Text

License Amendment Request to Amend the Accident Monitoring Instrumentation Technical Specifications
	ML16323A279
Person / Time
Site:	Salem
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

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.

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.

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.

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.

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 10 CFR 50.90 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 Comm itment Tracking Coordinator

LR-N16-0003 LAR 516-01 Attachment 1 Evaluation of Proposed Changes Table of Contents

1.0 DESCRIPTION

.................................... . .................................... . . ..................................... 1 2.0 PROPOS E D CHANGE .......................... . .. . ........... . . ................. . ....... . ..... . ................. ........ 1

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

7.0 REFERENCES

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

LR-N16-0003 LAR S16-01

1.0 DESCRIPTION

The proposed amendment revises the TS 3/4. 3 . 3 . 7, Accident Monitoring I nstrumentation.

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

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 I ndicator*

Pressurizer Safety Valve Position Indicator*

Containm ent 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 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 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 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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.

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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT S HUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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 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.

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.

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-N16-00 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",

Revision 22 dated December 1980 and the N UREG 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:

(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.

( 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.

(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.

(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:

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).

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.

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

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.

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, i n 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 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 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months for less than minimum number of OPERABLE channels) that are overly restrictive and are not commensurate with the safety significance of the i nstruments.

4.0 TECHNICAL ANALYSIS

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:

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 s h o u ld be removed because they are not Type A variable or not Category 1 were identified .

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).

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 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 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 S M M 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 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 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.

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 .

6

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.

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 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 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.

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.

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.

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 .

7

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 .

Therefore, the PORV position indication is removed from the Salem Unit 2 T S 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.

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.

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.

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."

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. "

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.

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.

8

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.

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.

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 .

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.

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."

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.

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 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.

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 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 .

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 .

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).

10

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.

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.

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.

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.

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.

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.

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.

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

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.

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)

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 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.

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 ).

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 .

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 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:

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.

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.

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 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.

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 .

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.

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.

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 .

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.

13

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).

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.

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 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , respectively). Other m inor changes to the l isted actions are discussed below.

14

LR-N16-0003 LAR 516-01 Instrument Action

1. Reactor Coolant Outlet Temperature- THot (Wide Range) 1' 2

2. Reactor Coolant Outlet Temperature- Tcold (Wide Range) 1' 2

3. 1' 2 1' 2 Reactor Coolant Pressure (Wide Range)

4. Pressurizer Water Level

5. Steam Line Pressure 1' 2

6. Steam Generator Water Level (Narrow Range) 1' 2

7. Steam Generator Water Level (Wide Range) 1' 2

8. Refueling Water Storage Tank Water Level 1' 2

10. Auxiliary Feedwater Flow Rate 4,6

16. 7,2 17.

Containment Pressure -Wide Range 7,2 18.

Containment Water Level- Wide Range Core Exit Thermocouples 1' 2

22. Wide Range Neutron Flux Monitors 1' 2

23. AFW (Condensate) Storage Tank Water Level 1' 2 The following TS Table 3 . 3- 1 1 Actions are revised:

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.

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 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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 .

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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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.

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

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.

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 .

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 .

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;

1. Does the proposed change involve a significant increase in the probability or conseq uences of an accident previously evaluated?

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).

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.

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 .

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 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.

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 .

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.

17

LR*N16*0003 LAR 816-01

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?

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.

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.

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.

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.

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.

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.

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.

The proposed changes only m od ify the i n stru me nts listed in TS 3/4. 3 . 3 . 7 .

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,

( 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 .

18

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 .

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

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 ,

"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) , "

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:

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:

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

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.

APPL I CAB I LITY: M O DE S 1 , 2, and 3 .

ACT I O N :

a. As shown in T a b l e 3.3-1 1 .

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

1. Reactor Coolant Outlet Temperature - 2 1, 2 T Hor (Wide Range)

2. Reactor Coolant Inlet Temperature - 2 1, 2 T coLo (Wide Ran ge)

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 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)

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**

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

1 9. Reactor Vessel Level Instrumentation 2 1 8, 9 System (RVUS)

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

( 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-,

Delete SALEM - UNIT 1 Amendment No. 3 1 0

TABLE 3 . 3 - ll ( continued)

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 .

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 .

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.

II\

ACTION 5 deleted

OTH E RWISE, restore the inoperable channel to OPERABLE status with in 30 days, or submit a s pecial report in accordance with !-

Specification 6 . 9 .4 .

3 /4 3 - 5 6 Amendment 225 SALSM - UNIT l No .

H OT STA N D BY within the following 6 next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in TABLE 3 . 3 - 1 1 ( 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 .

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 ) .

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 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 .

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 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and :

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 .

SALEM - UNIT 1 3 / 4 3 -5 6a Amendment No . 2 72

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

1. Reactor Coolant Outlet Temperature - N.A.

THOT (Wide Range)

2. Reactor Coolant I nlet Temperature - N.A.

TcoLo (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.

N:M N-A MargiA MeAiter y 1 1 . Reaeter GeelaAt SysteFA Sl:leeeeliA§ 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-: .-------.

SALEM - U N IT 1 3/4 3-57 Amendment No. 299

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

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

- ed....,....,

1 5. GGmaemHffiGe 1 6 . Containment Pressure - Wide N.A.

1 7. Containment Water Level - Wide Range N .A.

1 8. Core Exit Thermocouples N. A .

1 9. Reactor Vessel Level Instrumentation N.A .

System (RVLIS)

20. Containment High Range Accident Radiati on Monitor 2 1 . Main Steamline 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.

. R--..,...--I- ----------------------N -

, 2-t 2-. WJd e a n g e N e u t ro n F ux Mo- n or-s--

it- .A.

2 3 . Auxiliary Feed Water Storage Tan k (Condensate Storage Tank) Water Level N .A .

Unless the blosk-valve is closed in order to meet the ration b, or o In speGifieation aA

'3' Delete SALEM - UNIT 1 3/4 3-57a Amendment No. 3 1 0

AQMINISTRATIVE CONTROLS

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,

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,

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.

I nsert: 1 , 4 ,

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.

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

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.

AP PLICAB I L ITY: MODES 1, 2, and 3.

ACT I O N :

a. A s shown in Table 3.3-11.

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 .

()li E: -------------------------------------------------------

Separate Condition entry is allowed for each Function.

L--- 1 SALEM - UNIT 2 3/4 3-50 Amend ment N o . 282

TABLE 3 . 3-11 ACCIDENT MONITORING INSTRUMENTATION REQUIRED MINIMUM NO . OF NO . OF INSTRUMENT CHANNELS CHANNELS ACTION

1. Reactor Coolant Out1et Temperature - 2 1 1, 2 TBo:r (Wide Range}

2. Reactor Cooant Inet Temperature - 2 1 1, 2 Tcow (Wide Range)

3. Reactor Cooant Pressure {Wide Range) 2 1 1, 2

4. Pressurizer Water Level 2 1 1, 2

5. Steam L:ine Pressure 2 /Steam Generator 1/Steam Generator 1, 2

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

7. S team Generator Water Level (Wide 4 (1 /Steam Generator) 3 {1/Steam Generator) 1, 2 I Range)

8. Refueling Water Storage Tank Water 2 1 1, 2 Level

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:

¥oar¢a }aer . y Deleted I 2

12 . PORV Pes1ea

Iadi:eai;;ez +/- J__,- 2 SALEM - UNIT 2 Yoeleted I 3 / 4 3-51 Amendment No . 206

TABLE 3 . 3- 1 1 ( C ontinued)

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.. ;)_

.2 14 . n ....- ...... _r:::w--"'!'_.._.., .; --...-. -

==

:t "==*= L::r=wz co-.__,f;_....... \7......, 1 ... ,..,-, n ..... ,.... .; +- ..:: ......

2/ v al oe** +/- -+/--, i!

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

Deleted I 15 . Containment Pressure Narre-;; Ral:'!:§'C 2 +/- h Deleted I 2 1 2

16. Cont ainment P r e s s ure - Wide Range 7 r 17 . ContaiPent Water Level - Wide Range 2 1 7, 2

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 )

20. Containment High Range Accident 2 2 10 Radi a ti on Monitor

21. Main Steamline 1/ 1/ 10 Di s ch a rge ( Safety MS Line M S Line Dump s )

Va lves and Atmo spheri c S t e am Monitor

/

f-+/-.:L) .F V .L

+-. 1 I.... .I. J.C Ter.:tpe+/-:atures Delet e I SALEM - UNIT 2 3/4 3-5 1 a Amendment No . 263

22. Wid e Range Neutron Flux Monitors 2 1 1, 2 Auxil iary Feed Water Stora g e Tank 2 1 1, 2

( Condensate Storage Tank) Water Level

HOT STA N D BY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in submit a s pecial report in accorda nce with Specification 6 . 9 .4 TABLE 3 . 3 - l l ( continued)

TABLE NOTATION Wi th the n ber Channels able ACTION 1 o f OPERABLE accident moni torin channels less 3 . 3-11 ,

ohannel ( s ) to than the Re 1ired Number of shown in res tore the noperable OPERABLE .atus within +

days , or

n the nexe 2 ACTION 2 Wi th the numb 1r of OPERABLE acci dent moni toring channels less

\un Number of Channels shown in Table 3 . 3-11 ,

ohannel ( s )

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:::;-;- ---,

ACTION 3 deleted Yfollowing 6 I ACTION 4 Wi th the number of OPERABLE channels one

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

OTH E RWI SE, restore the inoperable channel to OPERABLE status within 30 days, or submit a special report i n accordance with

S pecification 6 . 9 .4 .

SALEM - UNI'l' 2 3/4 3 - 5 1b Amendment No . 206

3 . 3 -11 ( continued ) H OT STAN DBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 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 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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 8 With one RVLIS channel inoperable , restore the RVLI S channel t o accordance with Specif icat ion 6 . 9 . 4 .

OPERABLE s ta tus wi thin 30 days , or submit a spe c i a l report i n

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 .

ACTION 1 0 With the number o f O PERABLE Channels less than required b¥ 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 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , 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 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 .

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.

T Hor (Wide Range)

2. Reactor Coolant Inlet Temperature -

Tc o LD (Wide Range)

N.A.

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. Stea m 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.

11. ReaetoF Gaalant System S: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 s:beoolin§ maf§in are ealibrateeHA-aeeeffia.A.ee.

'v'll'ith the S:rveillance Freq: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§ram 1'\:

SALEM - U N IT 2 3/4 3-52 Amendment No. 282

TABLE 4.3-1 1 (Contin ued)

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 !

1 4. Pressurizer Safety Valve Position Indicator N.A r::-e-=-le-:t-1 5 . Gofifia nment Pressure Narro*N Rafl§-0 N-A

!Lj o ed-:-o

. Deleted I 1 6. Containment Pressure - Wide Range N .A.

1 7. Containment Water Level - Wide Range N .A.

1 8. Core Exit Thermocouples N .A.

1 9. Reactor Vessel Level I nstrumentation N.A.

System (RVLIS)

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 .

! 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 N eutron Flux Monitors N .A.

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

SALEM - U N IT 2 314 3-52a Amendment No. 282

ADM I NISTRATIVE CONTROLS

e. Number of t ubes plugged d u ring the inspection outage for each deg radation mechanism ,

f. The n u m ber and percentage of tubes plugged to date, and the effective plugg ing percentage in each steam generator,

g. The results of condition monitoring, including t h e results o f tube p ulls 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.

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.

SALEM - U NIT 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:

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 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 .

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 ,

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 .

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 .

The Wide Range Neutron F lux Mon itors are the Gamma-Metrics Post-Accident Neutron Monitors.

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

( 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.

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 . 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."

\\...__ The Wide Range N eutron F lux Monitors are the Gamma-Metrics Post-Accident N eutron Monitors .

SALEM - U N I T 2 B 3/4 3-3 Amendment No. 265 (PSEG I ssued)

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

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7.0 REFERENCES

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2.0 PROPOSED CHANGE

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

Text

1.0 DESCRIPTION

3.0 BACKGROUND

7.0 REFERENCES

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

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