron Monitoring)
Wide Range Neutron Flux Monitors (Gamma-Metrics Post-Accident Neutron Monitoring)
The Neutron Flux is listed as an instrument within the STS, and is listed as Category 1 in RG 1.97, and is Category 1 in Engineering Eval uation S-C-GSC-CEE-0470.
The Neutron Flux is listed as an instrument within the STS, and is listed as Category 1 in RG 1.97, and is Category 1 in Engineering Evaluation S-C-GSC-CEE-0470.
The bases in Westinghouse STS (Reference 4) state that power range and source range n eutron flux ind ication is provided to verify reactor shutdown . The two ranges are necessary to cover the full range of flux that may occur post-accident. It also states that neutron flux is used for accident d iag nosis, verification of sub-criticality, and diag nosis of positive reactivity insertion.
The bases in Westinghouse STS (Reference 4) state that power range and source range neutron flux indication is provided to verify reactor shutdown. The two ranges are necessary to cover the full range of flux that may occur post-accident. It also states that neutron flux is used for accident diagnosis, verification of sub-criticality, and diagnosis of positive reactivity insertion.
Gamma-Metrics Post-Accident Neutron Monitoring (PANM) system was installed at Salem to provide reliable neutron flux monitoring in a harsh envimnment from plant shutdown to full power, and thereby specifically satisfies RG 1 .97 requirements. I n the wide-range mode, the PANM output range meets the requi rement of RG 1.97 (Reference 1 ).
Gamma-Metrics Post-Accident Neutron Monitoring (PANM) system was installed at Salem to provide reliable neutron flux monitoring in a harsh envimnment from plant shutdown to full power, and thereby specifically satisfies RG 1.97 requirements. In the wide-range mode, the PANM output range meets the requirement of RG 1.97 (Reference 1 ).
Therefore, the Wide Range Neutron Flux Monitors are added to the Salem Unit 1 and Unit 2 TS Tables 3.3-11 and 4.3-11 .
Therefore, the Wide Range Neutron Flux Monitors are added to the Salem Unit 1 and Unit 2 TS Tables 3.3-11 and 4.3-11.
Auxiliary Feed Water (AFW) Storage Tank (Condensate Storage Tank 5) Water Level The AFW storage tank water level is listed as a Type A variable per the Salem UFSAR Table 7.5-3 and Category 1 by Table 3 in RG 1 .97. It is also listed as Type A and Category 1 in Eng ineering Evaluation S-C-GCS-CEE-0470. Because of this site-specific determination the i nstrum entation should be added to TS Tables 3.3-11 and 4.3-1 1 for Unit 1 and Unit 2 in accordance with the guidance of the reviewer's notes in NUREG-1431 .
Auxiliary Feed Water (AFW) Storage Tank (Condensate Storage Tank5) Water Level The AFW storage tank water level is listed as a Type A variable per the Salem UFSAR Table 7.5-3 and Category 1 by Table 3 in RG 1.97. It is also listed as Type A and Category 1 in Engineering Evaluation S-C-GCS-CEE-0470. Because of this site-specific determination the instrumentation should be added to TS Tables 3.3-11 and 4.3-1 1 for Unit 1 and Unit 2 in accordance with the guidance of the reviewer's notes in NUREG-1431.
Therefore, the Auxiliary Feed Water (AFW) Storage Tank (Condensate Storage Tank) Water Level is added to the Salem Unit 1 and Unit 2 TS Tables 3.3-11 and 4.3-1 1 .
Therefore, the Auxiliary Feed Water (AFW) Storage Tank (Condensate Storage Tank) Water Level is added to the Salem Unit 1 and Unit 2 TS Tables 3.3-11 and 4.3-1 1.
Exceptions:
Exceptions:
All i nstruments that are Type A variables, or non-Type A, but Category 1 were evaluated and are proposed to be included in the Salem Unit 1 and Unit 2 TS Tables 3.3-1 1 and 4.3-1 1 with the following exceptions:
All instruments that are Type A variables, or non-Type A, but Category 1 were evaluated and are proposed to be included in the Salem Unit 1 and Unit 2 TS Tables 3.3-1 1 and 4.3-1 1 with the following exceptions:
* Containment Isolation Valve (CIV) Position I ndicators PSEG is not proposing to add CIV position indication to TS Tables 3.3-11 and 4 . 3- 1 1.
Containment Isolation Valve (CIV) Position Indicators PSEG is not proposing to add CIV position indication to TS Tables 3.3-11 and 4.3-11.
For valves which receive an automatic isolation sig nal to close in order to accomplish contain ment isolation, valve position is a Type B, Category 1 variable. Type B variables are those which provide information to i ndicate whether plant safety functions are being accomplished . I n the Reference 8 letter to vendor owners 5 In various PAM instrument documents, Condensate Storage Tank water-level is listed because the CST is frequently the source of auxiliary feedwater. At Salem, the AFW storage tank is the source of auxiliary feedwater.
For valves which receive an automatic isolation signal to close in order to accomplish containment isolation, valve position is a Type B, Category 1 variable. Type B variables are those which provide information to indicate whether plant safety functions are being accomplished. In the Reference 8 letter to vendor owners 5 In various PAM instrument documents, Condensate Storage Tank water-level is listed because the CST is frequently the source of auxiliary feedwater. At Salem, the AFW storage tank is the source of auxiliary feedwater.
Hence AFW storage level and CST water level are used interchangeably in this document.
Hence AFW storage level and CST water level are used interchangeably in this document.
12
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LR-N16-0003                                                                             LAR 516-01 groups on application of the policy statement criteria to standard technical specifications, the NRC staff concurred that that only Type A variables meet Criterion 3 as a structure. system or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the I ntegrity of a fission product barrier. However, the staff was unable to confirm the owners groups' conclusion that non-Type A Category 1 instrumentation can be relocated from Technical Specifications I nclusion of CIV position indication in Salem TS Tables 3.3-1 0 and 4.3-1 0 is not required to ensure adequate information is available to operators to verify containment operability and the accomplishment of containment isolation when required . The operability requirements for the containment are controlled by TS 3.6. 1 . Limiti ng Condition for Operation (LCO) 3.6.1 .1 requi res primary CONTAI NMENT I NTEGRITY to be maintained in Modes 1 , 2, 3 and 4.
LR-N16-0003 LAR 516-01 groups on application of the policy statement criteria to standard technical specifications, the NRC staff concurred that that only Type A variables meet Criterion 3 as a structure. system or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the Integrity of a fission product barrier. However, the staff was unable to confirm the owners groups' conclusion that non-Type A Category 1 instrumentation can be relocated from Technical Specifications Inclusion of CIV position indication in Salem TS Tables 3.3-1 0 and 4.3-1 0 is not required to ensure adequate information is available to operators to verify containment operability and the accomplishment of containment isolation when required. The operability requirements for the containment are controlled by TS 3.6.1. Limiting Condition for Operation (LCO) 3.6.1.1 requires primary CONTAINMENT INTEGRITY to be maintained in Modes 1, 2, 3 and 4.
CONTAI NMENT I NTEGRITY requires that all penetrations required to be closed during accident conditions are either capable of being closed by an OPERABLE containment automatic isolation valve system, or closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except for valves that are open under administrative control as permitted by TS 3.6.3.1 .
CONTAINMENT INTEGRITY requires that all penetrations required to be closed during accident conditions are either capable of being closed by an OPERABLE containment automatic isolation valve system, or closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except for valves that are open under administrative control as permitted by TS 3.6.3.1.
The operability requirements for automatic containment isolation valves are controlled in TS 3.6.3.1 (unit 1 ) and TS 3.6.3 (Unit 2). LCO 3.6.3.1 (3.6.3 Unit 2) requires each containment isolation valve to be OPERABLE in Modes 1 , 2, 3 and 4.
The operability requirements for automatic containment isolation valves are controlled in TS 3.6.3.1 (unit 1 ) and TS 3.6.3 (Unit 2). LCO 3.6.3.1 (3.6.3 Unit 2) requires each containment isolation valve to be OPERABLE in Modes 1, 2, 3 and 4.
Surveillance Requirement (SR) 4.6.3. 1 .2 (SR 4.6.3.2 Unit 2) requires each containment isolation valve to be demonstrated OPERABLE by periodically verifying the valves actuate to their isolation positions in response to the required containment isolation signals. SR 4.6.3. 1 .3 (SR 4.6.3.3 U nit 2) requires periodic verification that on a main steam Isolation test signal, each main steam isolation valve actuates to its isolation position. S R 4.6.3. 1 .4 (SR 4.6.3.4 Unit 2) requires the isolation time of each automatic containment isolation valve to be determined to be within its limit when tested pursuant to the l nservice Testing Program. Proper position indication is verified during conduct of surveillance testing of the associated valves.
Surveillance Requirement (SR) 4.6.3. 1.2 (SR 4.6.3.2 Unit 2) requires each containment isolation valve to be demonstrated OPERABLE by periodically verifying the valves actuate to their isolation positions in response to the required containment isolation signals. SR 4.6.3.1.3 (SR 4.6.3.3 Unit 2) requires periodic verification that on a main steam Isolation test signal, each main steam isolation valve actuates to its isolation position. SR 4.6.3.1.4 (SR 4.6.3.4 Unit 2) requires the isolation time of each automatic containment isolation valve to be determined to be within its limit when tested pursuant to the lnservice Testing Program. Proper position indication is verified during conduct of surveillance testing of the associated valves.
CIV position indication was not included in the Amendment which added accident monitoring instrumentation to the Unit 1 Technical Specifications (Reference 9) or in the original U nit 2 Technical Specifications (Reference 1 0). The CIV position indication meets the associated Regulatory Guide 1 .97 requirements and the current TS requirements are sufficient to ensure adequate information is available to operators to verify containment operability and the accomplishment of containment isolation when required.
CIV position indication was not included in the Amendment which added accident monitoring instrumentation to the Unit 1 Technical Specifications (Reference 9) or in the original Unit 2 Technical Specifications (Reference 1 0). The CIV position indication meets the associated Regulatory Guide 1.97 requirements and the current TS requirements are sufficient to ensure adequate information is available to operators to verify containment operability and the accomplishment of containment isolation when required.
Therefore, the Containment Isolation Valve Position I nd icators is not added to the Salem Unit 1 and U nit 2 TS Tables 3.3-1 1 and 4.3-1 1 .
Therefore, the Containment Isolation Valve Position Indicators is not added to the Salem Unit 1 and Unit 2 TS Tables 3.3-1 1 and 4.3-1 1.
* Containment Hydrogen Concentration Containment Hydrogen Concentration is listed as a Type A, Category I variable in Engineering Evaluation S-C-GSC-CEE-0470, not listed as Type A in UFSAR Table 7.5-3, but also listed as Category I in RG 1 .97. However, it is not listed in the STS.
Containment Hydrogen Concentration Containment Hydrogen Concentration is listed as a Type A, Category I variable in Engineering Evaluation S-C-GSC-CEE-0470, not listed as Type A in UFSAR Table 7.5-3, but also listed as Category I in RG 1.97. However, it is not listed in the STS.
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LR-N1 6-0 003                                                                          LAR 516-01 Attachment 1 This instrument was included in the rulemaking for 1 0 CFR 50.44. The statement of considerations for the 50.44 rulemaking states that this instrument can be relocated from the Technical Specifications and can be re-classified as Type C, Category 3 per the RG 1 .97 definitions. Not including these instruments is consistent with Salem LCR S06-05 and Salem TS Amendments 281 and 264 (Reference 7).
LR-N16-0003 LAR 516-01 This instrument was included in the rulemaking for 1 0 CFR 50.44. The statement of considerations for the 50.44 rulemaking states that this instrument can be relocated from the Technical Specifications and can be re-classified as Type C, Category 3 per the RG 1.97 definitions. Not including these instruments is consistent with Salem LCR S06-05 and Salem TS Amendments 281 and 264 (Reference 7).
Therefore, the Containment Hydrogen Monitors are not added to the Salem Unit 1 and Unit 2 TS Tables 3.3-1 1 and 4 .3-1 1 .
Therefore, the Containment Hydrogen Monitors are not added to the Salem Unit 1 and Unit 2 TS Tables 3.3-1 1 and 4.3-1 1.
4.4     LCO Times and Separate Condition Entry The allowed outage times (AOT) in current Technical Specification 3.3.3.7 have their origin in NUREG-0452, "Standard Technical Specifications for Westinghouse Pressurized Water Reactors". The STS (NUREG-1431 ) extended the 7-day completion time for one inoperable instrument channel to 30 days and the 48-hour completion time for two inoperable channels to 7 days. Additionally, the STS removed the shutdown requirement for a sing le inoperable instrument channel.
4.4 LCO Times and Separate Condition Entry The allowed outage times (AOT) in current Technical Specification 3.3.3.7 have their origin in NUREG-0452, "Standard Technical Specifications for Westinghouse Pressurized Water Reactors". The STS (NUREG-1431) extended the 7-day completion time for one inoperable instrument channel to 30 days and the 48-hour completion time for two inoperable channels to 7 days. Additionally, the STS removed the shutdown requirement for a single inoperable instrument channel.
With one channel inoperable beyond 30 days, a special report outlining the preplanned method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels must be submitted to the NRC within the next 14 days. With two channels inoperable for more than 7 days, the STS requires either a plant shutdown or submittal of a special report, as discussed above, depending on the particular channel that is out of service. The STS also contain provisions that permit a separate condition entry for each inoperable instrument function.
With one channel inoperable beyond 30 days, a special report outlining the preplanned method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels must be submitted to the NRC within the next 14 days. With two channels inoperable for more than 7 days, the STS requires either a plant shutdown or submittal of a special report, as discussed above, depending on the particular channel that is out of service. The STS also contain provisions that permit a separate condition entry for each inoperable instrument function.
The STS 30 day completion time is based on operating experience and takes into account the remaining OPERABLE channel (or in the case of a function that has only one required channel, other non-Regulatory Guide 1.97 instrument channels to monitor the Function), the passive nature of the instrument (no critical automatic action is assumed to occur from these instruments), and the low probability of an event requiring PAM instrumentation during this interval.
The STS 30 day completion time is based on operating experience and takes into account the remaining OPERABLE channel (or in the case of a function that has only one required channel, other non-Regulatory Guide 1.97 instrument channels to monitor the Function), the passive nature of the instrument (no critical automatic action is assumed to occur from these instruments), and the low probability of an event requiring PAM instrumentation during this interval.
When the required actions and 30 day completion time is not met, a written report is required to be submitted. This report discusses the results of the root cause evaluation of the inoperability and identifies proposed restorative actions. This action is appropriate in lieu of a shutdown requirement since alternative actions are identified before loss of functional capability, and given the likelihood of unit conditions that would require information provided by this instrumentation.
When the required actions and 30 day completion time is not met, a written report is required to be submitted. This report discusses the results of the root cause evaluation of the inoperability and identifies proposed restorative actions. This action is appropriate in lieu of a shutdown requirement since alternative actions are identified before loss of functional capability, and given the likelihood of unit conditions that would require information provided by this instrumentation.
The STS completion time of 7 days is based on the relatively low probability of an event requiring PAM instrument operation and the availability of alternate means to obtain the required information.
The STS completion time of 7 days is based on the relatively low probability of an event requiring PAM instrument operation and the availability of alternate means to obtain the required information.
Since the TS is being revised to better align to the STS, this amendment revises the following instruments within the Salem Unit 1 and Unit 2 TS Table 3.3-11 to adopt the allowable outage times in STS of 30 days for one inoperable channel less than the required number of channels and 7 days for two or more channels less than the required number of channels (versus the current 7 days and 48 hours, respectively). Other m inor changes to the l isted actions are discussed below.
Since the TS is being revised to better align to the STS, this amendment revises the following instruments within the Salem Unit 1 and Unit 2 TS Table 3.3-11 to adopt the allowable outage times in STS of 30 days for one inoperable channel less than the required number of channels and 7 days for two or more channels less than the required number of channels (versus the current 7 days and 48 hours, respectively). Other minor changes to the listed actions are discussed below.
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LR-N16-0003                                                                             LAR 516-01 Instrument                          Action
LR-N16-0003
: 1. Reactor Coolant Outlet Temperature- THot (Wide Range)         1' 2
: 1.
: 2. Reactor Coolant Outlet Temperature- Tcold (Wide Range)       1' 2
: 2.
: 3.                                                                  1' 2 1' 2 Reactor Coolant Pressure (Wide Range)
: 3.
: 4. Pressurizer Water Level
: 4.
: 5. Steam Line Pressure                                           1' 2
: 5.
: 6. Steam Generator Water Level (Narrow Range)                     1' 2
: 6.
: 7. Steam Generator Water Level (Wide Range)                       1' 2
: 7.
: 8. Refueling Water Storage Tank Water Level                       1' 2
: 8.
: 10. Auxiliary Feedwater Flow Rate                                 4,6
: 10.
: 16.                                                                7,2 17.
: 16.
Containment Pressure -Wide Range 7,2 18.
: 17.
Containment Water Level- Wide Range Core Exit Thermocouples                                       1' 2
: 18.
: 22. Wide Range Neutron Flux Monitors                               1' 2
: 22.
: 23. AFW (Condensate) Storage Tank Water Level                     1' 2 The following TS Table     3 . 3- 1 1 Actions are revised:
: 23.
Action 1 With the number of OPERABLE accident m onitoring channels less than the Required Number of Channels shown in Table 3 . 3-1 1 , restore the inoperable channel(s) to OPERABLE status within 30 days, or submit a special report in accordance with Specification 6 . 9.4.
Instrument Reactor Coolant Outlet Temperature-T Hot (Wide Range)
Action 2 With the number of OPERABLE accident m onitoring channels less than the Minimum Number of Channels shown in Table 3.3- 1 1 , restore the inoperable channel(s) to OPERABLE status within 7 days or be in HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
Reactor Coolant Outlet Temperature-T cold (Wide Range)
Action 4 With the number of OPERABLE channels one less than the Required Number of channels shown in Table 3 . 3- 1 1 , operation may proceed provided that an OPERABLE Steam Generator Wide Range level channel is available as an alternate m eans of indication for the Steam Generator with no OPERAB LE Auxiliary Feedwater Flow Rate channel; OTHE RW ISE, restore the inoperable channel to OPERABLE status within 30 days, or submit a special report in accordance with Specification 6 . 9 . 4 .
Reactor Coolant Pressure (Wide Range)
Action 6 With the number of OPERABLE channels less than the Minimum Number of channels shown in Table 3 . 3-1 1 , restore the inoperable channel(s) to OPERABLE status within 7 days, or be in HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
Pressurizer Water Level Steam Line Pressure Steam Generator Water Level (Narrow Range)
This amendm ent also revises the Salem TS 3 . 3 . 3 . 7 to include the STS note which allows for a separate condition entry for each instrument. This wil l allow each instrument the full associated LCO time, if m ultiple PAM instruments become inoperable at different times.
Steam Generator Water Level (Wide Range)
The surveillance frequencies for Line Items 22 and 23 that are being added to TS Table 4.3-1 1 will be contained in the Surveillance Frequency Control Program with the exception of the 15
Refueling Water Storage Tank Water Level Auxiliary Feedwater Flow Rate Containment Pressure -Wide Range Containment Water Level-Wide Range Core Exit Thermocouples Wide Range Neutron Flux Monitors AFW (Condensate) Storage Tank Water Level The following TS Table 3.3-1 1 Actions are revised:
Action 1 LAR 516-01 Action 1' 2 1' 2 1' 2 1' 2 1' 2 1' 2 1' 2 1' 2 4,6 7,2 7,2 1' 2 1' 2 1' 2 With the number of OPERABLE accident monitoring channels less than the Required Number of Channels shown in Table 3.3-1 1, restore the inoperable channel(s) to OPERABLE status within 30 days, or submit a special report in accordance with Specification 6.9.4.
Action 2 With the number of OPERABLE accident monitoring channels less than the Minimum Number of Channels shown in Table 3.3-1 1, restore the inoperable channel(s) to OPERABLE status within 7 days or be in HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
Action 4 With the number of OPERABLE channels one less than the Required Number of channels shown in Table 3.3-1 1, operation may proceed provided that an OPERABLE Steam Generator Wide Range level channel is available as an alternate means of indication for the Steam Generator with no OPERABLE Auxiliary Feedwater Flow Rate channel; OTHERWISE, restore the inoperable channel to OPERABLE status within 30 days, or submit a special report in accordance with Specification 6.9.4.
Action 6 With the number of OPERABLE channels less than the Minimum Number of channels shown in Table 3.3-1 1, restore the inoperable channel(s) to OPERABLE status within 7 days, or be in HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
This amendment also revises the Salem TS 3.3.3. 7 to include the STS note which allows for a separate condition entry for each instrument. This will allow each instrument the full associated LCO time, if multiple PAM instruments become inoperable at different times.
The surveillance frequencies for Line Items 22 and 23 that are being added to TS Table 4.3-1 1 will be contained in the Surveillance Frequency Control Program with the exception of the 1 5


LR-N16-0 0 0 3                                                                                      LAR S16-0 1 C h a n n e l Functional Test w h i ch is not appl icable for these i n stru ments. Salem has i m pleme nted TSTF Change Trave l e r TSTF-42 5 , Revision 3 ( Reference 1 1 ) as approved by the N RC with TS A m e n d ments 299 and 282 ( Refere n ce 1 2 ) for U nits 1 and 2 respectively. The in itial s u rveilla nce freq u en cy will be 31 -days for the CHA N N E L CH ECK and 1 8-months for the CHAN N EL CAL I B RAT I O N consistent with N U RE G - 1 431 , Stan d a rd Techn ical Specifications for Westi ng house Plants.
LR-N16-0003 LAR S16-01 Channel Functional Test which is not applicable for these instruments. Salem has implemented TSTF Change Traveler TSTF-425, Revision 3 ( Reference 1 1 ) as approved by the NRC with TS Amendments 299 and 282 ( Reference 1 2) for Units 1 and 2 respectively. The initial surveillance frequency will be 31 -days for the CHANNEL CHECK and 1 8-months for the CHANNEL CALIBRATION consistent with NUREG-1 431, Standard Technical Specifications for Westinghouse Plants.
The criteria for relocation of a s u rve i l l ance freq uency to a l icensee controlled p rogram in accord ance with TSTF-425 were reviewe d . These s u rvei l lance freq ue n cies a re pe riod ic s u rvei lla nces that: 1 ) d o not reference othe r approved p rograms for the specified i nterva l , 2 ) are not event d rive n , 3) do not h ave a ti me component based on event occurrence, and 4) a re not related to a s pecific co nditi o n for performance. The refore the periodic s u rvei l lance freq uencies a re with i n the scope of TSTF-425 for location in the l i censee controlled S u rvei l lance F req uency Control Prog ra m .
The criteria for relocation of a surveillance frequency to a licensee controlled program in accordance with TSTF-425 were reviewed. These surveillance frequencies are periodic surveillances that: 1 ) do not reference other approved programs for the specified interval, 2) are not event driven, 3) do not have a time component based on event occurrence, and 4) are not related to a specific condition for performance. Therefore the periodic surveillance freq uencies are within the scope of TSTF-425 for location in the licensee controlled Surveillance Frequency Control Program.
Fi nally, T S 6 . 9.4, Special Repo rts , is revised to include Actions 1 a n d 4 of TS Ta ble 3 . 3-1 1 .
Finally, TS 6.9.4, Special Reports, is revised to include Actions 1 and 4 of TS Table 3.3-1 1.  
5.0        REG ULATORY ANALYSIS
: 5. 1      No Significant H azards Consideratio n P S E G req uests a n a m e nd ment to t h e Salem U nit 1 and U n it 2 Operating Licenses. The p roposed changes wou l d m od ify Accident Mo nitoring Tech n ical Specifications (TS) Tables 3 . 3-1 1 and 4 . 3-1 1 , titled "Accident Mon itoring I nstru me ntation" and " S u rvei l lance Req u i reme nts for Accident Mon itori n g I n strumentation" respectively. S pecifica l ly, these changes wo uld add Reg u l atory G u i d e ( RG ) 1 . 97 Type A or Category 1 i nstrume nts that are cu rrently absent i nto Tables 3 . 3-1 1 and 4 . 3- 1 1 , and wou ld re move i nstru m ents that a re not Type A, nor Category 1 from Tables 3 . 3-1 1 a n d 4 . 3 - 1 1 . The specific i nstrume nts to be removed or added a re s u m ma rized below:
I n stru ments rem oved from TS Table 3 . 3- 1 1 and TS Table 4 . 3-1 1 because they are not reg u latory g u ide ( RG ) 1 . 97 Type A, or Category 1 :
* Reacto r Coolant System Sub-Cooling Margin Mon itor*
* PORV Position I n d icator ( U nit 2 only)*
* PORV B lock Valve Position I n d i cator*
* Pressu rizer Safety Valve Position I n d icator*
* Conta i n m ent Press u re (Na rrow Ra nge)
* Th e associated footnotes a re also removed.
I nstru ments added to TS Table 3 . 3-1 1 and TS Table 4.3-1 1 :
* Wide Ran g e N eutron Flux Monitors
* Auxi l i a ry Feed Water (AFW) Storage Ta n k ( Condensate Storage Tank) Water Level Add itional ly, allowed outage times (AOT) and req u i red actions of TS Ta ble 3 . 3-1 1 Actions 1 , 2 ,
4 a n d 6 wo u l d b e revised to align with the Westinghouse Standard Tec h n i ca l Specifications (STS), N U R E G - 1 431 . TS Section 6 . 9 . 4 is revised to reflect these changes to Actions.
16


LR-N1 6-0003                                                                             LAR 516-01 PSEG has evaluated the proposed changes to the TS using the criteria in 1 0 CFR 50.92, and determ ined that the proposed changes do not involve a significant hazards consideration. The following information is provided to support a finding of no significant hazards;
==5.0 REGULATORY ANALYSIS==
: 1. Does the proposed change involve a significant increase in the probability or conseq uences of an accident previously evaluated?
: 5. 1 No Significant Hazards Consideration PSEG req uests an amend ment to the Salem Unit 1 and U nit 2 Operating Licenses. The proposed changes would modify Accident Monitoring Technical Specifications (TS) Tables 3.3-1 1 and 4.3-1 1, titled "Accident Monitoring Instrumentation" and "Surveillance Requirements for Accident Monitoring Instrumentation" respectively. Specifically, these changes would add Regulatory Guide ( RG ) 1.97 Type A or Category 1 instruments that are currently absent into Tables 3.3-1 1 and 4.3-1 1, and would remove instruments that are not Type A, nor Category 1 from Tables 3.3-1 1 and 4.3-1 1. The specific instruments to be removed or added are summarized below:
Response: No The proposed changes to the TS modify Accident Monitoring I nstrumentation TS Tables 3.3- 1 1 and 4.3-1 1 of Salem Units 1 and 2 by removing or adding instruments as listed above, and updating the AOT and required actions to better align with the Westing house STS, N U REG- 1 43 1 . The instruments listed above are not assumed to be initiators of any analyzed event of Chapter 1 5 in the U pdated Final Safety Analysis Report ( U FSAR).
Instruments removed from TS Table 3.3-1 1 and TS Table 4.3-1 1 because they are not regulatory guide ( RG ) 1.97 Type A, or Category 1 :
Reactor Coolant System Sub-Cooling Margin Monitor*
PORV Position Indicator (Unit 2 only)*
PORV Block Valve Position Indicator*
Pressurizer Safety Valve Position Indicator*
Containment Pressure (Narrow Range)
* The associated footnotes are also removed.
I nstruments added to TS Table 3.3-1 1 and TS Table 4.3-1 1 :
Wide Range Neutron Flux Monitors Auxiliary Feed Water (AFW) Storage Tank (Condensate Storage Tank) Water Level Additionally, allowed outage times (AOT) and required actions of TS Table 3.3-1 1 Actions 1, 2, 4 and 6 would be revised to align with the Westinghouse Standard Technical Specifications (STS), NUREG-1 431. TS Section 6.9.4 is revised to reflect these changes to Actions.
1 6
 
LR-N16-0003 LAR 516-01 PSEG has evaluated the proposed changes to the TS using the criteria in 1 0 CFR 50.92, and determined that the proposed changes do not involve a significant hazards consideration. The following information is provided to support a finding of no significant hazards; 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No The proposed changes to the TS modify Accident Monitoring Instrumentation TS Tables 3.3-1 1 and 4.3-1 1 of Salem Units 1 and 2 by removing or adding instruments as listed above, and updating the AOT and required actions to better align with the Westinghouse STS, NUREG-1 431. The instruments listed above are not assumed to be initiators of any analyzed event of Chapter 1 5 in the Updated Final Safety Analysis Report (UFSAR).
Therefore the probability of an accident previously evaluated is not significantly increased.
Therefore the probability of an accident previously evaluated is not significantly increased.
The proposed changes do not alter the design of any system , structure, or component (SSC). The proposed changes conform to NRC regulatory gu idance regard ing the content of plant TS, as identified in 1 0 CFR 50.36, NUREG- 1 43 1 , and the NRC Final Policy Statement in 58 FR 391 32.
The proposed changes do not alter the design of any system, structure, or component (SSC). The proposed changes conform to NRC regulatory guidance regarding the content of plant TS, as identified in 1 0 CFR 50.36, NUREG-1 431, and the NRC Final Policy Statement in 58 FR 391 32.
TS O perability req uirements are retained for Type A and Category 1 variables. Operability of these instruments ensures sufficient information is available to monitor and assess plant status d u ring and fol lowing an accident. Alternate means for d iagnosing and responding to instrument malfunctions are u naffected by the proposed change. Therefore, the consequences of an accident previously evaluated are not significantly increased .
TS Operability requirements are retained for Type A and Category 1 variables. Operability of these instruments ensures sufficient information is available to monitor and assess plant status during and following an accident. Alternate means for diagnosing and responding to instrument malfunctions are unaffected by the proposed change. Therefore, the consequences of an accident previously evaluated are not significantly increased.
Therefore, these proposed changes do not represent a significant increase in the probability or consequences of an accident previously evaluated.
Therefore, these proposed changes do not represent a significant increase in the probability or consequences of an accident previously evaluated.
: 2. Does the proposed change create the possibility of a new o r different kind of accident from any accident previously evaluated?
2. Does the proposed change create the possibility of a new o r different kind of accident from any accident previously evaluated?
Response: No The proposed changes to the TS would modify the TS Tables 3.3- 1 1 and 4 . 3- 1 1 of Salem U nits 1 and 2, by removing or adding instru ments as listed above, and updating the AOT and req uired actions to better align with the Westing house STS. The proposed changes do not involve a modification to the physical configuration of the plant or changes in the methods governing normal plant operation. The proposed changes will not im pose any new or different req uirement or introduce a new accident initiator, accident precursor, or malfunction mechanism.
Response: No The proposed changes to the TS would modify the TS Tables 3.3-1 1 and 4.3-1 1 of Salem Units 1 and 2, by removing or adding instruments as listed above, and updating the AOT and required actions to better align with the Westinghouse STS. The proposed changes do not involve a modification to the physical configuration of the plant or changes in the methods governing normal plant operation. The proposed changes will not impose any new or different requirement or introduce a new accident initiator, accident precursor, or malfunction mechanism.
Additionally, there is no change in the types or increases in the amounts of any effl uent that may be released off-site and there is no increase in individual or cumulative occupational exposure .
Additionally, there is no change in the types or increases in the amounts of any effluent that may be released off-site and there is no increase in individual or cumulative occupational exposure.
Therefore , the proposed changes do not create t h e possibility o f a new or different kind of accident from any accident previously evaluated.
Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
17
1 7


LR*N16*0003                                                                                                 LAR 816-01
LR*N16*0003
: 3. Do t h e proposed changes involve a s i g n ificant red uction i n a m a rg i n o f safety?
: 3.
Response: No The proposed changes to the TS wou ld mod ify the TS Tables 3 . 3- 1 1 and 4.3-1 1 of Salem U n its 1 and 2 , by rem oving o r a d d i ng instru ments as l isted above , and u pdating the AOT a n d req u i red actions to better a l i g n with the Westi ng house STS . T h e instru ments removed from Tables 3 . 3- 1 1 and 4 . 3 - 1 1 a re n ot needed for manual operator action necessary for safety systems to accomplish their safety fu n ction for the design basis events. The i nstru ments listed for removal a re i n d ication-only with the exceptio n of contai nm ent pressure n a rrow range instrume nts ; t h u s , they do not p rovide an in put to any a utomatic tri p fu nctions.
Do the proposed changes involve a significant reduction in a margin of safety?
In the case where s i m i l a r o r related i n stru m e nts (e . g . , conta i n ment p ressure-na rrow ra nge) a re associated with i m portant trips ( i. e . , RPS or E S F trips), such i nstru ments are governed by separate existing TS sections which a re not altered by this req uest.
Response: No LAR 816-01 The proposed changes to the TS would modify the TS Tables 3.3-1 1 and 4.3-1 1 of Salem Units 1 and 2, by removing or adding instruments as listed above, and updating the AOT and required actions to better align with the Westinghouse STS. The instruments removed from Tables 3.3-1 1 and 4.3-1 1 are not needed for manual operator action necessary for safety systems to accomplish their safety function for the design basis events. The instruments listed for removal are indication-only with the exception of containment pressure narrow range instruments; thus, they do not provide an input to any automatic trip functions.
Therefore, since the proposed changes do not im pact the response of the plant to a design basis accident, the proposed changes do not involve a sig n ificant red uction in a marg i n of safety.
In the case where similar or related instruments (e.g., containment pressure-narrow range) are associated with important trips (i. e., RPS or ESF trips), such instruments are governed by separate existing TS sections which are not altered by this req uest.
Based u pon the above , PSEG concl udes that the proposed a m e n d m e nt p resents no significant haza rd s consideration under the sta ndards set forth in 1 0 C F R 5 0 . 92(c), and , accord ing ly, a fi n d i ng of "no sign ificant h azards conside ration" is j u stified.
Therefore, since the proposed changes do not impact the response of the plant to a design basis accident, the proposed changes do not involve a significant reduction in a margin of safety.
In concl usio n , based on the consideration s discussed above, ( 1 ) there is a reasonable assura n ce that the h ea lth and safety of the public wi l l not be endangered by operation i n the p ro posed m a n ner, (2) such activities wi ll be co nducted i n com pliance with the N RC's reg u lations, and (3) the issuance of the amend ment wi l l not be i n i m ical to the com mon defense a n d secu rity or to the health and safety of the public.
Based upon the above, PSEG concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 1 0 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.
5.2         Appl icable Regulatory Req u i re m e nts/Criteria 1 0 C F R 5 0 . 36(c)(2), Limiting conditions for operation, states: ( i ) L i m iting co nditions for o p e rati o n are the lowest fu nctional capab i l ity or pe rformance l eve ls of eq u i pm ent req u i red for safe operation of the facility. When a l i m iti ng co nd ition for operation of a n u clear reactor is not m et, the l icensee shall shut down the reactor or follow any remed ial action permitted by the tech nical specificatio ns u nti l the condition can be met.
In conclusion, based on the considerations discussed above, ( 1 ) there is a reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in com pliance with the NRC's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Reg u latory G u ide (RG) 1 . 97, "I n stru m e ntation for Light-Water-Cooled N u clear Power Plants to Assess Plant and Environs Cond itions During and Following a n Accid e nt". The proposed changes to the Salem TS do not alter the classification of instru m e ntation req u ired by RG 1 . 97.
5.2 Applicable Regulatory Requirements/Criteria 1 0 CFR 50.36(c)(2), Limiting conditions for operation, states: (i) Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met.
The proposed changes only m od ify the i n stru me nts listed in TS 3/4. 3 . 3 . 7 .
Regulatory G uide (RG) 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident". The proposed changes to the Salem TS do not alter the classification of instrumentation required by RG 1.97.
T h e refore , based on t h e considerations d iscussed above, ( 1 ) there is reaso nable assurance that the h ea lth and safety of the p u b l i c wi l l not be enda ngered by operation i n the proposed manne r,
The proposed changes only modify the instruments listed in TS 3/4.3.3.7.
( 2 ) such activities wi l l be con d u cted i n compliance with the Com m ission's reg u lations, and (3) the issuance of the amendment will not be i n i m ical to the com m o n d efense and secu rity or to the h ealth a n d safety of the p u b l i c .
Therefore, based on the considerations discussed above, ( 1 ) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
18
1 8


LR-N16-000 3                                                                                              LAR S16-01 6.0      ENVIRON M ENTAL CONSIDERATION A review has determ i n ed that the proposed a m e n d m ent would change a req u i rement with respect to instal lation or use of a faci lity component located within the restricted a rea, as defined i n 1 0 C F R 2 0 , or wo u l d change a n inspection or surve i l l a n ce req u i reme nt. H owever, the p roposed am endm ent does n ot involve (i) a s i g n ificant h azards co nsiderati o n , (ii) a s i g n ificant change in the types or s i g n ificant i n crease in the a m o u nts of any effl uent that may be released offs ite, or ( i i i ) a sig n ifica nt i ncrease i n ind ivid ual or cu m u lative occupational rad iation exposu re .
LR-N16-0003
Accord i ng ly, t h e proposed a m e nd m e nt m eets t h e e l i g i b i l ity criterion for categorical exclusion set forth in 1 0 C F R 51 . 2 2 ( c)(9). Therefore , p u rsuant to 1 0 C F R 5 1 . 22(b), no e nvi ron mental im pact state m e nt or enviro n m e ntal assessment need be p re p a red in connection with the proposed a m e n d me nt.
7.0      REF ERENCES
: 1. Reg ulatory G u i d e 1 . 97, " I nstrume ntation for Lig ht-Water-Cooled Nuclear Power Plants to Assess Plant a nd Environs Conditions D u ri n g and Fol lowing a n Accid ent" , Revision 2 d ated Dece m b e r 1 980
: 2. Salem Generating Station (SGS) U FSAR, Section 7.5, "Safety-Related Display I nstru mentation"
: 3. Salem U n i t 1 a n d U n it 2 Techn ical Specification Table, 3 . 3- 1 1 , Accident Monitoring I nstrum entatio n , and Table 4.3-1 1 , Surve i l la n ce Req u i reme nts for Accident Mon itoring I nstrum ents
: 4. N U REG-1 43 1 , Vol u m e 1 , Specifications, Revisi o n 4 . 0 , "Stand ard Tech n ical Specifications - Westi n g ho use Plants," dated April 2 0 1 2 (ADAMS Accession N o .
M L 1 2 1 00A222 )
: 5. S-C-GSC-CE E-0470, " E n g i neering Eval uation o f S G S 1 & 2 Reg ulatory G u i d e 1 . 97 I nstru mentation Com p l i a n ce with Physical Separation and Electrical Isolation Crite ri a , "
Revis ion 1 , J u l y 1 996
: 6. PSEG Emerg e n cy License Am endm ent Req u est to Remove Pressu rizer Power Operated Rel ief Valve (PORV) Position I n d icatio n I nstru mentation from the Accident Monitoring I nstru me ntation Tech n ical Specifications, dated Aug ust 31 , 201 5 (ADAMS Accession No. ML15243A491 )
: 7. PSEG License Amendment Req u est rega rd i ng E l i m i nation of Req u i rements for H yd rogen Reco m b i n e rs and H yd rogen Analyzers Tech n i cal Specifications, d ated J u n e 7 , 2006
: 8. N RC Letter to Owne rs G ro u p o n the N RC Staff Review of N u clear Steam Supply System Vendor Owners Groups' Application of the Com m ission's I nterim Pol icy Statement Criteria to Sta n d a rd Tech n i cal Specifications, d ated May 9, 1 988 19


LR-N16-0003                                                                                  LAR 516-01
==6.0 ENVIRONMENTAL CONSIDERATION==
: 9. N RC approval letter reg ard i n g Amendment N o . 39 which i n corporates the req u i rements for i m plementatio n of the T M I-2 Lessons Learned Category "A" Items for Salem N uclear Generating Stati o n , U n it No. 1 , dated October 08, 1 98 1 1 0 . N U REG-0546 , "Technical Specifications for S a l e m N u clear Generati ng Station U nit N o . 2 , " dated May 1 98 1 1 1 . Tech n i cal Specification Task Force (TSTF) Change Traveler TSTF-42 5 , Revision 3 ,
LAR S16-01 A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 1 0 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cum ulative occupational radiation exposure.
          "Relocate S u rvei l lance Freq u encies t o Licensee Control - RITSTF I n itiative 5 b , " N RC approved J u l y 6, 2009 1 2 . Letter Richard B. E n n i s ( N RC) to Thomas Joyce ( P S E G ) , "Salem N u clear Generati ng Statio n , U n it Nos. 1 and 2, Issuan ce of Ame n d m e nts RE: Relocation of Specific S u rvei l lance Freq u e n cies to a Licensee-Controlled Program Based on Technical Specification Task Force (TSTF ) Change TSTF-425 (TAC Nos. M E3574 and M E3575) , "
Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 1 0 CFR 51.22(c)(9). Therefore, pursuant to 1 0 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.  
March 2 1 , 2 0 1 1 1 3 . N U REG-0737, "Clarification of T M I Action Plan Req u i rem ents , " d ated Nove m ber 1 980 20


LR-N16-0003                                                                    LAR S16-01 Attachment 2 Mark-up of Proposed Technical Specification Pages The following Technical Specifications pages for Renewed Facility Operating License DPR-70 are affected by this change req uest:
==7.0 REFERENCES==
Technical Specification 3 . 3 . 3 . 7, Accident Monitoring I nstrumentation                3/4 3-53 Table 3.3-1 1 , Accident Monitoring I nstrumentation                3/4 3-54 Table 3.3-1 1 , Accident Monitoring I nstrumentation               3/4 3-55 Table 3.3-1 1 , Accident Mon itoring I nstrumentation              3/4 3-56 Table 3.3-1 1 , Accident Monitoring I nstrumentation               3/4 3-56a Table 4 . 3-1 1 , Surveillance Req uirements for Accident Monitoring I nstrumentation                    3/4 3-57 Table 4.3-1 1 , Surveillance Req uirements for Accident Mon itoring I nstrumentation                    3/4 3-57a 6 . 9.4, Special Reports                                            6-24b The following Technical Specifications pages for Renewed Facility Operating License DPR-75 are affected by this change req uest:
1. Regulatory Guide 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident", Revision 2 dated December 1 980
Technical Specification 3 . 3 . 3. 7 , Accident Monitoring I nstrumentation                3/4 3-50 Table 3.3-1 1 , Accident Monitoring I nstru mentation              3/4 3-51 Table 3.3-1 1 , Accident Monitoring I nstrumentation               3/4 3-5 1 a Table 3.3-1 1 , Accident Monitoring I nstrumentation               3/4 3-5 1 b Table 3 . 3-1 1 , Accident Monitoring I nstrumentation             3/4 3-5 1 c Table 4.3-1 1 , Surveillance Requirements for Accident Monitoring I nstrumentation                    3/4 3-52 Table 4.3-1 1 , Surveillance Req uirements for Accident Monitoring I nstrumentation                    3/4 3-52a 6 . 9 .4, Special Reports                                          6-24b 1
: 2.
Salem Generating Station (SGS) UFSAR, Section 7.5, "Safety-Related Display I nstrumentation"
: 3.
Salem Unit 1 and Unit 2 Technical Specification Table, 3.3-1 1, Accident Monitoring I nstrumentation, and Table 4.3-1 1, Surveillance Requirements for Accident Monitoring I nstruments
: 4.
N UREG-1 431, Volume 1, Specifications, Revision 4.0, "Standard Technical Specifications - Westinghouse Plants," dated April 201 2 (ADAMS Accession No. ML12100A222)
: 5. S-C-GSC-CE E-0470, "Engineering Evaluation of SGS 1 & 2 Regulatory Guide 1.97 I nstrumentation Compliance with Physical Separation and Electrical Isolation Criteria,"
Revision 1, July 1 996
: 6. PSEG Emergency License Amendment Request to Remove Pressurizer Power Operated Relief Valve (PORV) Position Indication I nstrumentation from the Accident Monitoring I nstrumentation Technical Specifications, dated August 31, 201 5 (ADAMS Accession No. ML15243A491 )
: 7.
PSEG License Amendment Request regarding Elimination of Req uirements for Hyd rogen Recombiners and Hydrogen Analyzers Technical Specifications, dated June 7, 2006 8. N RC Letter to Owners Group o n the NRC Staff Review of Nuclear Steam Supply System Vendor Owners Groups' Application of the Com mission's I nterim Policy Statement Criteria to Standard Technical Specifications, dated May 9, 1 988 1 9


I N STRU M E NTAT I O N ACC I DENT M O N ITOR I N G I N STR U M ENTAT I O N LI M IT I NG CONDITION FOR OPE RATION 3 . 3 . 3 . 7 The accident monitoring instrumentation ch a nnels s h own in Table 3.3 1 1 s hall be operable.
LR-N16-0003 LAR 516-01
APPL I CAB I LITY: M O DE S 1 , 2, and 3 .
: 9.
ACT I O N :
NRC approval letter regarding Amendment No. 39 which incorporates the requirements for implementation of the TMI-2 Lessons Learned Category "A" Items for Salem Nuclear Generating Station, U nit No. 1, dated October 08, 1 981 1 0. NUREG-0546, "Technical Specifications for Salem Nuclear Generating Station Unit No. 2," dated May 1 981 1 1. Technical Specification Task Force (TSTF) Change Traveler TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5b," NRC approved July 6, 2009 1 2. Letter Richard B. Ennis (NRC) to Thomas Joyce (PSEG), "Salem Nuclear Generating Station, Unit Nos. 1 and 2, Issuance of Amendments RE: Relocation of Specific Surveillance Frequencies to a Licensee-Controlled Program Based on Technical Specification Task Force (TSTF) Change TSTF-425 (TAC Nos. M E3574 and M E3575),"
: a. As shown in T a b l e 3.3-1 1 .
March 21, 201 1 1 3. NUREG-0737, "Clarification of TMI Action Plan Requirements," dated November 1 980 20
SURVEILLA N C E REQU I R E M E N TS 4 . 3 . 3 . 7 Each accident mon itoring instrumentation channel s h a l l be dem onstrated OPERABLE by performa n ce of the CHAN N E L CHECK, CHAN N EL CALI B RATI O N and C HAN N EL FUNCTIONAL TEST operati ons at the frequencies specified in the S u rve i l la nce Frequency Control Program u n less otherwise noted in Table 4. 3- 1 1 .
        -------------------------------------------- ()liE:-------------------------------------------------
..____-! Separate Condition entry is allowed for each Function .
SALEM - U N I T 1                                   3/4 3-53                              Ame ndment No. 299


TABLE 3.3- 1 1 8CCI DENT MONITORING INSTRUMENTATION REQUIRED      MINIMUM N O. O F    N O . OF I NSTRUMENT                                      C HANN ELS  C HANNELS      ACTION
LR-N16-0003 LAR S16-01 Mark-up of Proposed Technical Specification Pages The following Technical Specifications pages for Renewed Facility Operating License DPR-70 are affected by this change request:
: 1. Reactor Coolant Outlet Temperature -              2                      1, 2 T Hor (Wide Range)
Technical Specification 3.3.3. 7, Accident Monitoring Instrumentation Table 3.3-1 1, Accident Monitoring Instrumentation Table 3.3-1 1, Accident Monitoring Instrumentation Table 3.3-1 1, Accident Monitoring Instrumentation Table 3.3-1 1, Accident Monitoring Instrumentation Table 4.3-1 1, Surveillance Requirements for Accident Monitoring Instrumentation Table 4.3-1 1, Surveillance Requirements for Accident Monitoring Instrumentation 6.9.4, Special Reports 3/4 3-53 3/4 3-54 3/4 3-55 3/4 3-56 3/4 3-56a 3/4 3-57 3/4 3-57a 6-24b The following Technical Specifications pages for Renewed Facility Operating License DPR-75 are affected by this change request:
: 2. Reactor Coolant Inlet Temperature -               2                      1, 2 T coLo (Wide Ran ge)
Technical Specification 3.3.3.7, Accident Monitoring I nstrumentation Table 3.3-1 1, Accident Monitoring Instrumentation Table 3.3-1 1, Accident Monitoring I nstrumentation Table 3.3-1 1, Accident Monitoring Instrumentation Table 3.3-1 1, Accident Monitoring Instrumentation Table 4.3-1 1, Surveillance Requirements for Accident Monitoring I nstrumentation Table 4.3-1 1, Surveillance Requirements for Accident Monitoring Instrumentation 6.9.4, Special Reports 1
: 3. Reactor Coolant Pressure                          2                      1, 2 (Wide Range)
3/4 3-50 3/4 3-51 3/4 3-51 a 3/4 3-51 b 3/4 3-51 c 3/4 3-52 3/4 3-52a 6-24b
: 4. Pressurizer Water Level                            2            1         1' 2
: 5. Steam Line Pressure                          2/Steam      1/Steam        1, 2 Generator    Generator
: 6. Steam Generator Water Level                  2/Steam      1 /Steam        1, 2 (Narrow Range)                              Generator    Generator
: 7. Steam Generator Water Level                4 ( 1 /Steam 3 ( 1 /Steam      1' 2 (Wide Range)                                Generator)  Generator)
: 8. Refueling Water Storage Tank                      2                      1, 2 Water Level
: 9. deleted 1 0. Auxiliary Feedwater Flow Rate                4 ( 1 /Steam 3 ( 1 /Steam      4, 6 Generator)  Generator) 1 1 . Reactor Geelant System                            a            +
Margin Monlrof 1 2 . Deleted Yoeleted]
SALEM - U NIT 1                            3/4 3-54                  Amendment No. 3 1 0


TABLE 3.3- 1 1 (CONTINUED)
I NSTRUMENTATION ACCI DENT MON ITORI NG I N STRUMENTATION LIMITI NG CONDITION FOR OPERATION 3.3.3.7 The accident monitoring instrumentation channels shown in Table 3.3 1 1 shall be operable.
ACCI DENT MONITO R I NG INSTRUMENTATION REQUIRED            M INIMUM N O. OF            NO. OF I N STRUM ENT 1 3. PORV Blook Val'le Position Indicator CHANNELS 2/valve**
APPLICABI LITY: MODES 1, 2, and 3.
CHANNELS ACTION 1 4 . Pressurizer Safety Valve Position                  2/valve                              4-;--2 lndioator Deleted 1 5. Gontainment Pressure        Narro*HRange              2                            4-;-4 1 6. Containment Pressure      - Wide Ran }e--1 Deleted    12                  1           7, 2 1 7 . Containment Water Level -                             2                  1           7, 2 Wide R ange 1 8. Core Exit Thermocouples                        4/core quadrant    2/core quadrant      1 2 t
ACTION:
1 9. Reactor Vessel Level Instrumentation                    2                  1            8, 9 System (RVUS)
: a.
: 20. Containment High Range Accident                          2                  2              10 Radiation Monitor 2 1 . Main Steamline Discharge ( Safety                1 /MS Line          1 /MS Line          10 Valves and Atmospheric Steam Dumps) Monitor Wide Range N eutron Flux Monitors                    2                    1            1t 2 2 3 . Auxiliary Feed Water Storage Tank                      2                    1           1' 2
As shown in Table 3.3-1 1.
( Condensate Storage Tank ) Water Level Total number ofls is aonsid to be t\vo (2) with ono (1 of the ohannol& bemg any one (1 ) of the follo*Ning alternate means of determining PORV BlocK, or Safety Valve Tailpipe TemperatuFOs for the valves, Pressurizer Relief Tank Temperature Pressurizer Relief Tank Level OPERABb-,
SURVEILLANCE REQUIREMENTS 4.3.3.7 Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, CHAN N EL CALI BRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies specified in the S u rveillance Frequency Control Program unless otherwise noted in Table 4.3-1 1.
Delete SALEM - UNIT 1                                                                   Amendment No. 3 1 0
!--------------------------------------------"()liE:-------------------------------#-------$---------$--%%%
..___-!Separate Condition entry is allowed for each Function.
SALEM - U N IT 1 3/4 3-53 Amendment No. 299


TABLE 3   . 3 - ll ( continued)
TABLE 3.3-1 1 8CCIDENT MONITORING INSTRUMENTATION REQUIRED MINIMUM NO. OF NO. OF INSTRUMENT CHANNELS CHANNELS ACTION 1. Reactor Coolant Outlet Temperature -
HOT STA N D BY with in the next 6                                           submit a special report in TABLE NOTATION accordance with         ecification 6 . 9.4 Wi th the number of OPERABLE ac c i den t moni toring channels                         than the Required Number of Channels             shown in Table         3 . 3 - 11 , res     the wi thin   f   days , or HM'   BHO'l'DOWN within tihe fti!iXtl 12 hG'U:Il'S .
2 1, 2 THor (Wide Range)
inoperable c hannel ( s )    to OPERABLE s tatus With the numb er of OPERABLE accident mon i t or in g channels less the MINIMUM Number of Chann e l s shown in T ab l e             3 . 3 - l l , restore the than inoperable chann e l ( s ) to OPBRABLE s tatus wi thin   or be in HOT   SHUTDOWN wi thi n the   hours .
: 2. Reactor Coolant Inlet Temperature -
ACTION 3  del eted                          following        6 I ACTION 4                                                            le s s than the Re quired Numb er of Ch ann e l s shown i n Table       3 . 3 11 , operation may proceed With the number of OPERABLE channe l s one provided that an OPERABLE S team Generator Wide Range Level channel is avai l ab l e as an a l tern a t e means of indication f o r the S t e am Generator with no OPERATABLE Auxi liary reedwa ter P l ow Ra t e channel.
2 1, 2 T coLo (Wide Range)
II\
: 3. Reactor Coolant Pressure 2
ACTION 5    deleted
1, 2 (Wide Range)
                                          ; OTH E RWISE, restore the inoperable channel to OPERABLE status with in 30 days, or submit a s pecial report in accordance with                 !-
: 4. Pressurizer Water Level 2
Specification 6 . 9 .4 .
1 1 ' 2
3 /4 3 - 5 6                                   Amendment       225 SALSM  - UNIT l                                                                                      No .
: 5. Steam Line Pressure 2/Steam 1/Steam 1, 2 Generator Generator
: 6.
Steam Generator Water Level 2/Steam 1/Steam 1, 2 (Narrow Range)
Generator Generator
: 7.
Steam Generator Water Level 4 (1/Steam 3 (1/Steam 1 ' 2 (Wide Range)
Generator)
Generator)
: 8. Refueling Water Storage Tank 2
1, 2 Water Level
: 9. deleted 1 0. Auxiliary Feedwater Flow Rate 4 (1/Steam 3 ( 1/Steam 4, 6 Generator)
Generator) 1 1. Reactor Geelant SystemA a
+
B Margin Monlrof Yoeleted]
1 2. Deleted SALEM - UNIT 1 3/4 3-54 Amendment No. 310
 
TABLE 3.3-1 1 (CONTINUED)
ACCIDENT MONITORING INSTRUMENTATION INSTRUMENT 
1 3. PORV Blook Val'le Position Indicator
: 14. Pressurizer Safety Valve Position lndioator 7Deleted 1 5. Gontainment Pressure Narro*HRange REQUIRED NO. OF CHANNELS 2/valve**
2/valve**
2 1 6. Containment Pressure - Wide Ran}e--1 Deleted 1 2 1 7. Containment Water Level -
2 Wide Range 1 8. Core Exit Thermocouples 4/core quadrant 1 9. Reactor Vessel Level Instrumentation 2
System (RVUS)
: 20. Containment High Range Accident 2
Radiation Monitor 2 1. Main Steamline Discharge (Safety 1/MS Line Valves and Atmospheric Steam Dumps) Monitor Wide Range Neutron Flux Monitors 2
: 23. Auxiliary Feed Water Storage Tank 2
(Condensate Storage Tank) Water Level MINIMUM NO. OF CHANNELS 1 1
2/core quadrant 1
2 1 /MS Line 1
1 ACTION 4-;--2 4-;-4 7, 2 7, 2 1 t 2 8, 9 10 10 1 t 2 1 ' 2 Total number ofls is aonsid to be t\\vo (2) with ono (1 of the ohannol& bemg any one (1) of the follo*Ning alternate means of determining PORV BlocK, or Safety Valve
@ Tailpipe TemperatuFOs for the valves, Pressurizer Relief Tank Temperature Pressurizer Relief Tank Level OPERABb-, 7Delete SALEM - UNIT 1 Amendment No. 310
 
TABLE 3
. 3 - ll (continued)
HOT STANDBY within the next 6 submit a special report in TABLE NOTATION accordance with ecification 6.9.4 ACTION 3 ACTION 4 Wi th the number of OPERABLE accident moni toring channels than the Required Number of Channels shown in Table 3. 3 - 11, res the inoperable channel ( s ) to OPERABLE status wi thin f days, or HM' BHO'l'DOWN within tihe fti!iXtl 12 hG'U:Il'S.
6 With the number of OPERABLE accident monitoring channels less than the MINIMUM Number of Channels shown in Tabl e 3. 3 - ll, restore the inoperable channel ( s ) to OPBRABLE s tatus wi thin or be in HOT SHUTDOWN within the 9 hours.
5 deleted 4following 6 I With the number of OPERABLE channels one les s than the Required Number of Channels shown in Table 3. 3 11, operation may proceed provided that an OPERABLE Steam Generator Wide Range Level channel is avai lable as an al ternate means of indication for the Steam Generator with no OPERATABLE Auxiliary reedwater Plow Rate channel.
ACTION 5 deleted II\\
SALSM UNIT l
; OTHERWISE, restore the inoperable channel to OPERABLE status within 30 days, or submit a special report in accordance with Specification 6.9.4.
3 /4 3 - 5 6 Amendment No. 225


H OT STA N D BY within the following 6                                               next 6 hours and in TABLE 3 . 3 - 1 1 ( continued)
following 6 TABLE 3. 3 - 11 (continued)
TABLE NOTATION ACTION 6     With the n mber of OPERABLE channels less than the Mi .imum Number of channe l s shown in Table 3 . 3 -11 , restore the inope . le OPERABLE s tatus within 7 days , or be in HOT SHUTDOWN within the   hours .
TABLE NOTATION HOT STANDBY within the next 6 hours and in ACTION 6 With the n mber of OPERABLE channels less than the Mi.imum Number of channels shown in Table 3. 3 -11,
channel ( s )
restore the inope. le channel ( s )
ACTION 7   With the number of OPERABLE channels one less than the Required Number of Channels shown in Table 3 . 3 - 1 1 , operation may proceed until the next CHANNEL CALIBRATION (which shall be performed upon the next entry into MODE 5 , COLD SHUTDOWN ) .
OPERABLE status within 7 days, or be in HOT SHUTDOWN within the hours.
ACTION 8   With one RVLI S channel inoperable , restore the RVLIS channel t o OPERABLE s t atus within 3 0 days , or submi t a special report in accordance with Specificat ion 6 . 9 . 4 .
ACTION 7 With the number of OPERABLE channels one less than the Required Number of Channels shown in Table 3. 3 -11,
ACTION 9   With both RVLIS channel s i noperable , restore one channel to OPERABLE s tatus within 7 days or submit a special report in accordance with Spe c i f i ca t i on 6 . 9 . 4 .
operation may proceed until the next CHANNEL CALIBRATION (which shall be performed upon the next entry into MODE 5, COLD SHUTDOWN).
ACTION 1 0 Wi th the number of OPERABLE Channel s less than required by the minimum channels OPERABLE requirements , ini t iate the preplanned al ternate method of moni toring the appropriate parameter within 7 2 hours , and :
ACTION 8 With one RVLIS channel inoperable, restore the RVLIS channel to OPERABLE status within 3 0 days, or submi t a special report in accordance with Specification 6. 9. 4.
: 1) ei ther res t ore the inoperable Channe l ( s )   to OPERABLE status within 7 days of the event , or 2 ) prepare and submit a Special Report to the Commi ssion pursuant to Speci f i cation 6 . 9 . 2 within 14 days f o llowing the event outlining the actions taken , the cause of the inoperabi l i ty and the plans and schedule for res toring the system to OPERABLE s-tatus .
ACTION 9 With both RVLIS channels inoperable, restore one channel to OPERABLE status within 7 days or submit a special report in accordance with Specification 6. 9. 4.
SALEM - UNIT 1                             3 / 4 3 -5 6a                     Amendment No . 2 72
ACTION 1 0 With the number of OPERABLE Channels less than required by the minimum channels OPERABLE requirements,
init iate the preplanned al ternate method of monitoring the appropriate parameter within 72 hours,
and :
1 ) either restore the inoperable Channel ( s ) to OPERABLE status within 7 days of the event, or 2 ) prepare and submit a Special Report to the Commission pursuant to Specification 6. 9. 2 within 14 days following the event outlining the actions taken,
the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE s-tatus.
SALEM -
UNIT 1 3 / 4 3 -5 6a Amendment No. 272


TABLE 4.3-1 1 SURVEILLANCE REQUIREMENTS FOR ACCIDENT MONITORING I NSTRUMENTATION CHANNEL CHANN E L      CHAN N E L        FU NCTIONAL I NSTRUMENT                                     CHECK' 1 l CALIBRATION ' 1l           TEST'1 l
TABLE 4.3-1 1 SURVEILLANCE REQUIREMENTS FOR ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHANNEL CHANNEL FUNCTIONAL I NSTRUMENT CHECK'1l CALIBRATION'1l TEST'1l 1. Reactor Coolant Outlet Temperature -
: 1. Reactor Coolant Outlet Temperature -                                             N.A.
N.A.
THOT (Wide Range)
T HOT (Wide Range)
: 2. Reactor Coolant I nlet Temperature -                                                 N.A.
: 2. Reactor Coolant Inlet Temperature -
TcoLo (Wide Range)
N.A.
: 3. Reactor Coolant Pressure (Wide Range)                                             N.A.
T coLo (Wide Range)
: 4. Pressurizer Water Level                                                           N.A.
: 3. Reactor Coolant Pressure (Wide Range)
: 5. Steam Line Pressure                                                               N .A .
N.A.
: 6. Steam Generator Water Level                                                       N .A.
: 4. Pressurizer Water Level N.A.
: 5.
Steam Line Pressure N.A.
: 6. Steam Generator Water Level N.A.
(Narrow Range)
(Narrow Range)
: 7. Steam Generator Water Level                                                       N.A.
: 7. Steam Generator Water Level N.A.
(Wide Range)
(Wide Range)
: 8. Refueling Water Storage Tank Water Level                                             N.A.
: 8. Refueling Water Storage Tank Water Level N.A.
: 9. deleted 1 0. Auxiliary Feedwater Flow Rate                 S/U#                               N.A.
: 9. deleted 1 0. Auxiliary Feedwater Flow Rate S/U#
N:M                N-A MargiA MeAiter    y 1 1 . Reaeter GeelaAt SysteFA Sl:leeeeliA§ Deleted I
N.A.
#        Auxiliary Feedwater System is used on each startup and flow rate indication is verified at that time.
1 1. Reaeter GeelaAt SysteFA Sl:leeeeliA§ N:M N-A MargiA MeAiter y Deleted I Auxiliary Feedwater System is used on each startup and flow rate indication is verified at that time.
: t.       The iAstevelep RGS suboooling FAar§tn are calf9rated in aeeordance wflh4he Surveillance Frequency Gentrol PrograFA; the FAeniter 'Nill be eeFApared with calculated sl.tBn for knovm input values in accordance with the S:rveilfa.Are Frequency Gentrel Progr-am-:       .-------.
: t.
SALEM - U N IT    1                         3/4 3-57                         Amendment No. 299
The iAstYevelep RGS suboooling FAar§tn are calf9rated in aeeordance wflh4he Surveillance Frequency Gentrol PrograFA; the FAeniter 'Nill be eeFApared with calculated sl.tBZn for knovm input values in accordance with the S[:rveilfa.Are Frequency Gentrel Progr-am-:  


SALEM - UNIT 1 3/4 3-57 Amendment No. 299  


TABLE 4 . 3-1 1 (Continued)
TABLE 4.3-1 1 (Continued)
S URVE ILlANCE REQUIREMENTS FO R ACCI DENT MONITORING lf::IS TRUMENTATION CHANNEL FU NCTIONAL CHECK<1l  CALIBRATIQN<1l CHANNEL I NST R U MENT                                                                                    TEST<1l C HANNEL 1 2 . Deleted         r-lDe leted      I 1 3. PORV Bloo   Position ln4teatef 1 4 . Pressurizer Safety Valvo PosltieA Indicator
SURVEILlANCE REQUIREMENTS FOR ACCIDENT MONITORING lf::ISTRUMENTATION INSTRUMENT
                                      =             .. t-0 e...,.le-
: 12. Deleted r-lDeleted I 1 3. PORV Bloo7Position ln4teatef
                                            -           ed....,....,
: 14. Pressurizer Safety Valvo PosltieA Indicator  
1 5. GGmaemHffiGe 1 6 . Containment Pressure - Wide                                                                 N.A.
=0-e...,..le-t-ed....,...,
1 7. Containment Water Level - Wide Range                                                         N .A.
1 5. GGmaemHfiGe
1 8. Core Exit Thermocouples                                                                       N. A .
: 16. Containment Pressure -
1 9. Reactor Vessel Level Instrumentation                                                         N.A  .
Wide 1 7. Containment Water Level - Wide Range 1 8. Core Exit Thermocouples 1 9. Reactor Vessel Level Instrumentation System (RVLIS)
System (RVLIS)
: 20. Containment High Range Accident Radiation Monitor 21. Main Steamline Discharge (Safety Valves and Atmospheric Steam Dumps)
: 20. Containment High Range Accident Radiati on Monitor 2 1 . Main Steamline Discharge (Safety Valves and Atmospheric Steam Dumps)
Monitor CHANNEL CHANNEL CHECK<1l CALIBRATIQN<1l Table Notation CHANNEL FUNCTIONAL TEST<1l N.A.
Monitor
N.A.
>                                                      Table Notation (1 )     Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.
N.A.
            . R--..,...--I-                                   ----------------------N    -
N.A.
, 2-t  2-. WJd e a n g e N e u t ro n F ux Mo-       n  or-s--
(1 )
it-                                          .A.
Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.
2 3 . Auxiliary Feed Water Storage Tan k (Condensate Storage Tank) Water Level                                                     N .A .
t,2-2-. WJ.deRan-g-eN..,...eut-ro-nFI-uxMo-nit-or-s------------------------N.A-.
Unless the blosk-valve is closed in order to meet the ration b, or o In speGifieation aA
: 23. Auxiliary Feed Water Storage Tank (Condensate Storage Tank) Water Level N.A.
                            '3'     Delete SALEM     - UNIT 1                                     3/4 3-57a                     Amendment No. 3 1 0
Unless the blosk-valve is closed in order to meet the ra8tion b, or o In speGifieation a A
'3' Delete SALEM - UNIT 1 3/4 3-57a Amendment No. 310


AQMINISTRATIVE CONTROLS
AQMINISTRATIVE CONTROLS h.
: h.       The primary to secondary leakage rate observed i n each S G (if it i s not practical to assign the leakage to an individual SG, the entire primary to secondary leakage should be conservatively assumed to be from one SG) during the cycle preceding the inspection which is the subject of the report,
The primary to secondary leakage rate observed i n each SG (if it i s not practical to assign the leakage to an individual SG, the entire primary to secondary leakage should be conservatively assumed to be from one SG) during the cycle preceding the inspection which is the subject of the report,
: i.       The calculated accident induced leakage rate from the portion o f the tubes below 1 5 .21 inches from the top of the tubesheet for the most limiting accident in the most l imiting SG. In addition, if the calculated accident induced leakage rate from the most limiting accident is less than 2 . 1 6 times the maximum operational primary to secondary leakage rate, the report should describe how it was d etermined,
: i.
: j.       The results of monitoring for tube axial displacement (slippage). If slippage is discovered , the implications of the discovery and corrective action shall be provided.
The calculated accident induced leakage rate from the portion of the tubes below 1 5.21 inches from the top of the tubesheet for the most limiting accident in the most limiting SG. In addition, if the calculated accident induced leakage rate from the most limiting accident is less than 2. 1 6 times the maximum operational primary to secondary leakage rate, the report should describe how it was determined,
SPECIAL REPORTS
: j.
: 6. 9.2 Special reports shall be submitted to the U . S . Nuclear Regulatory Commission, Document Control Desk, Washington , D.C. 20555, with a copy to the Administrator, USNRC Region I within the time period specified for each report.
The results of monitoring for tube axial displacement (slippage). If slippage is discovered, the implications of the discovery and corrective action shall be provided.
I nsert: 1 , 4 ,
SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, D.C. 20555, with a copy to the Administrator, USNRC Region I within the time period specified for each report.
6.9.3 D E LETED 6.9.4 When a report is required b y ACTION 8 o r 9 of Table 3.3- 1 1 "Accident Monitoring Instrumentation", a report shall be submitted within the following 1 4 days. The report shall outline the preplan ned alternate method of monitoring for inadequate core cooling, the cause of the inoperabil ity, and the plans and schedule for restoring the instrument channels to OPERABLE status.
6.9.3 DELETED Insert: 1, 4, 6.9.4 When a report is required by ACTION 8 or 9 of Table 3.3-1 1 "Accident Monitoring Instrumentation", a report shall be submitted within the following 1 4 days. The report shall outline the preplan ned alternate method of monitoring for inadequate core cooling, the cause of the inoperability, and the plans and schedule for restoring the instrument channels to OPERABLE status.
SALEM - UNIT 1                                   6-24b                         Amendment No. 303
SALEM - UNIT 1 6-24b Amendment No. 303  


I NSTR U M E NTAT I O N ACCI DENT MONITO R I N G I NSTRU M ENTATION LI M ITI NG CON D ITION FOR OPERATION
I NSTRUMENTATION ACCI DENT MONITORING I NSTRUM ENTATION LIMITING CON DITION FOR OPERATION 3.3.3.7 The accident monitoring instrumentation channels shown in Table 3.3-1 1 shall be operable.
: 3. 3 . 3 . 7 The accident m o n itoring in strumentation channels shown in T a b l e 3 . 3- 1 1 shall be operable.
APPLICABILITY: MODES 1, 2, and 3.  
AP PLICAB I L ITY: MODES         1, 2, and 3.
.----?
ACT I O N :
L---1 ACTION:
: a.     A s shown in Table   3.3-11.
: a.
SURVE I L LANCE REQU I R E M E NTS 4.3.3.7     E ach accident monitoring i nstrumentation channel sh all be demo nstrated OPERABLE by perform ance of the C HA N N E L C H E C K, CHANNEL CAL IBRAT I O N and C HAN NEL FU NCTIONAL TEST ope ratio n s at the freq uencies specified in the Surveill ance F requency C ontrol Prog ram unless otherwise noted i n Table 4. 3-1 1 .
As shown in Table 3.3-11.
      --------------------------------------------- ()li E: -------------------------------------------------------
SURVEILLANCE REQUIREMENTS 4.3.3.7 Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the C HANNEL C HECK, CHANNEL CALIBRATION and C HANNEL FUNCTIONAL TEST operations at the frequencies specified in the Surveillance Frequency Control Program unless otherwise noted in Table 4.3-1 1.  
---------------------------------------------()liE:-------------------------------------------------------
Separate Condition entry is allowed for each Function.
Separate Condition entry is allowed for each Function.
L--- 1 SALEM - UNIT         2                                 3/4 3-50                               Amend ment N o . 282
SALEM - UNIT 2 3/4 3-50 Amendment No. 282  


TABLE 3 . 3-11 ACCIDENT MONITORING INSTRUMENTATION REQUIRED                  MINIMUM NO . OF                    NO . OF INSTRUMENT                                           CHANNELS                  CHANNELS                  ACTION
TABLE 3. 3-11 ACCIDENT MONITORING INSTRUMENTATION INSTRUMENT 1.
: 1. Reactor Coolant Out1et Temperature     -            2                          1                      1,    2 TBo:r (Wide Range}
Reactor Coolant Out1et Temperature TBo:r (Wide Range}
: 2. Reactor Cooant Inet Temperature -                   2                          1                      1,    2 Tcow (Wide Range)
2.
: 3. Reactor Cooant Pressure     {Wide Range)             2                          1                      1,    2
Reactor Cooant Inet Temperature -
: 4. Pressurizer Water Level                               2                        1                      1, 2
Tcow (Wide Range) 3.
: 5. Steam L:ine Pressure                           2 /Steam Generator          1/Steam Generator            1,    2
Reactor Cooant Pressure
: 6. Steam Generator Water Level     (Narrow       2/Steam Generator          1/Steam Generator            1,    2 Range)
{Wide Range) 4.
: 7. S team Generator Water Level     (Wide         4 (1 /Steam Generator)    3  {1/Steam Generator)      1,    2      I Range)
Pressurizer Water Level 5.
: 8. Refueling Water Storage Tank Water             2                          1                            1,    2 Level
Steam L:ine Pressure 6.
: 9. deeted 10 . Auxi1.iary Feedwater Fl ow Rate                 4  (1 /Stett Generator)  3  (1/ Steam Generator)      4,    6 Syst:em Subeeol::ing 11 . RGacter Csal;mt:
Steam Generator Water Level (Narrow Range) 7.
      &#xa5;oar&#xa2;a   }aer    .          y Deleted   I 2
Steam Generator Water Level (Wide Range) 8.
12 . PORV Pes1ea
Refueling Water Storage Tank Water Level 9.
* Iadi:eai;;ez                                               +/-                            J__,- 2 SALEM   - UNIT 2 Yoeleted     I                   3 / 4 3-51                               Amendment No . 206
deeted 10. Auxi1.iary Feedwater Flow Rate 11.
12.
RGacter Csal;mt: Syst:em Subeeol:ing
&#xa5;oar&#xa2;a }1a23er y Deleted I PORV Pes1ea
* Iadi:eai;;ez SALEM  
- UNIT 2 Yoeleted I REQUIRED NO.
OF CHANNELS 2
2 2
2 2/Steam Generator 2/Steam Generator 4 (1 /Steam Generator) 2 4
(1/Stett Generator) 2 3/4 3-51 MINIMUM NO. OF CHANNELS ACTION 1
1, 2 1
1, 2 1
1, 2 1
1, 2 1/Steam Generator 1, 2 1/Steam Generator 1, 2 3
{1/Steam Generator) 1, 2 1
1, 2 3
(1/Steam Generator) 4, 6
+/-
J__,- 2 Amendment No. 206 I


TABLE   3 . 3- 1 1 ( C ontinued)
TABLE 3. 3-1 1
ACCI DENT MONITORING         INSTRUMENTATI ON REQUIRED                        MINIMUM NO . OF                      NO . OF rNsTRu""MENT                                               Deleted                               ]                                                         CH.llliN ELS                    CHANNELS                    ACT IOK 13 . PORV Block 'lalvc Position Indicator                                                                                                               2/valve**                            3..                     ;)_
( Continued)
                                                                                                                                                                                                                                .2 14 . n ....- ...... _r:::w--"'!'_.._.., .; --...-. -
ACCIDENT MONITORING INSTRUMENTATION rNsTRu""MENT  
          ;:::::==::t "==*= L::r=wz                        co-.__,f;_.......  \7......, 1 ... ,..,-, n ..... ,.... .; +- ..:: ......
,Deleted ]
2/ v al oe**                        +/-                      -+/--, i!
13.
          ""'""""'"' ca "0"'
PORV Block 'lalvc Position Indicator 14.
Deleted               I 15 . Containment Pressure                                                                         Narre-;; Ral:'!:&sect;'C                                   2                                   +/-                      h Deleted               I 2                                    1                             2
n....-..... _r:::w--"'!'_.. _.., W.; X--...-.-
: 16.      Cont ainment                                        P r e s s ure                    - Wide Range                                                                                                               7 r 17 . ContaiPent Water Level                                                                           - Wide                   Range                 2                                     1                       7,    2
co-.
: 18.     Core Exit Thermocoup les                                                                                                                            4 / core    quadrant                  2 / core quadrant       1,    2 19 . Reactor Ve s s el                                             Level                 I n s t rumentation                                          2                                    1                      8,    9 Sys t em                               { RVL I S )
__,f;_..... T UV \\7.....,1...,..,-,
: 20.     Containment High Range Accident                                                                                                                     2                                    2                      10 Radi a ti on Monitor
n....,....; +-..::..........  
: 21.      Main Steamline                                                                                                                                     1/                                  1/                      10 Di s ch a rge                                    ( Safety                                                                                     MS  Line                            M S Line Dump s )
-;:::::.==/::t "==*= L0::r=wz
Va lves                                and Atmo spheri c S t e am                                                Monitor
""'"""'"' ca "0"'
      /
Deleted I REQUIRED NO.
f-+/-.:L)                                                                                                                                                                                                                               .F V .L
OF CH.llliNELS 2/valve**
                                                                                                                                                                                                                                          +-. 1 I.... .I. J.C Ter.:tpe+/-:atures Delet e I SALEM - UNIT 2                                                                                                                                                               3/4 3-5 1 a                           Amendment No . 263
2/ valoe**
: 22. Wid e Range Neutron Flux Monitors                                                                                                                          2                                   1                           1, 2 Auxil iary Feed Water Stora g e Tank                                                                                                                    2                                   1                           1, 2
15.
( Condensate Storage Tank) Water Level
Containment Pressure Narre-;; Ral:'!:&sect;'C 2  
Deleted I 1 6.
Containment Pressure -
Wide Range 2
17.
ContaiPent Water Level Wide Range 2
1 8. Core Exit Thermocouples 4 /core quadrant 19.
Reactor Ve s s el Level Instrumentation Sys t em
{RVL I S )
2 0. Containment High Range Accident Radi ati on Monitor 2 1.
Main Steamline Di s charge
( Safety Valves and Atmospheri c Steam Dumps ) Monitor  
/
f-+/-.:L)
Ter.:tpe+/-:atures SALEM -
UNIT 2
: 22. Wide Range Neutron Flux Monitors Auxiliary Feed Water Storage Tank (Condensate Storage Tank) Water Level 2
2 1/
M S Line 2
2 3 / 4 3-5 1 a MINIMUM NO.
OF CHANNELS 3..  
+/-
+/-
1 1
2 / core quadrant 1
2 1 /
M S Line
Delete I 1
1 ACT IOK
;)_
.2 S,
A
-+/--, i!
h 7 r 2 7, 2 1, 2 8, 9 1 0 1 0
.F
+-. 1 V.L I....I.J.C Amendment No.
2 6 3 1, 2 1, 2  


HOT STA N D BY within the next 6 hours and in                                       submit a s pecial report in accorda nce with Specification 6 . 9 .4 TABLE 3 . 3 - l l  ( continued)
HOT STANDBY within the next 6 hours and in submit a special report in accordance with Specification 6.9.4 ACTION 1 ACTION 2 ACTION 3 ACTION 4 ACTION 5 Wi th the n
* TABLE NOTATION Wi th the  n  ber Channels                 able ACTION 1                            o f OPERABLE accident moni torin          channels less 3 . 3-11 ,
than the Re restore the days,
ohannel ( s ) to than the Re    1ired Number of                    shown in res tore the    noperable                          OPERABLE    .atus within +
or TABLE 3. 3-ll
days ,  or
( continued)
* n the nexe 2 ACTION 2      Wi th the numb 1r of OPERABLE acci dent moni toring channels less
TABLE NOTATION ber of OPERABLE accident monitorin 1ired Number of Channels shown in noperable ohannel (s) to OPERABLE
                                  \un Number of Channels shown in Table 3 .           3-11 ,
* n the nexe 2 channels less able 3. 3-11,  
ohannel ( s )
.atus within +  
than the Mini re s tore the i    parable                      to OPERABLE status within 48 or be in HOT SHU'I'DOWN within the M-&t:l 71-2 hours .
+
                                          .                          t.                        - If\d r:::;-;- ---,
Wi th the numb 1r of OPERABLE accident monitoring channels less than the Mini \\un Number of Channels shown in Table 3. 3-11,
ACTION 3      deleted                                               Yfollowing 6 I ACTION 4      Wi th the number of OPERABLE channels one
res tore the i parable ohannel (s) to OPERABLE status within 48 or be in HOT SHU'I'DOWN within the M-&t:l 71-2 hours.
* l e s s than the Required Number of Channel s        shown in Table 3 . 3- 1 1 , operations may proceed provided that an OPERABLE Steam Generator Wide Range Level channel is availabl e as an alternate means of indication for the Steam Generator wi th no OPERABLE Auxiliary Feedwater Flow Rate Channe l ACTION 5      de leted
If\\ r:::;-;--,
                                  ; OTH E RWI SE, restore the inoperable channel to OPERABLE status within 30 days, or submit a special report i n accordance with
d t.
* S pecification 6 . 9 .4 .
- 
* SALEM -  UNI'l' 2                             3/4   3 - 5 1b            Amendment No . 206
deleted Yfollowing 6 I With the number of OPERABLE channels one
* less than the Required Number of Channels shown in Table 3. 3-11,
operations may proceed provided that an OPERABLE Steam Generator Wide Range Level channel is available as an alternate means of indication for the Steam Generator wi th no OPERABLE Auxiliary Feedwater Flow Rate Channel deleted
; OTHERWISE, restore the inoperable channel to OPERABLE status within 30 days, or submit a special report in accordance with Specification 6.9.4.
SALEM UNI'l' 2 3/4 3-51b Amendment No. 206  


3 . 3 -11 ( continued )   H OT STAN DBY within the next 6 hours and in TABLE following 6 TABLE NOTATION 6      With the n ber o f OPERABLE channel s less than the Mi ti.m:um Number of channels 3hown in Table 3 . 3 -11 , restore the inope le ACTION channel ( s ) t OPERABLE s tatus within 7 days , or be in HOT SHUTDOWN within the amet 12 hours .
ACTION 6 ACTION 7 ACTION 8
ACTION 7 Number o f Channel s shown in Table 3 . 3 -11 , opera t i on may proceed With the number o f OPERABLE channels one l e s s than the Required unt i l the next CHANNEL CALIBRATION (which sha l l be performed upon the next entry into MODE 5 , COLD SHUTDOWN ) .
*ACTION 9.
ACTION 8      With one RVLIS channel inoperable , restore the RVLI S channel t o accordance with Specif icat ion 6 . 9 . 4 .
ACTION 1 0 TABLE 3. 3-11 ( continued )
OPERABLE s ta tus wi thin 30 days , or submit a spe c i a l report i n
following 6 TABLE NOTATION HOT STANDBY within the next 6 hours and in With the n ber of OPERABLE channel s less than the Mi ti.m:um Number of channels 3hown in Table 3. 3 -11,
*ACTION 9 . With both RVLI S channels inoperable , restore one channel to OPERABLE s tatus within 7 days or submi t a special report in accordance with Spec i f ication 6 . 9 . 4 .
restore the inopele channel ( s ) t OPERABLE status within 7 days,
ACTION 1 0      With the number o f O PERABLE Channels less than required b&#xa5; the minimum channels O PERABLE requirements , ini t i ate the preplanned alternat e method of monit oring the appropriate parame t er within 72 hours , and :
or be in HOT SHUTDOWN within the amet 12 hours.
: 1)   either restore the inoperable Channel ( s )   to OPERABLE status within 7 days of the event , or 2 ) prepare and submit a Special Report to the Commi s s ion pursuant to Spec i fication 6 . 9 . 2 within 14 days following the event outlining the actions taken , the cause of the inoperability and the plans and schedule for res toring the sys t em to OPERABLE status .
With the number o f OPERABLE channels one less than the Required Number o f Channels shown in Table 3. 3-11, operation may proceed unti l the next CHANNEL CALIBRATION (which shall be performed upon the next entry into MODE 5, COLD SHUTDOWN).
SALEM - UNIT     2                   3 / 4 3 -Slc                 Amendment No . 253
With one RVLIS channel inoperable,
restore the RVLIS channel to OPERABLE s ta tus within 30 days,
or submit a special report in accordance with Specification 6. 9. 4.
With both RVLIS channels inoperable,
restore one channel to OPERABLE s tatus within 7 days or submit a special report in accordance with Specification 6. 9. 4.
With the number of OPERABLE Channels less than required b&#xa5; the minimum channels OPERABLE requirements,
initiate the preplanned alternat e method of monitoring the appropriate parameter within 72 hours,
and :
1 ) either restore the inoperable Channel ( s ) to OPERABLE status within 7 days of the event,
or 2 ) prepare and submit a Special Report to the Commis s ion pursuant to Speci fication 6. 9. 2 within 14 days following the event outlining the actions taken,
the cause of the inoperability and the plans and schedule for res toring the sys tem to OPERABLE status.
SALEM -
UNIT 2 3 /4 3-Slc Amendment No. 253  


TABLE 4. 3-1 1 SURVEI LLANCE REQ UIREMENTS FOR ACCI DENT MONITORI NG I NSTRUMENTATI ON CHANNEL CHANNEL       CHA N N E L      FUNCTIONAL I N STRUMENT                                                CHECKS<1l   CALI BRATIQN<1l       TEST<1l 1 . Reactor Coolant Outlet Temperature -                                                         N.A.
TABLE 4.3-1 1 SURVEILLANCE REQUIREMENTS FOR ACCIDENT MONITORI NG INSTRUMENTATION CHANNEL CHANNEL CHANNEL FUNCTIONAL INSTRUMENT CHECKS<1l CALIBRATIQN<1l TEST<1l 1. Reactor Coolant Outlet Temperature -
T Hor (Wide Range)
N.A.
: 2. Reactor Coolant Inlet Temperature -
THor (Wide Range)
Tc o LD (Wide Range)
: 2.
Reactor Coolant Inlet Temperature -
N.A.
T coLD (Wide Range)
: 3. Reactor Coolant Pressure (Wide Range)
N.A.
N.A.
: 3. Reactor Coolant Pressure (Wide Range)                                                        N.A.
: 4. Pressurizer Water Level N.A.
: 4. Pressurizer Water Level                                                                       N.A.
: 5. Steam Line Pressure N.A.
: 5. Steam Line Pressure                                                                           N.A.
: 6. Steam Generator Water Level N.A.
: 6. Steam Generator Water Level                                                               N.A.
(Narrow Range)
(Narrow Range)
: 7. Stea m Generator Water Level                                                               N.A.
: 7. Steam Generator Water Level N.A.
(Wide Range)
(Wide Range)
: 8. Refueling Water Storage Tank Water Level                                                     N.A.
: 8. Refueling Water Storage Tank Water Level N.A.
: 9. deleted 1 0. Auxiliary Feedwater Flow Rate                               S/U#                           N.A.
: 9. deleted 1 0. Auxiliary Feedwater Flow Rate S/U#
: 11. ReaetoF Gaalant System S:eeoolin&sect;                                                       N.,.A; Maf&sect;j                           Deleted
N.A.
#        Auxiliary Feedwater System is used on each startup and flow rate indication is verified at that time.
1 1. ReaetoF Gaalant System Sc:eeoolin&sect; N.,.A; Maf&sect;j 
        =Fhe-iftStfl:tments used to develop RGS s:beoolin&sect; maf&sect;in are ealibrateeHA-aeeeffia.A.ee.
Deleted Auxiliary Feedwater System is used on each startup and flow rate indication is verified at that time.  
          'v'll'ith the S:rveillance Freq:ency Gontrol Pro&sect;ram; the monitor will be compared with calculated subcooling margin for knovm input values in accordance with the Surveillance Frequency Gontrol Pro&sect;ram 1'\:
=Fhe-iftStfl:tments used to develop RGS sd:beoolin&sect; maf&sect;in are ealibrateeHA-aeeefia.A.e.  
SALEM - U N IT 2                                       3/4 3-52                       Amendment No. 282
'v'll'ith the Sc:rveillance Freqe:ency Gontrol Pro&sect;ram; the monitor will be compared with calculated subcooling margin for knovm input values in accordance with the Surveillance Frequency Gontrol Pro&sect;ramb 1'\\:  

SALEM - UNIT 2 3/4 3-52 Amendment No. 282  


TABLE 4.3-1 1 (Contin ued)
TABLE 4.3-1 1 (Continued)
S U RVE I L LANCE REQU IREME NTS FOR ACCI DENT MONITORING I NSTRUMENTATION CHAN NEL CHANNEL       C HAN N EL    FUNCTIONAL I NSTRUMENT                                                    CH ECKS(1J CALIBRATION(1l         TEST(1l 12*                   ..      .                                                N.A Deleted 1 3 . PORV Block Valve Position Indicator                                       N.A                  !
SURVEILLANCE REQUIREMENTS FOR ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHANNEL CHANNEL FUNCTIONAL INSTRUMENT CHECKS(1J CALIBRATION(1l TEST(1l 1 2 *
1 4. Pressurizer Safety Valve Position Indicator                                 N.A r::-e-=-le-:t-1 5 . Gofifia nment Pressure Narro*N Rafl&sect;-0                                                       N-A
*Deleted 1 3. PORV Block Valve Position Indicator 1 4. Pressurizer Safety Valve Position Indicator 1 5. Goffiainment Pressure Narro*N Rafl&sect;-0 !Ljr::o-e-=-le-:t-ed-:-o  
                                                            !Lj  o           ed-:-o
)Deleted I 1 6. Containment Pressure - Wide Range 1 7. Containment Water Level - Wide Range 1 8. Core Exit Thermocouples 1 9. Reactor Vessel Level Instrumentation System (RVLIS)
                        .                                Deleted         I 1 6.       Containment Pressure - Wide Range                                                       N .A.
: 20. Containment High Range Accident Radiation monitor 21. Main Steamline Discharge (Safety Valves and Atmospheric Steam Dumps) Monitor Table Notation N.A N.A N.A N-A N.A.
1 7.       Containment Water Level - Wide Range                                                   N .A.
N.A.
1 8.       Core Exit Thermocouples                                                                 N .A.
N.A.
1 9.       Reactor Vessel Level I nstrumentation                                                  N.A.
N.A.
System (RVLIS)
(1 )
: 20. Contain ment H igh Range Accident Radiation monitor 21 . Main Steam line Discharge (Safety Valves and Atmospheric Steam Dumps) Monitor Table Notation (1 )           Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table .
Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.
      !              I:Jnless the bloek valve is closed in order to meet the requirements of ,l\otion b, or o in
I:Jnless the bloek valve is closed in order to meet the requirements of,l\\otion b, or o in  
.-:-:-:--="'==--. specificatioo-.a-A,&sect;..;
.-:-:-:--="'=-.
fLj Delete 122. Wide Range N eutron Flux Monitors                                                           N .A.
specificatioo-.a-A,&sect;..;
: 23. Auxiliary Feed Water Storage Tank (Condensate Storage Tank) Water Level                                               N A.
fLj Delete 122. Wide Range Neutron Flux Monitors
SALEM - U N IT 2                                       314 3-52a                       Amendment No.     282
: 23. Auxiliary Feed Water Storage Tank (Condensate Storage Tank) Water Level SALEM - UNIT 2 314 3-52a N.A.
N A.
Amendment No. 282  


ADM I NISTRATIVE CONTROLS
ADMINISTRATIVE CONTROLS
: e.     Number of t ubes plugged d u ring the inspection outage for each deg radation mechanism ,
: e.
: f. The n u m ber and percentage of tubes plugged to date, and the effective plugg ing percentage in each steam generator,
Number of tubes plugged during the inspection outage for each deg radation mechanism,
: g. The results of condition monitoring, including t h e results o f tube p ulls and in-situ testing.
: f.
SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission , Document Control Desk, Wash ington, D.C. 20555, with a copy to the Administrator, USNRC Region I within the time period specified for each report.
The number and percentage of tubes plugged to date, and the effective plugging percentage in each steam generator,
I nsert: 1 , 4, 6.9.3 DELETED 6 . 9.4 When a report is required by ACTION 8 OR 9 of Table 3.3- 1 1 "Accident Monitoring Instrumentation", a report shall be submitted with in the following 1 4 days. The report shall outline the preplan ned alternate method of monitoring for inadequate core cooling , the cause of the inoperability, and the plans and schedule for restoring the instrument channels to OPERABLE status.
: g.
SALEM - U NIT 2                               6-24b                               Amendment No. 291
The results of condition monitoring, including the results of tube pulls and in-situ testing.
SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Document Control Desk, Wash ington, D.C. 20555, with a copy to the Administrator, USNRC Region I within the time period specified for each report.
6.9.3 DELETED Insert: 1, 4, 6.9.4 When a report is required by ACTION 8 OR 9 of Table 3.3-1 1 "Accident Monitoring Instrumentation", a report shall be submitted with in the following 14 days. The report shall outline the preplan ned alternate method of monitoring for inadequate core cooling, the cause of the inoperability, and the plans and schedule for restoring the instrument channels to OPERABLE status.
SALEM - UNIT 2 6-24b Amendment No. 291  


LR-N16-0003                                                                     LAR S16-01 Attachment 3 M ark-up of Proposed Technical Specification Bases Pages The following Tech nical Specifications pages for Renewed Facility Operating License DPR-70 are affected by this change req uest:
LR-N16-0003 LAR S16-01 Mark-up of Proposed Technical Specification Bases Pages The following Technical Specifications pages for Renewed Facility Operating License DPR-70 are affected by this change request:
Technical Specification Bases 3/4 . 3 . 3 . 7, Accident Mon itoring I nstrumentation               B 3/4 3-3 The following Technical Specifications pages for Renewed Facil ity Operating License DPR-75 are affected by this change req uest:
Technical Specification Bases 3/4.3.3.7, Accident Monitoring I nstrumentation B 3/4 3-3 The following Technical Specifications pages for Renewed Facility Operating License DPR-75 are affected by this change request:
Technical Specification Bases 3/4 . 3 . 3 . 7, Accident Monitoring I nstrumentation               B 3/4 3-3 1
Technical Specification Bases 3/4.3.3. 7, Accident Monitoring I nstrumentation B 3/4 3-3 1  


BAS E S 3/4 . 3 . 3 . 5   REMOTE       S H UT DOWN     I N S TRUMENTAT I ON The   OPERAB I L I T Y      of   the   remo t e      shut down       i n s t rume n t a t i o n    e n s ur e s   that sufficient          capab i l i t y i s       ava i l a b l e    t o p e rmi t     shut down       a n d ma i n t e nance      of   HOT S TAN DBY      of   the     facility        f r om l o c a t i on s     ou t s i d e    of   the     control       room . This 19          10          50 .
BASE S 3/ 4. 3. 3. 5 REMOTE SHUT DOWN INSTRUMENTAT I ON The OPERAB I L ITY of the remote shut down i n s t rumentat ion ensur e s that s u f f i c ient capab i l i t y i s ava i lable t o permi t shut down and maintenance of HOT S TANDBY of the f a c i l i t y from loca t i ons out s ide of the control room.
c ap a b i l i t y is    r e qu i r e d  in    the     event control             r o om h ab i t ab i l i t y     is    lost   and is cons i s tent with General                   D e s i gn   Criteria             of        C FR 3/4 . 3 . 3 . 6     THIS      S E C T I ON   DELETE D 3/4 . 3 . 3 . 7     AC C I DENT     MON I T O R I N G    I N S TRUMENTAT I ON The   OPERAB I L I T Y      of   the   a c c i de n t mon i t o r i n g      i n s t rume n t a t i o n    ensures        that sufficient          i n forma t i o n      is   ava i l ab l e     on   s e l e c t e d p l ant     p a rame t e r s     to mon i t o r 1 . 97 ,
This c apabi l i t y i s required i n the event control room hab i t ab i l i t y i s lost and is cons i s tent with General D e s i gn Criteria 19 o f 10 C FR 50.
and   assess      the s e     v a r i ab l e s   f o l l ow i n g  an   a c c i de n t . This       capab i l i t y i s cons i s t ent     with     the   R e c omme n da t i o n s     o f Re gul a t o r y G u i d e                  " I n s trume n t a t i on 1975 .
3 / 4. 3. 3. 6 T H I S SECT I ON DELETE D 3 / 4. 3. 3. 7 ACC I DENT MON I TORING INSTRUMENTAT ION The OPERAB I L ITY of the acc i dent moni t oring ins trumentat i on ensures that s u f f i c i e n t in format ion is ava i l ab l e on s e l e cted pl ant parame t e r s to monitor and a s s e s s the s e variab l e s f o l l owing an accident.
for   L i gh t -Wa t e r - C o o l e d Nuc l e a r        Powe r     Plants       to A s s e s s     P l ant   Condi t i o n s  During and   Fo l l ow i n g    an A c c i d e n t , "     D e c emb e r 3                   RAD I OA C T I VE      L I QU I D E FFLUENT         MON I TORING         I N S TRUMENTAT I ON urpo s e     of   t an k   l evel     indicating         devi c e s     is  to   a s s ure     the   detect ion     and contr     l   of   leaks       that     if   not   c ont r o l l e d c ou l d p o t e n t i a l l y       result       in the t ra n s p  rt   of   r a d i o a ct i ve ma t e r i a l s     to   UNRE S T R I C T E D AREAS .
This capab i l i t y i s cons i s t ent with the Recommenda t i o n s o f Re gul atory Guide 1. 97, " I ns trumentation for L i ght-Water-Cooled Nuclear Powe r Plants to A s s e s s Pl ant Condi tions During and Fo l l owing an Accident, " December 1 9 7 5.
The Wide Range Neutron F lux Mon itors are the Gamma-Metrics Post-Accident Neutron Monitors.
3 RAD I OACTIVE L I QU I D E FFLUENT MON I TORING INSTRUMENTAT I ON urpo s e of t an k l evel indicating devi c e s i s to a s s ure the detect ion and contr l of leaks that if not cont r o l l e d could potent i a l l y result in the t ransp rt of radioact ive ma t e r i a l s to UNRESTRICTED AREAS.
THIS    S E C T I ON D E L E T E D 3/4 . 3 . 3 . 10 THIS    S E CT I ON   DELETED 3/4 . 3 . 3 . 11 THIS    S E CT I ON   DELETED 3/4 . 3 . 3 . 12 THI S   S E C T I ON  DELETED 3/4 . 3 . 3 . 13 TH I S S E C T I ON   DELETED SALEM     -   UN I T   1                               B   3/4 3-3                                               Ame ndme n t    No . 282
The Wide Range Neutron Flux Monitors are the Gamma-Metrics Post-Accident Neutron Monitors.
( PSEG I s sued)
T H I S SECT I ON DELETED 3 / 4. 3. 3. 1 0 T H I S SECT I ON DELETED 3 / 4. 3. 3. 1 1 T H I S SECT I ON DELETED 3 / 4. 3. 3. 12 THI S SECTION DELETED 3 / 4. 3. 3. 1 3 TH I S SECT I ON DELETED SALEM -
UN I T 1 B 3 / 4 3-3 Amendment No. 282
( PSEG I s sued)  


I N STRU M E NTATI ON BASES I mmediate action(s) , in accordance with the LCO Action Statements, means that the req uired action should be pursued without delay and in a controlled man ner.
INSTRUMENTATION BASES I mmediate action(s), in accordance with the LCO Action Statements, means that the required action should be pursued without delay and in a controlled manner.
3/4 . 3 . 3 . 2 THIS SECTI ON DE LETED 3/4 . 3 . 3 . 3 THIS SECTI ON D E LETED 3/4 . 3 . 3 .4 THIS S ECTI ON DE LETED 3/4 . 3 . 3 . 5 REMOTE SHUTDOWN I NSTRUME NTATI ON The OPERAB I L ITY of the remote sh utdown instrumentation ensures that sufficient capabil ity is available to permit sh utdown and mai ntenance of HOT STANDBY of the facility from locations outside of the control room . This capabi l ity i s req u i red in the event control room habitability is lost and is consistent with General Design Criterion 1 9 of 1 0 CFR 50.
3/4.3.3.2 THIS SECTION DELETED 3/4.3.3.3 THIS SECTION DELETED 3/4.3.3.4 THIS SECTION DELETED 3/4.3.3.5 REMOTE SHUTDOWN I NSTRUMENTATION The OPERABILITY of the remote shutdown instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of HOT STANDBY of the facility from locations outside of the control room. This capability is required in the event control room habitability is lost and is consistent with General Design Criterion 1 9 of 1 0 CFR 50.
3/4 . 3 . 3 . 6 TH I S S ECTI ON DE LETED 3/4 . 3 . 3 . 7 ACC I D ENT M O N ITORING I NSTRUME NTATION The OPERABI LITY of the accident monitoring instru mentation ensures that s ufficient i nformation is available on selected plant parameters to monitor and assess these variables following an accident. This capability is consistent with the Recom mendations of Regu latory Guide 1 . 97, " I nstrumentation for Light-Water-Cooled N u clear Power Plants to Assess Plant Cond itions During and Following an Accident," December 1 975 and N UREG-0578 , "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recom mendations."
3/4.3.3.6 THIS SECTION DELETED 3/4.3.3.7 ACCIDENT MONITORING I NSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables following an accident. This capability is consistent with the Recommendations of Regulatory Guide 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," December 1 975 and N UREG-0578,"TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations."  
    \\...__ The Wide Range N eutron F lux Monitors are the Gamma-Metrics Post-Accident N eutron Monitors .
\\\\..._The Wide Range Neutron Flux Monitors are the Gamma-Metrics Post-Accident Neutron Monitors.
SALEM - U N I T 2                             B 3/4 3-3                           Amendment No. 265 (PSEG I ssued)}}
SALEM - UNIT 2 B 3/4 3-3 Amendment No. 265 (PSEG Issued)}}

Latest revision as of 19:02, 9 January 2025

License Amendment Request to Amend the Accident Monitoring Instrumentation Technical Specifications
ML16323A279
Person / Time
Site: Salem  PSEG icon.png
Issue date: 11/17/2016
From: Mcfeaters C
Public Service Enterprise Group
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
LAR S1601, LR-N16-0003
Download: ML16323A279 (43)
v • d • e


Text

PSEG Nuclear LLC P.O.Box 236, Hancoclw Bridge, New Jersey 08038-0236 PSEG NudearLLC NOV 17 2016.

10 CFR 50.90 LR-N16-0003 LAR S16-01 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Salem Generating Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-70 and DPR-75 NRC Docket Nos. 50-272 and 50-311 License Amendment Request to Amend the Accident Monitoring Instrumentation Technical Specifications In accordance with the provisions of 10 CFR 50.90, PSEG Nuclear LLC (PSEG) is submitting a request for an amendment to the Technical Specifications (TS) for Salem Generating Station (Salem) Units 1 and 2.

The proposed amendment revises the TS 3/4.3.3.7, Accident Monitoring Instrumentation.

Specifically, this change modifies the list of instruments required to be operable based on implementation of Regulatory Guide (RG) 1.97 for Salem, and revises the allowed outage times and Actions for inoperable channels to be consistent with NUREG-1431, Revision 4, "Standard Technical Specifications - Westinghouse Plants." provides an evaluation supporting the proposed changes. Attachment 2 provides the existing TS pages marked up to show the proposed changes. Attachment 3 provides existing TS Bases pages marked up to show the proposed changes and are being provided for information only.

PSEG requests approval of this license amendment request (LAR) in accordance with standard NRC approval process and schedule. Once approved, the amendment will be implemented within 60 days from the date of issuance.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated State of New Jersey Official.

There are no regulatory commitments contained in this letter.

If you have any questions or require additional information, please contact Ms. Tanya Timberman at 856-339-1426.

LR-N16-0003 Page 2 I declare under penalty of perjury that the foregoing is true and correct.

Executed on __ t_t _/--.,.\\ _7_/_1 :=--6 (Date)

Respectfully, Charles V. McFeaters Site Vice President Salem Generating Station Attachments:

1. Evaluation of Proposed Changes
2. Mark-up of Proposed Technical Specification Pages
3. Mark-up of Proposed Technical Specification Bases Pages cc:

Mr. D. Dorman, Administrator, Region I, NRC Ms. C. Parker, Project Manager, NRC NRC Senior Resident Inspector, Salem Mr. P. Mulligan, Chief, NJBNE PSEG Corporate Commitment Tracking Coordinator Salem Commitment Tracking Coordinator 10 CFR 50.90

LR-N16-0003 LAR 516-01 Evaluation of Proposed Changes Table of Contents

1.0 DESCRIPTION

................................................................................................................ 1

2.0 PROPOSED CHANGE

.................................................................................................... 1

3.0 BACKGROUND

............................................................................................................... 2 4.0 TECHNICAL ANAL YSIS

.................................................................................................. 4 4.1 Methodology............................................................................................................ 4 4.2 Technical Evaluation

................................................................................................ 5 4.3 Individual Instrument Evaluations............................................................................ 5 4.4 LCO Times and Separate Condition Entry............................................................. 14

5.0 REGULATORY ANALYSIS

........................................................................................... 16 5.1 No Significant Hazards Consideration................................................................... 16 5.2 Applicable Regulatory Requirements/Criteria

......................................................... 18

6.0 ENVIRONMENTAL CONSIDERATION

......................................................................... 19

7.0 REFERENCES

.............................................................................................................. 19

LR-N16-0003

1.0 DESCRIPTION

LAR S16-01 The proposed amendment revises the TS 3/4.3.3.7, Accident Monitoring Instrumentation.

Specifically, this change modifies the list of instruments required to be operable based on implementation of Regulatory Guide (RG) 1.97 for Salem, and revises the allowed outage times and Actions for inoperable channels to be consistent with NUREG-1 431, Revision 4, "Standard Technical Specifications - Westinghouse Plants."

Changes are proposed to Technical Specification (TS) Table 3.3-1 1, to Actions 1, 2, 4 and 6, and to TS Section 6.9.4. In addition, consistent with NUREG-1 431 a note is added to Section 3.3.3. 7 allowing separate condition entry for each function (i.e., affected instrument).

2.0 PROPOSED CHANGE

This License Amendment Request revises Salem Units 1 and 2 TS Table 3.3-1 1 and Table 4.3-1 1, Accident Monitoring Instrumentation as follows:

Instruments removed from TS Table 3.3-1 1 and TS Table 4.3-1 1 because they are not regulatory guide (RG) 1.97 Type A, or Category 1 :

Reactor Coolant System Sub-Cooling Margin Monitor*

PORV Position Indicator (Unit 2 only)*

PORV Block Valve Position Indicator*

Pressurizer Safety Valve Position Indicator*

Containment Pressure (Narrow Range)

  • The associated footnotes are also removed.

Instruments added to TS Table 3.3-1 1 and TS Table 4.3-1 1 because they are RG 1.97 Type A, or Category 1 instruments that are not currently in the tables:

Wide Range Neutron Flux Monitors Auxiliary Feed Water (AFW) Storage Tank (Condensate Storage Tank) Water Level The following TS Table 3.3-1 1 Actions are revised:

Action 1 With the number of OPERABLE accident monitoring channels less than the Required Number of Channels shown in Table 3.3-1 1, restore the inoperable channel(s) to OPERABLE status within 30 days, or submit a special report in accordance with Specification 6.9.4.

Action 2 With the number of OPERABLE accident monitoring channels less than the Minimum Number of Channels shown in Table 3.3-1 1, restore the inoperable channel(s) to OPERABLE status within 7 days 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 />.

1

LR-N16-0003 Action 4 LAR 516-01 With the number of OPERABLE channels one less than the Required Number of channels shown in Table 3.3-11, operation may proceed provided that an OPERABLE Steam Generator Wide Range level channel is available as an alternate means of indication for the Steam Generator with no OPERABLE Auxiliary Feedwater Flow Rate channel: OTHERWISE, restore the inoperable channel to OPERABLE status within 30 days, or submit a special report in accordance with Specification 6.9.4.

Action 6 With the number of OPERABLE channels less than the Minimum Number of channels shown in Table 3.3-11, restore the inoperable channel(s) to OPERABLE status within 7 days, 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 />.

TS 3.3.3.7 is revised to add a note which allows for a Separate Condition entry for each instrument.

The surveillance frequencies for Line Items 22 and 23 that are being added to TS Table 4.3-11 will be contained in the Surveillance Frequency Control Program with the exception of the Channel Functional Test which is not applicable for these instruments. Salem has implemented TSTF Change Traveler TSTF-425, Revision 3 (Reference 11) as approved by the NRC with TS Amendments 299 and 282 (Reference 12} for Units 1 and 2 respectively. The initial surveillance frequency will be 31-days for the CHANNEL CHECK and 18-months for the CHANNEL CALIBRATION consistent with NUREG-1431, Standard Technical Specifications for Westinghouse Plants.

The criteria for relocation of a surveillance frequency to a licensee controlled program in accordance with TSTF-425 were reviewed. These surveillance frequencies are periodic surveillances that: 1) do not reference other approved programs for the specified interval, 2) are not event driven, 3) do not have a time component based on event occurrence, and 4) are not related to a specific condition for performance. Therefore the periodic surveillance frequencies are within the scope of TSTF-425 for location in the licensee controlled Surveillance Frequency Control Program.

Finally, TS 6.9.4, Special Reports, is revised to include Actions 1 and 4 of TS Table 3.3-11.

3.0 BACKGROUND

In August 2015, pursuant to 10 CFR 50.90, PSEG requested an amendment to the Salem facility operating license to remove the Pressurizer Power Operated Relief Valve (PORV) position indication from the Unit 1 Technical Specifications 3/4.3.3. 7, Accident Monitoring Instrumentation. The change was requested and approved on an emergency basis as permitted by 10 CFR 50.91(a)(5), and was the result of a 10 CFR Part 21 evaluation associated with the Namco limit switches. Per LAR S15-04, 1 due to the emergency nature of the request, the LAR was only for the specific line item for Salem Unit 1.

1 PSEG Emergency License Amendment Request to Remove Pressurizer Power Operated Relief Valve {PORV)

Position Indication Instrumentation from the Accident Monitoring Instrumentation Technical Specifications, dated August 31, 2015 (accession number ML15243A491).

2

LR-N16-0003 LAR 816-01 PSEG is submitting this LAR to revise Unit 1 and Unit 2 TS Tables 3.3-11 and 4.3-1 1, Accident Monitoring Instrumentation. The purpose is to align the remainder of the TS tables with the requirements of Regulatory Guide (RG) 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident",

Revision 22 dated December 1980 and the NUREG 1431, Revision 4, Standard Technical Specifications (STS).

NUREG-0737, "Clarification of TMI Action Plan Requirements" (Reference 1 3), included requirements for accident monitoring instrumentation to be listed in plant Technical Specifications. Subsequently, the NRC Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, which was later codified by changes to 10 CFR 50.36, provided a specific set of objective criteria as guidance for determining which regulatory requirements and operating restrictions should be included In Technical Specifications:

(A) Criterion 1. Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

(B) Criterion 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(C) Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(D) Criterion 4. A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

RG 1.97 describes a method acceptable to the NRC for complying with the regulations to provide instrumentation to monitor plant variables and systems during and following an accident in a light-water-cooled nuclear power plant. Variables and Categories are defined as follows:

Type A Variables are those variables to be monitored that provide the primary information required to permit the control room operator to take specific manually controlled actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for design basis accident events.

Type B Variables are those variables that provide information to indicate whether plant safety functions are being accomplished. Plant safety functions are (1) reactivity control, (2) core cooling, (3) maintaining reactor coolant system integrity, and (4) maintaining containment integrity (including radioactive effluent control).

2 Salem was designed prior to the issuance of RG 1.97. The UFSAR was subsequently updated to demonstrate compliance with the intent of RG 1.97, Revision 2. In various site-specific evaluations of PAM instruments, Revision 3 was used by Salem as more clearly presenting guidance for PWR variables while at the same time being essentially equivalent to Revision 2 guidance. The current RG 1.97 (Revision 4) states that it is primarily intended for new reactors and that previous versions of the RG remain in effect for licensees of current operating reactors.

Revision 2 remains the Salem licensing basis for RG 1.97 compliance.

3

LR-N16-0003 LAR 516-01 Type C Variables are those variables that provide information to indicate the potential for being breached or the actual breach of the barriers to fission product releases. The barriers are (1) fuel cladding, (2) primary coolant pressure boundary, and (3) containment.

Type D Variables are those variables that provide information to indicate the operation of individual safety systems and other systems important to safety. These variables are to help the operator make appropriate decisions in using the individual systems important to safety in mitigating the consequences of an accident.

Type E Variables are those variables to be monitored as required for use in determining the magnitude of the release of radioactive materials and continually assessing such releases.

Category 1, in general, provides for full qualification, redundancy, and continuous real time display and requires onsite (standby) power.

Category 2, in general, provides for qualification but is less stringent in that it does not (of itself) include seismic qualification, redundancy, or continuous display and requires only a high-reliability power source (not necessarily standby power).

Category 3 is the least stringent. It provides for high-quality commercial-grade equipment that requires only offsite power.

The current TS contains instrumentation that does not meet any of the 10 CFR 50.36(c)(2)(ii) screening criteria for inclusion in TS. It also contains TS allowed outage times (7 days for less than required number of OPERABLE channels, 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> for less than minimum number of OPERABLE channels) that are overly restrictive and are not commensurate with the safety significance of the instruments.

4.0 TECHNICAL ANALYSIS

4.1 Methodology The accident monitoring instrumentation for the Salem Unit 1 and Unit 2 TS Tables 3.3-11 and 4.3-11 was assessed as follows:

The following documents were reviewed and compared with each other, specifically to identify post-accident monitoring (PAM) instruments as Type A or Category 1:

Document Reference Regulatory Guide 1.97 1

Salem UFSAR 2

Salem TS Tables 3.3-11/4.3-11 3

Standard Westinghouse TS 4

Engineering Evaluation 5

S-C-GSC-CE E-04 70 4

LR-N16-0003 LAR 516-01 Based on the comparisons, instruments that should be removed because they are not Type A variable or not Category 1 were identified.

Based on the comparisons, instruments that should be added to the TS because they are Type A variables, or non-Type A, but Category 1 were identified.

Type A Variables are defined as those variables to be monitored that provide the primary information required to permit the control room operator to take specific manually controlled actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for design basis accident events.

Category 1, in general, provides for full q ualification, redundancy, and continuous real-time display and requires onsite (standby) power.

The intent is to align the instruments in the Salem Unit 1 and U nit 2 TS Tables 3.3-1 1 and 4.3-1 1 to be consistent with the scope of the N UREG-1 431 reviewer's note which states that the TS should include all RG 1.97 Type A instruments and all RG 1.97, Category 1, non-Type A instruments in accordance with the Unit's RG 1.97 Safety Evaluation Report. The instruments proposed for removal from the TS tables are not being removed from the plant and will continue to satisfy their RG 1.97 requirements.

4.2 Technical Evaluation Each of the instruments listed in Table 1 below, were evaluated and determined to be removed or added to the Salem Unit 1 and U nit 2 TS Tables 3.3-1 1 and 4.3-1 1.

Table 1 Instrument Reactor Coolant System Sub-Cooling Margin Monitor PORV Position Indicator PORV Block Valve Position Indicator Pressurizer Safety Valve Position I ndicator Containment Pressure (Narrow Range)

Wide Range Neutron Flux Monitors AFW (Condensate) Storage Tank Water Level 4.3 Individual I nstrument Evaluations Reactor Coolant System Sub-Cooling Margin Monitor Salem Unit Action U 1 and U2 Remove Unit 2 only:;

Remove U 1 and U2 Remove U 1 and U2 Remove U 1 and U2 Remove U 1 and U2 Add U1 and U2 Add The reactor coolant system (RCS) sub-cooling margin monitor (SMM ) is not listed in the STS. It should be noted, that this instrument is categorized in Table 3 of RG 1.97 as a Category 2 variable.

For the Salem Generating Station, Eng ineering Evaluation S-C-GCS-CEE-0470 states that the sub-cooling margi n monitor is a Type B Category 2 variable in accordance with the requirements of RG 1.97. It is currently listed in U FSAR Table 7.5-3, as a Type A variable, Category 1. As discussed below, degrees of subcooling should be removed from U FSAR Table 7.5-3 because it does not meet the Type A or Category 1 criteria and should be classified as Type B, Category 2.

3 The PORV position indicator was removed from Salem Unit 1 by TS Amendment 310 5

LR-N16-0003 LAR 516-01 Type B Variables are defined as those variables that provide information to indicate whether plant safety functions are being accomplished. Plant safety functions are (1) reactivity control, (2) core cooling, (3) maintaining reactor coolant system integrity, and

( 4) maintaining containment integrity (including radioactive effluent control).

Category 2, in general, provides for qualification but is less stringent in that it does not (of itself) include seismic qualification, redundancy, or continuous display and requires only a high-reliability power source (not necessarily standby power).

The RCS sub-cooling margin indication provides information to the operators related to satisfying one of the Safety Injection (SI) termination criteria following a steam line break or Steam Generator Tube Rupture (SGTR) accident. The inputs to the RCS sub-cooling margin monitor are the core exit thermocouples for RCS temperature and the wide range RCS pressure indication for RCS pressure.

The RCS SMM does not detect or indicate a significant abnormal degradation of the reactor coolant pressure boundary, as required by Criterion 1. This is consistent with the NRC Final Policy Statement, which provided that Criterion 1 is intended to ensure that those instruments specifically installed to detect excessive reactor coolant system leakage be included in the TS.

The instrumentation that satisfies Criterion 1 is contained in Salem Unit 1 TS 3.4.6.1 and Unit 2 TS 3.4. 7.1 "Reactor Coolant System Leakage, Leakage Detection Systems."

RCS sub-cooling instrumentation is not a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis considered in Criterion 2.

The RCS SMM is not part of a primary success path as indicated in Criterion 3.

The loss of the RCS SMM instrumentation has no effect on the probabilistic safety assessment, and has not been shown to be significant to health and safety as considered in Criterion 4.

The RCS sub-cooling margin indication provides information to indicate whether the core cooling safety function is being accomplished, and is therefore considered a Type B variable.

The RCS sub-cooling indication is a backup to the core exit thermocouples and RCS pressure, and is therefore considered Category 2. In support of this LAR, the UFSAR (Tables 7.5-1, 7.5-3, and 7.5-4) and associated design documentation, as appropriate, are being revised to reflect that the subcooling margin monitor is not a Type A variable. PSEG has determined these documentation changes can be implemented in accordance with 1 0 CFR 50.59. The Salem UFSAR will be revised prior to implementation of this proposed TS change.

The RCS sub-cooling margin indication does not meet any of the four screening criteria of the NRC Final Policy Statement in 58 Federal Register 391 32 (58FR391 32), which provides criteria to determine which items are required to be included in the TS as LCOs. This conclusion is supported by the absence of operability and surveillance requirements for the sub-cooling margin indication in the improved standard TS presented in NUREG-1 431. Accordingly, removal of the RCS sub-cooling margin indication from the TS conforms to the STS.

Therefore, the reactor coolant system sub-cooling margin monitor is removed from the Salem Unit 1 and Unit 2 TS Tables 3.3-1 1 and 4.3-11.

6

LR-N16-0003 PORV Position Indication LAR 516-01 The reactor coolant system is protected against over-pressurization by control and protective circuits such as the pressurizer pressure high reactor trip and by the PORVs connected to the top of the pressurizer. The PORVs provide a means for pressure relief. Each PORV is equipped with two limit switches to provide Open and Closed indication (i.e. lights) in the control room.

The PORV limit switch position indicators provide information to the control room operators related to the position of the pressurizer PORV's. The Design Basis Accident (DBA) analysis of an inadvertent opening of the PORV does not rely on operator diagnosis and closure of the PORV or block valve, the DBA analysis assumes that automatic safety injection actuation will provide adequate protection. At Salem, this event is bounded by the more limiting inadvertent opening of a code safety valve as described in Salem UFSAR Section 15.2.12.1, Accidental Depressurization of the Reactor Coolant System. At Salem, no credit for diagnosis and re closure of a PORV, block valve, or code valve is assumed in the DBA analysis.

The Salem UFSAR identifies "Primary System Safety Relief Valve Positions (including PORV and code valves)" as Type D, Category 2 Variables, consistent with RG 1.97 (UFSAR Section 7.5). Type D variables are those variables that provide information to indicate the operation of individual safety systems and other systems important to safety. Category 2 instruments are designed to less stringent qualifications that do not require seismic qualification, redundancy, or continuous display, and require only a high reliability power source, not necessarily standby power. Removing the PORV position indication from the TS conforms with the NRC position on application of the screening criteria to post-accident monitoring instrumentation.

The PORVs themselves are part of the primary success path in the UFSAR accident analysis because they are assumed to actuate to mitigate a DBA and therefore meet Criterion 3 of the NRC Final Policy Statement. For example, they are credited in Salem UFSAR Section 15.2. 1 4, Spurious Operation of the Safety Injection System at Power. The operability of the PORVs is therefore required by TS 3.4.3, "Relief Valves." However, PORV position indication does not detect or indicate a significant abnormal degradation of the reactor coolant pressure boundary, as required by Criterion 1. PORV position indication is not a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis considered in Criterion 2.

While the function of the PORVs themselves is part of the primary success path in the UFSAR, PORV position indication is not part of the primary success path. UFSAR accident analysis assumes that the PORVs open as designed to reduce reactor pressure and no operator action based on PORV position indication is required. Therefore, PORV position indication is not part of the primary success path as indicated in Criterion 3.

The loss of PORV position indication instrumentation has no effect on the probabilistic safety assessment, and has not been shown to be significant to health and safety as considered in Criterion 4.

PORV position indication instrumentation does not meet any of the four screening criteria of the NRC Final Policy Statement in 58 Federal Register 391 32 (58FR39132), which provides criteria to determine which items are required to be included in the TS as LCOs. This conclusion is supported by the absence of operability and surveillance requirements for the PORV position indication instrumentation in the improved standard TS presented in NUREG-1 431.

7

LR-N16-0003 LAR 516-01 Accordingly, removal of PORVs from TS conforms to the STS. The PORV position indicator was previously removed from Salem Unit 1 through TS Amendment 31 0.

Therefore, the PORV position indication is removed from the Salem Unit 2 TS Tables 3.3-1 1 and 4.3-11.

PORV Block Valve Position Indication The PORVs can be isolated by the PORV block valves which are connected in line with the PORVs. The PORV block valves provide closure redundancy to the PORVs, to isolate PORVs with excessive seat leakage or that stick open. Use of the PORV block valves in the event of a PORV malfunction can prevent a severe depressurization of the RCS with potential for uncovering of the reactor core.

The PORV block valve limit switch position indication provides information to the control room operators on the position of the pressurizer PORV block valves. It could be used to diagnose the availability of the pressurizer PORV's for use in depressurizing the RCS or to indicate the isolation of a stuck open PORV (LOCA) at lower RCS temperatures. The PORV block valve limit switch position indication does not provide an indication for operator actions for which no automatic control is provided or impact the response of the PORVs to a design basis accident.

Furthermore, this instrumentation is not needed for manual operator action necessary for safety systems to accomplish their safety function for the design-basis events.

The PORVs themselves are part of the primary success path in the U FSAR accident analysis because they are assumed to actuate to mitigate a Design Basis Accident (DBA) and therefore meet Criterion 3 of the NRC Final Policy Statement. In order to provide this function, the PORV block valves m ust be open or capable of being manually opened. The operability of the PORV block valves is therefore required by Unit 1 TS 3.4.3 and Unit 2 TS 3.4.5, "Relief Valves."

However, PORV block valve position indication does not detect or indicate a significant abnormal degradation of the reactor coolant pressure boundary, as required by Criterion 1. This is consistent with the NRC Final Policy Statement, which provided that Criterion 1 is intended to ensure that those instruments specifically installed to detect excessive reactor coolant system leakage be included in the TS. Criterion 1 is not to be interpreted to include instrumentation installed to identify the source of actual leakage, for example valve position indicators. The instrumentation that satisfies Criterion 1 is contained in Salem Unit 1 TS 3.4.6.1 and Unit 2 TS 3.4. 7.1 "Reactor Coolant System Leakage, Leakage Detection Systems."

PORV block valve position indication is not a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis considered in Criterion 2.

While the function of the PORVs themselves is part of the primary success path in the UFSAR, the PORV block valve position indication is not part of the primary success path. U FSAR accident analysis assumes that the PORVs open as designed to reduce reactor pressure. In the event a PORV block valve is closed to isolate PORV valve seat leakage as allowed by the technical specifications, the emergency operating procedures direct the operators to manually open the PORV and associated block valve. This action does not rely on the PORV block valve position indication. Therefore, PORV block valve position indication is not part of the primary success path as indicated in Criterion 3.

8

LR-N16-0003 LAR 516-0 1 The loss of PORV block valve position indication instrumentation has no effect on the probabilistic safety assessment, and has not been shown to be significant to health and safety as considered in Criterion 4.

PORV block valve position indication does not meet any of the four screening criteria of the NRC Final Policy Statement in 58 Federal Register 391 32 (58FR391 32), which provides criteria to determine which items are required to be included in the TS as LCOs. This conclusion is supported by the absence of operability and surveillance requirements for the PORV block valve position indication in the improved standard TS presented in NU REG-1 431. Accordingly, removal of the PORV block valve position indication from the TS conforms to the STS.

Thus, since the PORV block valve limit switch positions provide information to indicate the status of the pressurizer PORV block valves which are used to isolate the PORVs in the event of excessive PORV leakage, they are a Type D4 variable. For the same reasons as the PORV position indications, the block valves position indicators do not meet Criterion 1, 2, 3 or 4 of the NRC Final Policy Statement and an LCO is not required concerning the PORV block valve position indicators. Because the PORV block valve position indicator is not classified as a Type A variable, or a non-Type A but Category 1 variable, the PORV block valve position indication is not required to be in the Salem Unit 1 and Unit 2 TS Tables 3.3-1 1 and 4.3-1 1.

Therefore, the PORV block valve position indication is removed from the Salem Unit 1 and Unit 2 TS Tables 3.3-1 1 and 4.3-1 1.

Pressurizer Safety Valve Position Indication The Pressurizer Safety Valve Position Indication provides information to the control room operators on the position of the pressurizer safety valves. It could be used to diagnose high RCS pressures or a stuck open safety valve (i.e. LOCA) at lower RCS pressures. They are both considered Type D variable, Category 2.

The Pressurizer safety relief valves themselves are part of the primary success path in the UFSAR accident analysis because they are assumed to actuate to mitigate a Design Basis Accident (DBA) and therefore meet Criterion 3 of the NRC Final Policy Statement. The operability of the Pressurizer safety valves is therefore required by Unit 1 TS 3.4.2 and Unit 2 TS 3.4.2 & 3.4.3, "Safety Valves." However, the Pressurizer safety valve position indication does not detect or indicate a significant abnormal degradation of the reactor coolant pressure boundary, as required by Criterion 1. This is consistent with the NRC Final Policy Statement, which provided that Criterion 1 is intended to ensure that those instruments specifically installed to detect excessive reactor coolant system leakage be included in the TS. Criterion 1 is not to be interpreted to include instrumentation installed to identify the source of actual leakage, for example valve position indicators. The instrumentation that satisfies Criterion 1 is contained in Salem Unit 1 TS 3.4.6.1 and Unit 2 TS 3.4. 7.1 "Reactor Coolant System Leakage, Leakage Detection Systems."

Pressurizer safety relief valve position indication is not a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis considered in Criterion 2.

4 UFSAR Section 7.5 describes implementation of RG 1.97 at Salem. Type D variables are defined in RG 1.97, as variables that provide information to indicate the operation of individual safety systems and other systems important to safety.

9

LR-N16-0003 LAR 516-01 While the function of the Pressurizer safety relief valves themselves is part of the primary success path in the UFSAR, the valve position indication is not part of the primary success path.

UFSAR accident analysis assumes that the Pressurizer safety relief valves open as designed to reduce reactor pressure and no operator action based on valve position indication is required.

Therefore, the Pressurizer safety valve position indication is not part of the primary success path as indicated in Criterion 3.

The loss of Pressurizer safety valve position indication instrumentation has no effect on the probabilistic safety assessment, and has not been shown to be significant to health and safety as considered in Criterion 4.

Pressurizer Safety Valve position indication does not meet any of the four screening criteria of the NRC Final Policy Statement in 58 Federal Register 391 32 (58FR391 32), which provides criteria to determine which items are required to be included in the TS as LCOs. This conclusion is supported by the absence of operability and surveillance requirements for the Pressurizer Safety Valve position indication in the improved standard TS presented in NUREG-1 431. Accordingly, removal of the Pressurizer Safety Valve position indication from the TS conforms to the STS.

Since the position indication for these valves does not provide an indication for operator actions for which no automatic control is provided, it does not satisfy Criterion 1, 2, 3 or 4 of the NRC Final Policy Statement for the same reasons as discussed for the PORVs above.

The Pressurizer Safety Valve indication is not included in the STS. They are not Category 1 in RG 1.97 nor are they Type A or Category 1 within the Salem UFSAR or Engineering Evaluation S-C-GCS-CEE-0470, thus, the Pressurizer Safety Valve Position Indication is not required to be in the Salem Unit 1 and Unit 2 TS Tables 3.3-1 1 and 4.3-1 1.

Therefore, the Pressurizer Safety Valve Position Indication is proposed to be removed from the Salem Unit 1 and Unit 2 TS Tables 3.3-1 1 and 4.3-1 1.

Containment Pressure (Narrow Range)

The containment pressure indication provides information for assessing an inadequate containment cooling condition and for determining the potential challenge to the containment pressure retaining integrity. The wide range containment pressure instrumentation, which is also listed in Salem Unit 1 and Unit 2 TS Table 3.3-1 1 (Line Item #16), provides an adequate range and sensitivity for this purpose. Only the wide range instrumentation is used in the emergency operating procedures (EOPs) to define the potential for a challenge to containment integrity due to over-pressurization.

If containment heat removal systems are functioning properly, no challenge to containment integrity should occur due to containment pressure. Containment pressure instrumentation is used in the severe accident management guidelines (SAMG) to indicate a possible containment integrity challenge and to initiate the assessment of containment venting strategies. It is also used as an indicator of the potential loss of a fission product barrier in the emergency plan.

Containment pressure is a key indicator in the declaration of a General Emergency level and the potential need for offsite radiological protection actions. Therefore, the containment wide range instrumentation (by itself) satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).

1 0

LR-N16-0003 LAR 516-01 The STS requires only the Containment Pressure (Wide Range) instrumentation be included in the PAM TS. Per Salem UFSAR 7.5-3, the measurement of Containment Pressure is a Type A, Category 1 variable; however, no distinction is made between the use of wide range or narrow range instruments.

Since the redundancy requirements and other requirements of RG 1.97 are met with the Containment Pressure (Wide Range) channels, the Containment Pressure (Narrow Range) channels are not required to be in the Salem Unit 1 and Unit 2 TS Tables 3.3-11 and 4.3-11.

In support of this LAR, the UFSAR (Tables 7.5-1, 7.5-3, and 7.5-4, as appropriate) and Item 11 of S-C-GCS-CEE-0470 will be revised to reflect that the narrow range containment pressure instrumentation is not a Type A variable. PSEG has determined these documentation changes can be implemented in accordance with 10 CFR 50.59. The Salem UFSAR will be revised prior to implementation of this proposed TS change.

The containment pressure narrow instrument channels themselves are part of the primary success path in the UFSAR accident analysis because they are assumed to actuate to mitigate a Design Basis Accident (DBA) and therefore meet Criterion 3 of the NRC Final Policy Statement. The operability of the containment pressure narrow range channels is therefore required by Unit 1 and 2 TS 3.3.2.1, "Engineered Safety Feature Actuation System Instrumentation." However, containment pressure narrow range indication does not detect or indicate a significant abnormal degradation of the reactor coolant pressure boundary, as required by Criterion 1. This is consistent with the NRC Final Policy Statement, which provided that Criterion 1 is intended to ensure that those instruments specifically installed to detect excessive reactor coolant system leakage be included in the TS. Criterion 1 is not to be interpreted to include instrumentation installed to identify the source of actual leakage, for example valve position indicators.

Containment pressure narrow range indication is not a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis considered in Criterion 2.

While the function of the containment pressure narrow range channels themselves is part of the primary success path in the UFSAR, the pressure indication is not part of the primary success path. UFSAR accident analyses assume automatic actuation of the safety injection and containment spray system based on containment narrow range pressures and no operator actions based on containment pressure narrow range indication are required. Therefore, the containment pressure narrow range indication is not part of the primary success path as indicated in Criterion 3.

The Containment Pressure narrow range indication has no effect on the probabilistic safety assessment, and has not been shown to be significant to health and safety as considered in Criterion 4.

Containment Pressure (Narrow Range) indication does not meet any of the four screening criteria of the NRC Final Policy Statement in 58 Federal Register 39132 (58FR39132), which provides criteria to determine which items are required to be included in the TS as LCOs. This conclusion is supported by the absence of operability and surveillance requirements for the Containment Pressure (Narrow Range) indication in the improved standard TS presented in 11

LR-N16-0003 LAR 516-01 NUREG-1431. Accordingly, removal of Containment Pressure (Narrow Range) indication from the TS conforms to the STS.

Therefore, the Containment Pressure (Narrow Range) is removed from the Salem Unit 1 and Unit 2 TS Tables 3.3-1 1 and 4.3-11.

Wide Range Neutron Flux Monitors (Gamma-Metrics Post-Accident Neutron Monitoring)

The Neutron Flux is listed as an instrument within the STS, and is listed as Category 1 in RG 1.97, and is Category 1 in Engineering Evaluation S-C-GSC-CEE-0470.

The bases in Westinghouse STS (Reference 4) state that power range and source range neutron flux indication is provided to verify reactor shutdown. The two ranges are necessary to cover the full range of flux that may occur post-accident. It also states that neutron flux is used for accident diagnosis, verification of sub-criticality, and diagnosis of positive reactivity insertion.

Gamma-Metrics Post-Accident Neutron Monitoring (PANM) system was installed at Salem to provide reliable neutron flux monitoring in a harsh envimnment from plant shutdown to full power, and thereby specifically satisfies RG 1.97 requirements. In the wide-range mode, the PANM output range meets the requirement of RG 1.97 (Reference 1 ).

Therefore, the Wide Range Neutron Flux Monitors are added to the Salem Unit 1 and Unit 2 TS Tables 3.3-11 and 4.3-11.

Auxiliary Feed Water (AFW) Storage Tank (Condensate Storage Tank5) Water Level The AFW storage tank water level is listed as a Type A variable per the Salem UFSAR Table 7.5-3 and Category 1 by Table 3 in RG 1.97. It is also listed as Type A and Category 1 in Engineering Evaluation S-C-GCS-CEE-0470. Because of this site-specific determination the instrumentation should be added to TS Tables 3.3-11 and 4.3-1 1 for Unit 1 and Unit 2 in accordance with the guidance of the reviewer's notes in NUREG-1431.

Therefore, the Auxiliary Feed Water (AFW) Storage Tank (Condensate Storage Tank) Water Level is added to the Salem Unit 1 and Unit 2 TS Tables 3.3-11 and 4.3-1 1.

Exceptions:

All instruments that are Type A variables, or non-Type A, but Category 1 were evaluated and are proposed to be included in the Salem Unit 1 and Unit 2 TS Tables 3.3-1 1 and 4.3-1 1 with the following exceptions:

Containment Isolation Valve (CIV) Position Indicators PSEG is not proposing to add CIV position indication to TS Tables 3.3-11 and 4.3-11.

For valves which receive an automatic isolation signal to close in order to accomplish containment isolation, valve position is a Type B, Category 1 variable. Type B variables are those which provide information to indicate whether plant safety functions are being accomplished. In the Reference 8 letter to vendor owners 5 In various PAM instrument documents, Condensate Storage Tank water-level is listed because the CST is frequently the source of auxiliary feedwater. At Salem, the AFW storage tank is the source of auxiliary feedwater.

Hence AFW storage level and CST water level are used interchangeably in this document.

12

LR-N16-0003 LAR 516-01 groups on application of the policy statement criteria to standard technical specifications, the NRC staff concurred that that only Type A variables meet Criterion 3 as a structure. system or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the Integrity of a fission product barrier. However, the staff was unable to confirm the owners groups' conclusion that non-Type A Category 1 instrumentation can be relocated from Technical Specifications Inclusion of CIV position indication in Salem TS Tables 3.3-1 0 and 4.3-1 0 is not required to ensure adequate information is available to operators to verify containment operability and the accomplishment of containment isolation when required. The operability requirements for the containment are controlled by TS 3.6.1. Limiting Condition for Operation (LCO) 3.6.1.1 requires primary CONTAINMENT INTEGRITY to be maintained in Modes 1, 2, 3 and 4.

CONTAINMENT INTEGRITY requires that all penetrations required to be closed during accident conditions are either capable of being closed by an OPERABLE containment automatic isolation valve system, or closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except for valves that are open under administrative control as permitted by TS 3.6.3.1.

The operability requirements for automatic containment isolation valves are controlled in TS 3.6.3.1 (unit 1 ) and TS 3.6.3 (Unit 2). LCO 3.6.3.1 (3.6.3 Unit 2) requires each containment isolation valve to be OPERABLE in Modes 1, 2, 3 and 4.

Surveillance Requirement (SR) 4.6.3. 1.2 (SR 4.6.3.2 Unit 2) requires each containment isolation valve to be demonstrated OPERABLE by periodically verifying the valves actuate to their isolation positions in response to the required containment isolation signals. SR 4.6.3.1.3 (SR 4.6.3.3 Unit 2) requires periodic verification that on a main steam Isolation test signal, each main steam isolation valve actuates to its isolation position. SR 4.6.3.1.4 (SR 4.6.3.4 Unit 2) requires the isolation time of each automatic containment isolation valve to be determined to be within its limit when tested pursuant to the lnservice Testing Program. Proper position indication is verified during conduct of surveillance testing of the associated valves.

CIV position indication was not included in the Amendment which added accident monitoring instrumentation to the Unit 1 Technical Specifications (Reference 9) or in the original Unit 2 Technical Specifications (Reference 1 0). The CIV position indication meets the associated Regulatory Guide 1.97 requirements and the current TS requirements are sufficient to ensure adequate information is available to operators to verify containment operability and the accomplishment of containment isolation when required.

Therefore, the Containment Isolation Valve Position Indicators is not added to the Salem Unit 1 and Unit 2 TS Tables 3.3-1 1 and 4.3-1 1.

Containment Hydrogen Concentration Containment Hydrogen Concentration is listed as a Type A, Category I variable in Engineering Evaluation S-C-GSC-CEE-0470, not listed as Type A in UFSAR Table 7.5-3, but also listed as Category I in RG 1.97. However, it is not listed in the STS.

1 3

LR-N16-0003 LAR 516-01 This instrument was included in the rulemaking for 1 0 CFR 50.44. The statement of considerations for the 50.44 rulemaking states that this instrument can be relocated from the Technical Specifications and can be re-classified as Type C, Category 3 per the RG 1.97 definitions. Not including these instruments is consistent with Salem LCR S06-05 and Salem TS Amendments 281 and 264 (Reference 7).

Therefore, the Containment Hydrogen Monitors are not added to the Salem Unit 1 and Unit 2 TS Tables 3.3-1 1 and 4.3-1 1.

4.4 LCO Times and Separate Condition Entry The allowed outage times (AOT) in current Technical Specification 3.3.3.7 have their origin in NUREG-0452, "Standard Technical Specifications for Westinghouse Pressurized Water Reactors". The STS (NUREG-1431) extended the 7-day completion time for one inoperable instrument channel to 30 days and the 48-hour completion time for two inoperable channels to 7 days. Additionally, the STS removed the shutdown requirement for a single inoperable instrument channel.

With one channel inoperable beyond 30 days, a special report outlining the preplanned method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels must be submitted to the NRC within the next 14 days. With two channels inoperable for more than 7 days, the STS requires either a plant shutdown or submittal of a special report, as discussed above, depending on the particular channel that is out of service. The STS also contain provisions that permit a separate condition entry for each inoperable instrument function.

The STS 30 day completion time is based on operating experience and takes into account the remaining OPERABLE channel (or in the case of a function that has only one required channel, other non-Regulatory Guide 1.97 instrument channels to monitor the Function), the passive nature of the instrument (no critical automatic action is assumed to occur from these instruments), and the low probability of an event requiring PAM instrumentation during this interval.

When the required actions and 30 day completion time is not met, a written report is required to be submitted. This report discusses the results of the root cause evaluation of the inoperability and identifies proposed restorative actions. This action is appropriate in lieu of a shutdown requirement since alternative actions are identified before loss of functional capability, and given the likelihood of unit conditions that would require information provided by this instrumentation.

The STS completion time of 7 days is based on the relatively low probability of an event requiring PAM instrument operation and the availability of alternate means to obtain the required information.

Since the TS is being revised to better align to the STS, this amendment revises the following instruments within the Salem Unit 1 and Unit 2 TS Table 3.3-11 to adopt the allowable outage times in STS of 30 days for one inoperable channel less than the required number of channels and 7 days for two or more channels less than the required number of channels (versus the current 7 days and 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, respectively). Other minor changes to the listed actions are discussed below.

1 4

LR-N16-0003

1.
2.
3.
4.
5.
6.
7.
8.
10.
16.
17.
18.
22.
23.

Instrument Reactor Coolant Outlet Temperature-T Hot (Wide Range)

Reactor Coolant Outlet Temperature-T cold (Wide Range)

Reactor Coolant Pressure (Wide Range)

Pressurizer Water Level Steam Line Pressure Steam Generator Water Level (Narrow Range)

Steam Generator Water Level (Wide Range)

Refueling Water Storage Tank Water Level Auxiliary Feedwater Flow Rate Containment Pressure -Wide Range Containment Water Level-Wide Range Core Exit Thermocouples Wide Range Neutron Flux Monitors AFW (Condensate) Storage Tank Water Level The following TS Table 3.3-1 1 Actions are revised:

Action 1 LAR 516-01 Action 1' 2 1' 2 1' 2 1' 2 1' 2 1' 2 1' 2 1' 2 4,6 7,2 7,2 1' 2 1' 2 1' 2 With the number of OPERABLE accident monitoring channels less than the Required Number of Channels shown in Table 3.3-1 1, restore the inoperable channel(s) to OPERABLE status within 30 days, or submit a special report in accordance with Specification 6.9.4.

Action 2 With the number of OPERABLE accident monitoring channels less than the Minimum Number of Channels shown in Table 3.3-1 1, restore the inoperable channel(s) to OPERABLE status within 7 days 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 />.

Action 4 With the number of OPERABLE channels one less than the Required Number of channels shown in Table 3.3-1 1, operation may proceed provided that an OPERABLE Steam Generator Wide Range level channel is available as an alternate means of indication for the Steam Generator with no OPERABLE Auxiliary Feedwater Flow Rate channel; OTHERWISE, restore the inoperable channel to OPERABLE status within 30 days, or submit a special report in accordance with Specification 6.9.4.

Action 6 With the number of OPERABLE channels less than the Minimum Number of channels shown in Table 3.3-1 1, restore the inoperable channel(s) to OPERABLE status within 7 days, 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 />.

This amendment also revises the Salem TS 3.3.3. 7 to include the STS note which allows for a separate condition entry for each instrument. This will allow each instrument the full associated LCO time, if multiple PAM instruments become inoperable at different times.

The surveillance frequencies for Line Items 22 and 23 that are being added to TS Table 4.3-1 1 will be contained in the Surveillance Frequency Control Program with the exception of the 1 5

LR-N16-0003 LAR S16-01 Channel Functional Test which is not applicable for these instruments. Salem has implemented TSTF Change Traveler TSTF-425, Revision 3 ( Reference 1 1 ) as approved by the NRC with TS Amendments 299 and 282 ( Reference 1 2) for Units 1 and 2 respectively. The initial surveillance frequency will be 31 -days for the CHANNEL CHECK and 1 8-months for the CHANNEL CALIBRATION consistent with NUREG-1 431, Standard Technical Specifications for Westinghouse Plants.

The criteria for relocation of a surveillance frequency to a licensee controlled program in accordance with TSTF-425 were reviewed. These surveillance frequencies are periodic surveillances that: 1 ) do not reference other approved programs for the specified interval, 2) are not event driven, 3) do not have a time component based on event occurrence, and 4) are not related to a specific condition for performance. Therefore the periodic surveillance freq uencies are within the scope of TSTF-425 for location in the licensee controlled Surveillance Frequency Control Program.

Finally, TS 6.9.4, Special Reports, is revised to include Actions 1 and 4 of TS Table 3.3-1 1.

5.0 REGULATORY ANALYSIS

5. 1 No Significant Hazards Consideration PSEG req uests an amend ment to the Salem Unit 1 and U nit 2 Operating Licenses. The proposed changes would modify Accident Monitoring Technical Specifications (TS) Tables 3.3-1 1 and 4.3-1 1, titled "Accident Monitoring Instrumentation" and "Surveillance Requirements for Accident Monitoring Instrumentation" respectively. Specifically, these changes would add Regulatory Guide ( RG ) 1.97 Type A or Category 1 instruments that are currently absent into Tables 3.3-1 1 and 4.3-1 1, and would remove instruments that are not Type A, nor Category 1 from Tables 3.3-1 1 and 4.3-1 1. The specific instruments to be removed or added are summarized below:

Instruments removed from TS Table 3.3-1 1 and TS Table 4.3-1 1 because they are not regulatory guide ( RG ) 1.97 Type A, or Category 1 :

Reactor Coolant System Sub-Cooling Margin Monitor*

PORV Position Indicator (Unit 2 only)*

PORV Block Valve Position Indicator*

Pressurizer Safety Valve Position Indicator*

Containment Pressure (Narrow Range)

  • The associated footnotes are also removed.

I nstruments added to TS Table 3.3-1 1 and TS Table 4.3-1 1 :

Wide Range Neutron Flux Monitors Auxiliary Feed Water (AFW) Storage Tank (Condensate Storage Tank) Water Level Additionally, allowed outage times (AOT) and required actions of TS Table 3.3-1 1 Actions 1, 2, 4 and 6 would be revised to align with the Westinghouse Standard Technical Specifications (STS), NUREG-1 431. TS Section 6.9.4 is revised to reflect these changes to Actions.

1 6

LR-N16-0003 LAR 516-01 PSEG has evaluated the proposed changes to the TS using the criteria in 1 0 CFR 50.92, and determined that the proposed changes do not involve a significant hazards consideration. The following information is provided to support a finding of no significant hazards; 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed changes to the TS modify Accident Monitoring Instrumentation TS Tables 3.3-1 1 and 4.3-1 1 of Salem Units 1 and 2 by removing or adding instruments as listed above, and updating the AOT and required actions to better align with the Westinghouse STS, NUREG-1 431. The instruments listed above are not assumed to be initiators of any analyzed event of Chapter 1 5 in the Updated Final Safety Analysis Report (UFSAR).

Therefore the probability of an accident previously evaluated is not significantly increased.

The proposed changes do not alter the design of any system, structure, or component (SSC). The proposed changes conform to NRC regulatory guidance regarding the content of plant TS, as identified in 1 0 CFR 50.36, NUREG-1 431, and the NRC Final Policy Statement in 58 FR 391 32.

TS Operability requirements are retained for Type A and Category 1 variables. Operability of these instruments ensures sufficient information is available to monitor and assess plant status during and following an accident. Alternate means for diagnosing and responding to instrument malfunctions are unaffected by the proposed change. Therefore, the consequences of an accident previously evaluated are not significantly increased.

Therefore, these proposed changes do not represent a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new o r different kind of accident from any accident previously evaluated?

Response: No The proposed changes to the TS would modify the TS Tables 3.3-1 1 and 4.3-1 1 of Salem Units 1 and 2, by removing or adding instruments as listed above, and updating the AOT and required actions to better align with the Westinghouse STS. The proposed changes do not involve a modification to the physical configuration of the plant or changes in the methods governing normal plant operation. The proposed changes will not impose any new or different requirement or introduce a new accident initiator, accident precursor, or malfunction mechanism.

Additionally, there is no change in the types or increases in the amounts of any effluent that may be released off-site and there is no increase in individual or cumulative occupational exposure.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

1 7

LR*N16*0003

3.

Do the proposed changes involve a significant reduction in a margin of safety?

Response: No LAR 816-01 The proposed changes to the TS would modify the TS Tables 3.3-1 1 and 4.3-1 1 of Salem Units 1 and 2, by removing or adding instruments as listed above, and updating the AOT and required actions to better align with the Westinghouse STS. The instruments removed from Tables 3.3-1 1 and 4.3-1 1 are not needed for manual operator action necessary for safety systems to accomplish their safety function for the design basis events. The instruments listed for removal are indication-only with the exception of containment pressure narrow range instruments; thus, they do not provide an input to any automatic trip functions.

In the case where similar or related instruments (e.g., containment pressure-narrow range) are associated with important trips (i. e., RPS or ESF trips), such instruments are governed by separate existing TS sections which are not altered by this req uest.

Therefore, since the proposed changes do not impact the response of the plant to a design basis accident, the proposed changes do not involve a significant reduction in a margin of safety.

Based upon the above, PSEG concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 1 0 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

In conclusion, based on the considerations discussed above, ( 1 ) there is a reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in com pliance with the NRC's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.2 Applicable Regulatory Requirements/Criteria 1 0 CFR 50.36(c)(2), Limiting conditions for operation, states: (i) Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met.

Regulatory G uide (RG) 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident". The proposed changes to the Salem TS do not alter the classification of instrumentation required by RG 1.97.

The proposed changes only modify the instruments listed in TS 3/4.3.3.7.

Therefore, based on the considerations discussed above, ( 1 ) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

1 8

LR-N16-0003

6.0 ENVIRONMENTAL CONSIDERATION

LAR S16-01 A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 1 0 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cum ulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 1 0 CFR 51.22(c)(9). Therefore, pursuant to 1 0 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 REFERENCES

1. Regulatory Guide 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident", Revision 2 dated December 1 980

2.

Salem Generating Station (SGS) UFSAR, Section 7.5, "Safety-Related Display I nstrumentation"

3.

Salem Unit 1 and Unit 2 Technical Specification Table, 3.3-1 1, Accident Monitoring I nstrumentation, and Table 4.3-1 1, Surveillance Requirements for Accident Monitoring I nstruments

4.

N UREG-1 431, Volume 1, Specifications, Revision 4.0, "Standard Technical Specifications - Westinghouse Plants," dated April 201 2 (ADAMS Accession No. ML12100A222)

5. S-C-GSC-CE E-0470, "Engineering Evaluation of SGS 1 & 2 Regulatory Guide 1.97 I nstrumentation Compliance with Physical Separation and Electrical Isolation Criteria,"

Revision 1, July 1 996

6. PSEG Emergency License Amendment Request to Remove Pressurizer Power Operated Relief Valve (PORV) Position Indication I nstrumentation from the Accident Monitoring I nstrumentation Technical Specifications, dated August 31, 201 5 (ADAMS Accession No. ML15243A491 )
7.

PSEG License Amendment Request regarding Elimination of Req uirements for Hyd rogen Recombiners and Hydrogen Analyzers Technical Specifications, dated June 7, 2006 8. N RC Letter to Owners Group o n the NRC Staff Review of Nuclear Steam Supply System Vendor Owners Groups' Application of the Com mission's I nterim Policy Statement Criteria to Standard Technical Specifications, dated May 9, 1 988 1 9

LR-N16-0003 LAR 516-01

9.

NRC approval letter regarding Amendment No. 39 which incorporates the requirements for implementation of the TMI-2 Lessons Learned Category "A" Items for Salem Nuclear Generating Station, U nit No. 1, dated October 08, 1 981 1 0. NUREG-0546, "Technical Specifications for Salem Nuclear Generating Station Unit No. 2," dated May 1 981 1 1. Technical Specification Task Force (TSTF) Change Traveler TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5b," NRC approved July 6, 2009 1 2. Letter Richard B. Ennis (NRC) to Thomas Joyce (PSEG), "Salem Nuclear Generating Station, Unit Nos. 1 and 2, Issuance of Amendments RE: Relocation of Specific Surveillance Frequencies to a Licensee-Controlled Program Based on Technical Specification Task Force (TSTF) Change TSTF-425 (TAC Nos. M E3574 and M E3575),"

March 21, 201 1 1 3. NUREG-0737, "Clarification of TMI Action Plan Requirements," dated November 1 980 20

LR-N16-0003 LAR S16-01 Mark-up of Proposed Technical Specification Pages The following Technical Specifications pages for Renewed Facility Operating License DPR-70 are affected by this change request:

Technical Specification 3.3.3. 7, Accident Monitoring Instrumentation Table 3.3-1 1, Accident Monitoring Instrumentation Table 3.3-1 1, Accident Monitoring Instrumentation Table 3.3-1 1, Accident Monitoring Instrumentation Table 3.3-1 1, Accident Monitoring Instrumentation Table 4.3-1 1, Surveillance Requirements for Accident Monitoring Instrumentation Table 4.3-1 1, Surveillance Requirements for Accident Monitoring Instrumentation 6.9.4, Special Reports 3/4 3-53 3/4 3-54 3/4 3-55 3/4 3-56 3/4 3-56a 3/4 3-57 3/4 3-57a 6-24b The following Technical Specifications pages for Renewed Facility Operating License DPR-75 are affected by this change request:

Technical Specification 3.3.3.7, Accident Monitoring I nstrumentation Table 3.3-1 1, Accident Monitoring Instrumentation Table 3.3-1 1, Accident Monitoring I nstrumentation Table 3.3-1 1, Accident Monitoring Instrumentation Table 3.3-1 1, Accident Monitoring Instrumentation Table 4.3-1 1, Surveillance Requirements for Accident Monitoring I nstrumentation Table 4.3-1 1, Surveillance Requirements for Accident Monitoring Instrumentation 6.9.4, Special Reports 1

3/4 3-50 3/4 3-51 3/4 3-51 a 3/4 3-51 b 3/4 3-51 c 3/4 3-52 3/4 3-52a 6-24b

I NSTRUMENTATION ACCI DENT MON ITORI NG I N STRUMENTATION LIMITI NG CONDITION FOR OPERATION 3.3.3.7 The accident monitoring instrumentation channels shown in Table 3.3 1 1 shall be operable.

APPLICABI LITY: MODES 1, 2, and 3.

ACTION:

a.

As shown in Table 3.3-1 1.

SURVEILLANCE REQUIREMENTS 4.3.3.7 Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, CHAN N EL CALI BRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies specified in the S u rveillance Frequency Control Program unless otherwise noted in Table 4.3-1 1.

!--------------------------------------------"()liE:-------------------------------#-------$---------$--%%%

..___-!Separate Condition entry is allowed for each Function.

SALEM - U N IT 1 3/4 3-53 Amendment No. 299

TABLE 3.3-1 1 8CCIDENT MONITORING INSTRUMENTATION REQUIRED MINIMUM NO. OF NO. OF INSTRUMENT CHANNELS CHANNELS ACTION 1. Reactor Coolant Outlet Temperature -

2 1, 2 THor (Wide Range)

2. Reactor Coolant Inlet Temperature -

2 1, 2 T coLo (Wide Range)

3. Reactor Coolant Pressure 2

1, 2 (Wide Range)

4. Pressurizer Water Level 2

1 1 ' 2

5. Steam Line Pressure 2/Steam 1/Steam 1, 2 Generator Generator
6.

Steam Generator Water Level 2/Steam 1/Steam 1, 2 (Narrow Range)

Generator Generator

7.

Steam Generator Water Level 4 (1/Steam 3 (1/Steam 1 ' 2 (Wide Range)

Generator)

Generator)

8. Refueling Water Storage Tank 2

1, 2 Water Level

9. deleted 1 0. Auxiliary Feedwater Flow Rate 4 (1/Steam 3 ( 1/Steam 4, 6 Generator)

Generator) 1 1. Reactor Geelant SystemA a

+

B Margin Monlrof Yoeleted]

1 2. Deleted SALEM - UNIT 1 3/4 3-54 Amendment No. 310

TABLE 3.3-1 1 (CONTINUED)

ACCIDENT MONITORING INSTRUMENTATION INSTRUMENT 

1 3. PORV Blook Val'le Position Indicator

14. Pressurizer Safety Valve Position lndioator 7Deleted 1 5. Gontainment Pressure Narro*HRange REQUIRED NO. OF CHANNELS 2/valve**

2/valve**

2 1 6. Containment Pressure - Wide Ran}e--1 Deleted 1 2 1 7. Containment Water Level -

2 Wide Range 1 8. Core Exit Thermocouples 4/core quadrant 1 9. Reactor Vessel Level Instrumentation 2

System (RVUS)

20. Containment High Range Accident 2

Radiation Monitor 2 1. Main Steamline Discharge (Safety 1/MS Line Valves and Atmospheric Steam Dumps) Monitor Wide Range Neutron Flux Monitors 2

23. Auxiliary Feed Water Storage Tank 2

(Condensate Storage Tank) Water Level MINIMUM NO. OF CHANNELS 1 1

2/core quadrant 1

2 1 /MS Line 1

1 ACTION 4-;--2 4-;-4 7, 2 7, 2 1 t 2 8, 9 10 10 1 t 2 1 ' 2 Total number ofls is aonsid to be t\\vo (2) with ono (1 of the ohannol& bemg any one (1) of the follo*Ning alternate means of determining PORV BlocK, or Safety Valve

@ Tailpipe TemperatuFOs for the valves, Pressurizer Relief Tank Temperature Pressurizer Relief Tank Level OPERABb-, 7Delete SALEM - UNIT 1 Amendment No. 310

TABLE 3

. 3 - ll (continued)

HOT STANDBY within the next 6 submit a special report in TABLE NOTATION accordance with ecification 6.9.4 ACTION 3 ACTION 4 Wi th the number of OPERABLE accident moni toring channels than the Required Number of Channels shown in Table 3. 3 - 11, res the inoperable channel ( s ) to OPERABLE status wi thin f days, or HM' BHO'l'DOWN within tihe fti!iXtl 12 hG'U:Il'S.

6 With the number of OPERABLE accident monitoring channels less than the MINIMUM Number of Channels shown in Tabl e 3. 3 - ll, restore the inoperable channel ( s ) to OPBRABLE s tatus wi thin or be in HOT SHUTDOWN within the 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />.

5 deleted 4following 6 I With the number of OPERABLE channels one les s than the Required Number of Channels shown in Table 3. 3 11, operation may proceed provided that an OPERABLE Steam Generator Wide Range Level channel is avai lable as an al ternate means of indication for the Steam Generator with no OPERATABLE Auxiliary reedwater Plow Rate channel.

ACTION 5 deleted II\\

SALSM UNIT l

OTHERWISE, restore the inoperable channel to OPERABLE status within 30 days, or submit a special report in accordance with Specification 6.9.4.

3 /4 3 - 5 6 Amendment No. 225

following 6 TABLE 3. 3 - 11 (continued)

TABLE NOTATION 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 ACTION 6 With the n mber of OPERABLE channels less than the Mi.imum Number of channels shown in Table 3. 3 -11,

restore the inope. le channel ( s )

OPERABLE status within 7 days, or be in HOT SHUTDOWN within the hours.

ACTION 7 With the number of OPERABLE channels one less than the Required Number of Channels shown in Table 3. 3 -11,

operation may proceed until the next CHANNEL CALIBRATION (which shall be performed upon the next entry into MODE 5, COLD SHUTDOWN).

ACTION 8 With one RVLIS channel inoperable, restore the RVLIS channel to OPERABLE status within 3 0 days, or submi t a special report in accordance with Specification 6. 9. 4.

ACTION 9 With both RVLIS channels inoperable, restore one channel to OPERABLE status within 7 days or submit a special report in accordance with Specification 6. 9. 4.

ACTION 1 0 With the number of OPERABLE Channels less than required by the minimum channels OPERABLE requirements,

init iate the preplanned al ternate method of monitoring the appropriate parameter within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />,

and :

1 ) either restore the inoperable Channel ( s ) to OPERABLE status within 7 days of the event, or 2 ) prepare and submit a Special Report to the Commission pursuant to Specification 6. 9. 2 within 14 days following the event outlining the actions taken,

the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE s-tatus.

SALEM -

UNIT 1 3 / 4 3 -5 6a Amendment No. 272

TABLE 4.3-1 1 SURVEILLANCE REQUIREMENTS FOR ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHANNEL CHANNEL FUNCTIONAL I NSTRUMENT CHECK'1l CALIBRATION'1l TEST'1l 1. Reactor Coolant Outlet Temperature -

N.A.

T HOT (Wide Range)

2. Reactor Coolant Inlet Temperature -

N.A.

T coLo (Wide Range)

3. Reactor Coolant Pressure (Wide Range)

N.A.

4. Pressurizer Water Level N.A.
5.

Steam Line Pressure N.A.

6. Steam Generator Water Level N.A.

(Narrow Range)

7. Steam Generator Water Level N.A.

(Wide Range)

8. Refueling Water Storage Tank Water Level N.A.
9. deleted 1 0. Auxiliary Feedwater Flow Rate S/U#

N.A.

1 1. Reaeter GeelaAt SysteFA Sl:leeeeliA§ N:M N-A MargiA MeAiter y Deleted I Auxiliary Feedwater System is used on each startup and flow rate indication is verified at that time.

t.

The iAstYevelep RGS suboooling FAar§tn are calf9rated in aeeordance wflh4he Surveillance Frequency Gentrol PrograFA; the FAeniter 'Nill be eeFApared with calculated sl.tBZn for knovm input values in accordance with the S[:rveilfa.Are Frequency Gentrel Progr-am-:





SALEM - UNIT 1 3/4 3-57 Amendment No. 299

TABLE 4.3-1 1 (Continued)

SURVEILlANCE REQUIREMENTS FOR ACCIDENT MONITORING lf::ISTRUMENTATION INSTRUMENT

12. Deleted r-lDeleted I 1 3. PORV Bloo7Position ln4teatef
14. Pressurizer Safety Valvo PosltieA Indicator

=0-e...,..le-t-ed....,...,

1 5. GGmaemHfiGe

16. Containment Pressure -

Wide 1 7. Containment Water Level - Wide Range 1 8. Core Exit Thermocouples 1 9. Reactor Vessel Level Instrumentation System (RVLIS)

20. Containment High Range Accident Radiation Monitor 21. Main Steamline Discharge (Safety Valves and Atmospheric Steam Dumps)

Monitor CHANNEL CHANNEL CHECK<1l CALIBRATIQN<1l Table Notation CHANNEL FUNCTIONAL TEST<1l N.A.

N.A.

N.A.

N.A.

(1 )

Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.

t,2-2-. WJ.deRan-g-eN..,...eut-ro-nFI-uxMo-nit-or-s------------------------N.A-.

23. Auxiliary Feed Water Storage Tank (Condensate Storage Tank) Water Level N.A.

Unless the blosk-valve is closed in order to meet the ra8tion b, or o In speGifieation a A

'3' Delete SALEM - UNIT 1 3/4 3-57a Amendment No. 310

AQMINISTRATIVE CONTROLS h.

The primary to secondary leakage rate observed i n each SG (if it i s not practical to assign the leakage to an individual SG, the entire primary to secondary leakage should be conservatively assumed to be from one SG) during the cycle preceding the inspection which is the subject of the report,

i.

The calculated accident induced leakage rate from the portion of the tubes below 1 5.21 inches from the top of the tubesheet for the most limiting accident in the most limiting SG. In addition, if the calculated accident induced leakage rate from the most limiting accident is less than 2. 1 6 times the maximum operational primary to secondary leakage rate, the report should describe how it was determined,

j.

The results of monitoring for tube axial displacement (slippage). If slippage is discovered, the implications of the discovery and corrective action shall be provided.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, D.C. 20555, with a copy to the Administrator, USNRC Region I within the time period specified for each report.

6.9.3 DELETED Insert: 1, 4, 6.9.4 When a report is required by ACTION 8 or 9 of Table 3.3-1 1 "Accident Monitoring Instrumentation", a report shall be submitted within the following 1 4 days. The report shall outline the preplan ned alternate method of monitoring for inadequate core cooling, the cause of the inoperability, and the plans and schedule for restoring the instrument channels to OPERABLE status.

SALEM - UNIT 1 6-24b Amendment No. 303

I NSTRUMENTATION ACCI DENT MONITORING I NSTRUM ENTATION LIMITING CON DITION FOR OPERATION 3.3.3.7 The accident monitoring instrumentation channels shown in Table 3.3-1 1 shall be operable.

APPLICABILITY: MODES 1, 2, and 3.

.----?

L---1 ACTION:

a.

As shown in Table 3.3-11.

SURVEILLANCE REQUIREMENTS 4.3.3.7 Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the C HANNEL C HECK, CHANNEL CALIBRATION and C HANNEL FUNCTIONAL TEST operations at the frequencies specified in the Surveillance Frequency Control Program unless otherwise noted in Table 4.3-1 1.

()liE:-------------------------------------------------------

Separate Condition entry is allowed for each Function.

SALEM - UNIT 2 3/4 3-50 Amendment No. 282

TABLE 3. 3-11 ACCIDENT MONITORING INSTRUMENTATION INSTRUMENT 1.

Reactor Coolant Out1et Temperature TBo:r (Wide Range}

2.

Reactor Cooant Inet Temperature -

Tcow (Wide Range) 3.

Reactor Cooant Pressure

{Wide Range) 4.

Pressurizer Water Level 5.

Steam L:ine Pressure 6.

Steam Generator Water Level (Narrow Range) 7.

Steam Generator Water Level (Wide Range) 8.

Refueling Water Storage Tank Water Level 9.

deeted 10. Auxi1.iary Feedwater Flow Rate 11.

12.

RGacter Csal;mt: Syst:em Subeeol:ing

¥oar¢a }1a23er y Deleted I PORV Pes1ea

  • Iadi:eai;;ez SALEM

- UNIT 2 Yoeleted I REQUIRED NO.

OF CHANNELS 2

2 2

2 2/Steam Generator 2/Steam Generator 4 (1 /Steam Generator) 2 4

(1/Stett Generator) 2 3/4 3-51 MINIMUM NO. OF CHANNELS ACTION 1

1, 2 1

1, 2 1

1, 2 1

1, 2 1/Steam Generator 1, 2 1/Steam Generator 1, 2 3

{1/Steam Generator) 1, 2 1

1, 2 3

(1/Steam Generator) 4, 6

+/-

J__,- 2 Amendment No. 206 I

TABLE 3. 3-1 1

( Continued)

ACCIDENT MONITORING INSTRUMENTATION rNsTRu""MENT

,Deleted ]

13.

PORV Block 'lalvc Position Indicator 14.

n....-..... _r:::w--"'!'_.. _.., W.; X--...-.-

co-.

__,f;_..... T UV \\7.....,1...,..,-,

n....,....; +-..::..........

-;:::::.==/::t "==*= L0::r=wz

""'"""'"' ca "0"'

Deleted I REQUIRED NO.

OF CH.llliNELS 2/valve**

2/ valoe**

15.

Containment Pressure Narre-;; Ral:'!:§'C 2

Deleted I 1 6.

Containment Pressure -

Wide Range 2

17.

ContaiPent Water Level Wide Range 2

1 8. Core Exit Thermocouples 4 /core quadrant 19.

Reactor Ve s s el Level Instrumentation Sys t em

{RVL I S )

2 0. Containment High Range Accident Radi ati on Monitor 2 1.

Main Steamline Di s charge

( Safety Valves and Atmospheri c Steam Dumps ) Monitor

/

f-+/-.:L)

Ter.:tpe+/-:atures SALEM -

UNIT 2

22. Wide Range Neutron Flux Monitors Auxiliary Feed Water Storage Tank (Condensate Storage Tank) Water Level 2

2 1/

M S Line 2

2 3 / 4 3-5 1 a MINIMUM NO.

OF CHANNELS 3..

+/-

+/-

1 1

2 / core quadrant 1

2 1 /

M S Line

Delete I 1

1 ACT IOK

)_

.2 S,

A

-+/--, i!

h 7 r 2 7, 2 1, 2 8, 9 1 0 1 0

.F

+-. 1 V.L I....I.J.C Amendment No.

2 6 3 1, 2 1, 2

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 submit a special report in accordance with Specification 6.9.4 ACTION 1 ACTION 2 ACTION 3 ACTION 4 ACTION 5 Wi th the n

than the Re restore the days,

or TABLE 3. 3-ll

( continued)

TABLE NOTATION ber of OPERABLE accident monitorin 1ired Number of Channels shown in noperable ohannel (s) to OPERABLE

  • n the nexe 2 channels less able 3. 3-11,

.atus within +

+

Wi th the numb 1r of OPERABLE accident monitoring channels less than the Mini \\un Number of Channels shown in Table 3. 3-11,

res tore the i parable ohannel (s) to OPERABLE status within 48 or be in HOT SHU'I'DOWN within the M-&t:l 71-2 hours.

If\\ r:::;-;--,

d t.

- 

deleted Yfollowing 6 I With the number of OPERABLE channels one

  • less than the Required Number of Channels shown in Table 3. 3-11,

operations may proceed provided that an OPERABLE Steam Generator Wide Range Level channel is available as an alternate means of indication for the Steam Generator wi th no OPERABLE Auxiliary Feedwater Flow Rate Channel deleted

OTHERWISE, restore the inoperable channel to OPERABLE status within 30 days, or submit a special report in accordance with Specification 6.9.4.

SALEM UNI'l' 2 3/4 3-51b Amendment No. 206

ACTION 6 ACTION 7 ACTION 8

  • ACTION 9.

ACTION 1 0 TABLE 3. 3-11 ( continued )

following 6 TABLE NOTATION 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 With the n ber of OPERABLE channel s less than the Mi ti.m:um Number of channels 3hown in Table 3. 3 -11,

restore the inopele channel ( s ) t OPERABLE status within 7 days,

or be in HOT SHUTDOWN within the amet 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

With the number o f OPERABLE channels one less than the Required Number o f Channels shown in Table 3. 3-11, operation may proceed unti l the next CHANNEL CALIBRATION (which shall be performed upon the next entry into MODE 5, COLD SHUTDOWN).

With one RVLIS channel inoperable,

restore the RVLIS channel to OPERABLE s ta tus within 30 days,

or submit a special report in accordance with Specification 6. 9. 4.

With both RVLIS channels inoperable,

restore one channel to OPERABLE s tatus within 7 days or submit a special report in accordance with Specification 6. 9. 4.

With the number of OPERABLE Channels less than required b¥ the minimum channels OPERABLE requirements,

initiate the preplanned alternat e method of monitoring the appropriate parameter within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />,

and :

1 ) either restore the inoperable Channel ( s ) to OPERABLE status within 7 days of the event,

or 2 ) prepare and submit a Special Report to the Commis s ion pursuant to Speci fication 6. 9. 2 within 14 days following the event outlining the actions taken,

the cause of the inoperability and the plans and schedule for res toring the sys tem to OPERABLE status.

SALEM -

UNIT 2 3 /4 3-Slc Amendment No. 253

TABLE 4.3-1 1 SURVEILLANCE REQUIREMENTS FOR ACCIDENT MONITORI NG INSTRUMENTATION CHANNEL CHANNEL CHANNEL FUNCTIONAL INSTRUMENT CHECKS<1l CALIBRATIQN<1l TEST<1l 1. Reactor Coolant Outlet Temperature -

N.A.

THor (Wide Range)

2.

Reactor Coolant Inlet Temperature -

N.A.

T coLD (Wide Range)

3. Reactor Coolant Pressure (Wide Range)

N.A.

4. Pressurizer Water Level N.A.
5. Steam Line Pressure N.A.
6. Steam Generator Water Level N.A.

(Narrow Range)

7. Steam Generator Water Level N.A.

(Wide Range)

8. Refueling Water Storage Tank Water Level N.A.
9. deleted 1 0. Auxiliary Feedwater Flow Rate S/U#

N.A.

1 1. ReaetoF Gaalant System Sc:eeoolin§ N.,.A; Maf§j 

Deleted Auxiliary Feedwater System is used on each startup and flow rate indication is verified at that time.

=Fhe-iftStfl:tments used to develop RGS sd:beoolin§ maf§in are ealibrateeHA-aeeefia.A.e.

'v'll'ith the Sc:rveillance Freqe:ency Gontrol Pro§ram; the monitor will be compared with calculated subcooling margin for knovm input values in accordance with the Surveillance Frequency Gontrol Pro§ramb 1'\\:



SALEM - UNIT 2 3/4 3-52 Amendment No. 282

TABLE 4.3-1 1 (Continued)

SURVEILLANCE REQUIREMENTS FOR ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHANNEL CHANNEL FUNCTIONAL INSTRUMENT CHECKS(1J CALIBRATION(1l TEST(1l 1 2 *

  • Deleted 1 3. PORV Block Valve Position Indicator 1 4. Pressurizer Safety Valve Position Indicator 1 5. Goffiainment Pressure Narro*N Rafl§-0 !Ljr::o-e-=-le-:t-ed-:-o

)Deleted I 1 6. Containment Pressure - Wide Range 1 7. Containment Water Level - Wide Range 1 8. Core Exit Thermocouples 1 9. Reactor Vessel Level Instrumentation System (RVLIS)

20. Containment High Range Accident Radiation monitor 21. Main Steamline Discharge (Safety Valves and Atmospheric Steam Dumps) Monitor Table Notation N.A N.A N.A N-A N.A.

N.A.

N.A.

N.A.

(1 )

Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.

I:Jnless the bloek valve is closed in order to meet the requirements of,l\\otion b, or o in

.-:-:-:--="'=-.

specificatioo-.a-A,§..;

fLj Delete 122. Wide Range Neutron Flux Monitors

23. Auxiliary Feed Water Storage Tank (Condensate Storage Tank) Water Level SALEM - UNIT 2 314 3-52a N.A.

N A.

Amendment No. 282

ADMINISTRATIVE CONTROLS

e.

Number of tubes plugged during the inspection outage for each deg radation mechanism,

f.

The number and percentage of tubes plugged to date, and the effective plugging percentage in each steam generator,

g.

The results of condition monitoring, including the results of tube pulls and in-situ testing.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Document Control Desk, Wash ington, D.C. 20555, with a copy to the Administrator, USNRC Region I within the time period specified for each report.

6.9.3 DELETED Insert: 1, 4, 6.9.4 When a report is required by ACTION 8 OR 9 of Table 3.3-1 1 "Accident Monitoring Instrumentation", a report shall be submitted with in the following 14 days. The report shall outline the preplan ned alternate method of monitoring for inadequate core cooling, the cause of the inoperability, and the plans and schedule for restoring the instrument channels to OPERABLE status.

SALEM - UNIT 2 6-24b Amendment No. 291

LR-N16-0003 LAR S16-01 Mark-up of Proposed Technical Specification Bases Pages The following Technical Specifications pages for Renewed Facility Operating License DPR-70 are affected by this change request:

Technical Specification Bases 3/4.3.3.7, Accident Monitoring I nstrumentation B 3/4 3-3 The following Technical Specifications pages for Renewed Facility Operating License DPR-75 are affected by this change request:

Technical Specification Bases 3/4.3.3. 7, Accident Monitoring I nstrumentation B 3/4 3-3 1

BASE S 3/ 4. 3. 3. 5 REMOTE SHUT DOWN INSTRUMENTAT I ON The OPERAB I L ITY of the remote shut down i n s t rumentat ion ensur e s that s u f f i c ient capab i l i t y i s ava i lable t o permi t shut down and maintenance of HOT S TANDBY of the f a c i l i t y from loca t i ons out s ide of the control room.

This c apabi l i t y i s required i n the event control room hab i t ab i l i t y i s lost and is cons i s tent with General D e s i gn Criteria 19 o f 10 C FR 50.

3 / 4. 3. 3. 6 T H I S SECT I ON DELETE D 3 / 4. 3. 3. 7 ACC I DENT MON I TORING INSTRUMENTAT ION The OPERAB I L ITY of the acc i dent moni t oring ins trumentat i on ensures that s u f f i c i e n t in format ion is ava i l ab l e on s e l e cted pl ant parame t e r s to monitor and a s s e s s the s e variab l e s f o l l owing an accident.

This capab i l i t y i s cons i s t ent with the Recommenda t i o n s o f Re gul atory Guide 1. 97, " I ns trumentation for L i ght-Water-Cooled Nuclear Powe r Plants to A s s e s s Pl ant Condi tions During and Fo l l owing an Accident, " December 1 9 7 5.

3 RAD I OACTIVE L I QU I D E FFLUENT MON I TORING INSTRUMENTAT I ON urpo s e of t an k l evel indicating devi c e s i s to a s s ure the detect ion and contr l of leaks that if not cont r o l l e d could potent i a l l y result in the t ransp rt of radioact ive ma t e r i a l s to UNRESTRICTED AREAS.

The Wide Range Neutron Flux Monitors are the Gamma-Metrics Post-Accident Neutron Monitors.

T H I S SECT I ON DELETED 3 / 4. 3. 3. 1 0 T H I S SECT I ON DELETED 3 / 4. 3. 3. 1 1 T H I S SECT I ON DELETED 3 / 4. 3. 3. 12 THI S SECTION DELETED 3 / 4. 3. 3. 1 3 TH I S SECT I ON DELETED SALEM -

UN I T 1 B 3 / 4 3-3 Amendment No. 282

( PSEG I s sued)

INSTRUMENTATION BASES I mmediate action(s), in accordance with the LCO Action Statements, means that the required action should be pursued without delay and in a controlled manner.

3/4.3.3.2 THIS SECTION DELETED 3/4.3.3.3 THIS SECTION DELETED 3/4.3.3.4 THIS SECTION DELETED 3/4.3.3.5 REMOTE SHUTDOWN I NSTRUMENTATION The OPERABILITY of the remote shutdown instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of HOT STANDBY of the facility from locations outside of the control room. This capability is required in the event control room habitability is lost and is consistent with General Design Criterion 1 9 of 1 0 CFR 50.

3/4.3.3.6 THIS SECTION DELETED 3/4.3.3.7 ACCIDENT MONITORING I NSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables following an accident. This capability is consistent with the Recommendations of Regulatory Guide 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," December 1 975 and N UREG-0578,"TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations."

\\\\..._The Wide Range Neutron Flux Monitors are the Gamma-Metrics Post-Accident Neutron Monitors.

SALEM - UNIT 2 B 3/4 3-3 Amendment No. 265 (PSEG Issued)

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