Text

Request for Changes to Technical Specifications Relocation of Response Time Testing Time Limits to the Updated Final Safety Analysis Report
	ML052420546
Person / Time
Site:	Salem
Issue date:	08/19/2005
From:	Joyce T; Public Service Enterprise Group
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	+kBR1SISP20051118, GL-93-008, IN-97-028, LCR S05-04, LR-N05-0336
	Download: ML052420546 (57)
	v • d • e

PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 LR-N05-0336 LCR S05-04 0 PSEG AUG 1 9 2005 NuclearLLC U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 REQUEST FOR CHANGES TO TECHNICAL SPECIFICATIONS RELOCATION OF RESPONSE TIME TESTING TIME LIMITS TO THE UPDATED FINAL SAFETY ANALYSIS REPORT SALEM NUCLEAR GENERATING STATION UNITS I and 2 FACILITY OPERATING LICENSES DPR-70 and DPR-75 DOCKET NOs. 50-272 and 50-311 Pursuant to 10 CFR 50.90, PSEG Nuclear LLC (PSEG) hereby requests a revision to the Technical Specifications (TS) for the Salem Nuclear Generating Station, Units 1 and

2. In accordance with 10 CFR 50.91 (b)(1), a copy of this submittal has been sent to the State of New Jersey.

PSEG Nuclear proposes to revise the Salem Unit 1 and 2 Technical Specifications to reflect the relocation of Response Time Testing Tables to the Updated Final Safety Analysis report in accordance with NRC guidance provided in Generic Letter 93-08, Relocation of Technical Specification Tables of Instrument Response Time Limits, dated December 29, 1993. Beaver Valley Power Station was issued a similar amendment dated January 20, 1998 (TAC Nos. M99671 and M99672).

PSEG is also revising the Definition of Engineered Safety Feature Response Time and Reactor Trip System Response Time using the guidance of Improved Technical Specifications (NUREG 1431), as modified by TSTF-111 and NRC Information Notice 97-28, Elimination of Instrument Response Time Testing under the Requirements of 10 CFR 50.59, dated May 30,1997.

PSEG has evaluated the proposed changes in accordance with 10 CFR 50.91 (a)(1),

using the criteria in 10 CFR 50.92 (c), and has determined this request involves no sig-nificant hazard considerations. This amendment to the Salem TS meets the criteria of 10 CFR 51.22 (c)(9) for categorical exclusion from an environmental impact statement.

The requested changes are provided in Attachment I to this letter. The proposed marked up Technical Specification pages are provided in Attachment 2. Attachment 3 contains the UFSAR pages, as they will appear in the Salem UFSAR following approval of this request. c) 3 95-2168 REV. 7/99

AUG 1 9 2005 Document Control Desk 2 LR-N05-0336 Should you have any questions regarding this request, please contact Steve Mannon at 856-339-1129.

I declare under penalty of perjury that the foregoing is true and correct.

Sincerely, Executed on _____________1___ _ ____la Thomas P. Joyce Site Vice-President Salem Units 1 and 2 Attachments (3) cc Mr. S. J. Collins, Administrator - Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 U. S. Nuclear Regulatory Commission ATTN: Mr. Stewart Bailey, Project Manager - Salem and Hope Creek Stations Mail Stop 08B1 Washington, DC 20555-0001 USNRC Senior Resident Inspector - Salem (X24)

Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering PO Box 415 Trenton, NJ 08625

EVALUATION OF REVISIONS TO THE TECHNICAL SPECIFICATIONS RELOCATION OF RESPONSE TIME TESTING TIME LIMITS TO THE UPDATED FINAL SAFETY ANALYSIS REPORT SALEM NUCLEAR GENERATING STATION UNITS I & 2 FACILITY OPERATING LICENSES DPR-70 AND DPR-75 DOCKET NOs. 50-272 AND 50-311 ATTACHMENT I DESCRIPTION AND EVALUATION OF REQUESTED CHANGES

I EVALUATION OF REVISIONS TO THE TECHNICAL SPECIFICATIONS RELOCATION OF RESPONSE TIME TESTING TIME LIMITS TO THE UPDATED FINAL SAFETY ANALYSIS REPORT Table of Contents

1. DESCRIPTION ........................................... 2

2. PROPOSED CHANGES ........................................... 2

3. EVALUATION ........................................... 3

4. REGULATORY SAFETY ANALYSIS ......................................... 5 4.1 No Significant Hazards Consideration ........................................... 5 4.2 Applicable Regulatory Requirements/Criteria ........................................... 6

5. ENVIRONMENTAL ASSESMENT/IMPACT STATEMENT ................................... 7

6. REFERENCES ........................................... 8 EVALUATION OF REVISIONS TO THE TECHNICAL SPECIFICATIONS RELOCATION OF RESPONSE TIME TESTING TIME LIMITS TO THE UPDATED FINAL SAFETY ANALYSIS REPORT

1.0 DESCRIPTION

PSEG requests changes to the Salem Units 1 and 2 Technical Specifications (TS). The requested changes would relocate the Reactor Trip System (RTS) and Engineered Safety Feature Actuation System (ESFAS) response times from TS Tables 3.3-2 and 3.3-5 to the Updated Final Safety Analysis Report (UFSAR). Neither the response time limits nor the surveillance requirements for performing response time testing would be altered by these proposed changes. Future changes to the response time limits included in the UFSAR will be controlled in accordance with the requirements of 10 CFR 50.59. De-letion of Response Time Testing Requirements will require prior NRC ap-proval. In addition, a change to the Definition of Reactor Trip System (RTS) and Engineered Safety Feature Actuation System (ESFAS), will require prior NRC review and approval of any methodology used to verify the response times for selected components in lieu of measuring them.

Editorial changes to Tables 7.2-5, Item 14 and Table 7.2-5A, Item 14, are in-cluded to correct the Initiating Signal from Station Blackout to Loss of Offsite Power, which is the correct terminology.

2.0 PROPOSED CHANGE

S The following TS pages are affected by this request and the appropriate mark-ups are included in Attachment 2:

Salem Unit 1 1-3 Definition 1.12 1-6 Definition 1.26 3/4 3-1 Reactor Trip System Instrumentation 3/4 3-9 Table 3.3-2 3/4 3-10 Table 3.3-2 3/4 3-14 Engineered Safety Feature Actuation System Instrumentation 3/4 3-27 Table 3.3-5 3/4 3-28 Table 3.3-5 3/4 3-29 Table 3.3-5 3/4 3-30 Table 3.3-5 3/4 3-31 Table 3.3-5 Notations Salem Unit 2 1-3 Definition 1.12 1-6 Definition 1.26 3/4 3-1 Reactor Trip System Instrumentation 3/4 3-9 Table 3.3-2 3/4 3-10 Table 3.3-2 3/4 3-14 Engineered Safety Feature Actuation System Instrumentation 3/4 3-28 Table 3.3-5 3/4 3-29 Table 3.3-5 EVALUATION OF REVISIONS TO THE TECHNICAL SPECIFICATIONS RELOCATION OF RESPONSE TIME TESTING TIME LIMITS TO THE UPDATED FINAL SAFETY ANALYSIS REPORT 3/4 3-30 Table 3.3-5 3/4 3-31 Table 3.3-5 3/4 3-32 Table 3.3-5 Notations No changes to the applicable Bases are proposed in accordance with the guid-ance of Generic Letter 93-08 and Improved Standard Technical Specifications.

Attachment 3 contains the UFSAR pages representing the proposed relocation of the Response Time Limits Tables.

Table 7.2.4, Salem Unit 1 Reactor Trip System Instrumentation Response Times Table 7.2.5, Salem Unit 1 Engineered Safety Features Response Times Table 7.2.4A, Salem Unit 2 Reactor Trip System Instrumentation Response Times Table 7.2.5A, Salem Unit 2 Engineered Safety Features Response Times Page 7.3-26 Chapter 7, Tables, Page 7-vi 3.0 EVALUATION The Commission's regulatory requirements related to the content of TS are set forth in 10 CFR 50.36. That regulation provides four criteria for determining whether particular limiting conditions for operation are required to be included in the TS. These are:

1. Installed instrumentation that is used to detect, and indicate in the con-trol room, a significant abnormal degradation of the reactor coolant pres-sure boundary;

2. A process variable, design feature, or operating restriction that is an ini-tial 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;

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;

4. A structure, system, or component which operating experience or prob-abilistic risk assessment has shown to be significant to public health and safety.

Existing TS limiting conditions for operation, which do not satisfy these four specified criteria, may be relocated to the UFSAR, such that future changes could be made to these provisions pursuant to 10 CFR 50.59.

NRC Generic Letter (GL) 93-08, "Relocation of Technical Specification Tables of Instrument Response Time Limits," dated December 29, 1993, provides guidance to licensees proposing to relocate RTS and ESFAS instrument re-EVALUATION OF REVISIONS TO THE TECHNICAL SPECIFICATIONS RELOCATION OF RESPONSE TIME TESTING TIME LIMITS TO THE UPDATED FINAL SAFETY ANALYSIS REPORT sponse time limits from the TS to the UFSAR. GL 93-08 provides that reloca-tion of the RTS and ESFAS instrument response time limits from the TS to the UFSAR should not alter the surveillance requirements. After relocation, the UFSAR will contain the response time limits for the RTS and ESFAS instru-ments, including those channels for which the response time limit is indicated as "NA"; that is, a response time is not applicable. The UFSAR also clarifies response time limits where footnotes are included in the tables that describe how those limits are applied. The limiting condition for operation (LCO) for the RTS and ESFAS instruments is modified to delete the phrases "with RE-SPONSE TIMES as shown in Table 3.3-2 (RTS) or 3.3-5 (ESFAS)" so as to simply state that this instrumentation "shall be OPERABLE." Although the sur-veillance requirements for the RTS and ESFAS instrument response time limits, do not reference the tables containing these limits and, therefore, do not need to be modified to implement this change, a footnote on TS Table 3.3-2 states that neutron detectors are exempt from response time testing. To retain this exception, which is stated in TS Table 3.3-2 (being removed from the TS by this amendment), the RTS surveillance requirements is modified to add the follow-ing statement: "Neutron detectors are exempt from response time testing."

The change to the Definition of Reactor Trip System (RTS) and Engineered Safety Feature Actuation System (ESFAS), will require prior NRC review and approval of any methodology used to verify the response times for selected components in lieu of measuring them. The additional wording proposed is in-cluded in the Improved Standard Technical Specifications and was incorpo-rated as part of TSTF-1 11.

The proposed changes relocate the RTS and ESFAS instrument response time limits from the TS to the UFSAR, which do not alter the surveillances for these instruments or change any of the response time limits, including those channels for which the response time limit is indicated as NA. The clarifications provided in the applicable TS footnotes describing how the response time limits are to be applied will also be relocated to the UFSAR. Any future changes to the RTS and ESFAS instrument response time limits will be performed in accordance with the requirements of 10 CFR 50.59. The proposed changes also delete from the LCO, the phrase "with response TIMES as shown in Table 3.3-2 (RTS) or 3.3-5 ESFAS)" so as to simply state "shall be OPERABLE". The sur-veillance requirements for the RTS are revised to include the footnote "Neutron detectors are exempt from response time testing", which was previously in-cluded on TS Table 3.3-2. These proposed changes are consistent with the guidance provided in GL 93-08.

EVALUATION OF REVISIONS TO THE TECHNICAL SPECIFICATIONS RELOCATION OF RESPONSE TIME TESTING TIME LIMITS TO THE UPDATED FINAL SAFETY ANALYSIS REPORT 4.0 REGULATORY SAFETY ANALYSIS 4.1 Basis for proposed no significant hazards consideration determination As required by 10 CFR 50.91 (a), PSEG provides its analysis of the no significant hazards consideration. According to 10 CFR 50.92(c), a pro-posed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated;

2. Create the possibility of a new or different kind of accident from any previously evaluated; or

3. Involve a significant reduction in a margin of safety.

The determinations that the criteria set forth in 10 CFR 50.92 are met for this amendment request are indicated below:

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

Response: No The proposed amendment relocates the instrument response time lim-its for the reactor trip system (RTS) and engineered safety feature ac-tuation system (ESFAS) from the technical specifications to the Up-dated Final Safety Analysis Report (UFSAR). The proposed amendment conforms to the guidance given in Enclosures 1 and 2 of Generic Letter 93-08. Neither the response time limits nor the surveil-lance requirements for performing response time testing will be altered by this submittal. The overall RTS and ESFAS functional capabilities will not be changed and assurance that action requirements of the re-actor trip and engineered safety features systems are completed within the time limits assumed in the accident analyses is unaffected by the proposed amendment.

Therefore, operation of the facility in accordance with the proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Create the possibility of a new or different kind of accident from any accident previously evaluated.

Response: No EVALUATION OF REVISIONS TO THE TECHNICAL SPECIFICATIONS RELOCATION OF RESPONSE TIME TESTING TIME LIMITS TO THE UPDATED FINAL SAFETY ANALYSIS REPORT The proposed amendment will not change the physical plant or the modes of plant operation defined in the operating license. The change does not involve the addition or modification of equipment nor does it alter the design or operation of plant systems.

Therefore, operation of the facility in accordance with the proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the change involve a significant reduction in a margin of safety?

Response: No The measurement of instrumentation response times at the frequen-cies specified in the technical specification provides assurance that ac-tions associated with the reactor trip and engineered safety features systems are accomplished within the time limits assumed in the acci-dent analyses. The response time limits and the measurement fre-quencies remain unchanged by the proposed amendment.

There will be no effect on the manner in which safety limits or limiting safety system settings are determined nor will there be any effect on those plant systems necessary to assure the accomplishment of pro-tection functions.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on this review, it is concluded that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, PSEG proposes that a finding of 'no significant hazards consideration" is justified.

4.2 Applicable Regulatory Requirements/Criteria The regulatory bases and guidance documents associated with the amendment request include the general design criteria that were followed in the design of the Salem Station which are the Atomic Industrial Forum (AIF) version, as published in a letter to the Atomic Energy Commission from E. A. Wiggin, Atomic Industrial Forum, dated October 2, 1967. In addition to the AIF General Design Criteria, the Salem Generating Station (SGS) was designed to comply with Public Service Electric & Gas (PSE&G's) understanding of the intent of the AEC's proposed General Design Criteria, as published for comment by the AEC in July, 1967. The application of the AEC's proposed General Design Criteria to the Salem Station is discussed in UFSAR Section 3.1.2.

EVALUATION OF REVISIONS TO THE TECHNICAL SPECIFICATIONS RELOCATION OF RESPONSE TIME TESTING TIME LIMITS TO THE UPDATED FINAL SAFETY ANALYSIS REPORT No changes to the RTS or ESFAS instrumentation design are requested, thus there would be no adverse impact to the General Design Criteria described above.

10CFR 50.36 Technical Specifications The Commission's regulatory requirements related to the content of TS are set forth in 10 CFR 50.36. That regulation provides four criteria for determining whether particular limiting conditions for operation are required to be included in the TS.

Existing TS limiting conditions for operation which do not satisfy the four 10 CFR 50.36 criteria may be relocated to the UFSAR, such that future changes could be made to these provisions pursuant to 10 CFR 50.59.

Based on the evaluation, PSEG believes that the proposed TS changes do not reduce the level of safety currently maintained by the TS and are in accor-dance with 10 CFR 50.36.

CONCLUSION PSEG has concluded, based on the considerations discussed above, that: (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 issu-ance of the amendments will not be inimical to the common defense and secu-rity or to the health and safety of the public.

5.0 ENVIRONMENTAL ASSESSMENT/IMPACT STATEMENT Pursuant to 10 CFR 51.22 (b), an evaluation of this license amendment re-quest has been performed to determine whether or not it meets the criteria for categorical exclusion set forth in 10 CFR 51.22 (c)(9) of the regulations.

The proposed amendment does not change a requirement with respect to in-stallation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and change surveillance requirements. PSEG has determined that the proposed amendment involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. Accordingly, the amendment re-quest meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issu-ance of the amendments.

EVALUATION OF REVISIONS TO THE TECHNICAL SPECIFICATIONS RELOCATION OF RESPONSE TIME TESTING TIME LIMITS TO THE UPDATED FINAL SAFETY ANALYSIS REPORT Therefore, it has been determined that there is:

1. No significant hazards consideration,

2. No significant change in the types, or significant increase in the amounts, of any effluents that may be released offsite, and

3. No significant increase in individual or cumulative occupational radiation exposures involved.

Therefore, this amendment request to the Salem Technical Specifications meets the criteria of 10 CFR 51.22 (c)(9) for categorical exclusion from an environmental impact statement.

6.0 REFERENCES

6.1 Code of Federal Regulations, General Design Criteria and 10 CFR 50.36.

6.2 PSEG Salem Units 1 and 2, Updated Final Safety Analysis Report.

6.3 PSEG Salem Units 1 and 2, Technical Specifications.

6.4 NRC Generic Letter 93-08, Relocation of Technical Specification Tables of Instrument Response Time Limits.

6.5 Beaver Valley Power Station Amendments 210 and 88 (TAC Nos M99671 and M99672) dated January 20,1998.

6.6 NRC Information Notice 97-28, Elimination of Response Time Testing under the Requirements of 10 CFR 50.59.

6.7 NUREG 1431, TSTF-111, Revise Bases for SRs 3.3.1.16 and 3.3.2.10 to eliminate pressure sensor response time testing.

EVALUATION OF REVISIONS TO THE TECHNICAL SPECIFICATIONS RELOCATION OF RESPONSE TIME TESTING TIME LIMITS TO THE UPDATED FINAL SAFETY ANALYSIS REPORT SALEM NUCLEAR GENERATING STATION UNITS I AND 2 FACILITY OPERATING LICENSES DPR-70 AND DPR-75 DOCKET NOs. 50-272 AND 50-311 ATTACHMENT 2 TECHNICAL SPECIFICATIONS MARKED-UP CHANGES

EVALUATION OF REVISIONS TO THE TECHNICAL SPECIFICATIONS RELOCATION OF RESPONSE TIME TESTING TIME LIMITS TO THE UPDATED FINAL SAFETY ANALYSIS REPORT Attachment 2 Salem Unit I Affected Pages 1-3 Definition 1.12 1-6 Definition 1.26 3/4 3-1 Reactor Trip System Instrumentation 3/4 3-9 Table 3.3-2 3/4 3-10 Table 3.3-2 3/4 3-14 Engineered Safety Feature Actuation System Instrumentation 3/4 3-27 Table 3.3-5 3/4 3-28 Table 3.3-5 3/4 3-29 Table 3.3-5 3/4 3-30 Table 3.3-5 3/4 3-31 Table 3.3-5 Notations

DEFINITIONS thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844 "Calculation of Distanice Factors for Power and Test Reactor Sites."

E - AVERAGE DISINTEGRATION ENERGY 1.11 E shall be the average (wei1gtda in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half-lives greater than 15 minutes, making up at least 95t of the total non-iodine activity in the coolant.

ENGINEERED SAFETY FEATURE RESPONSE TIME 1.12 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.

\ lIn lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.

FREQUENCY NOTATION 1.13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

FULLY WITHDRAWN 1.13a FULLY WITHDRAWN shall be the condition where control and/or shutdown banks are at a position which is within the interval of 222 to 228 steps withdrawn, inclusive. FULLY WITHDRAWN will be specified in the current reload analysis.

GASEOUS RADWASTE TREATMENT SYSTEM 1.14 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

IDENTIFIED LEAKAGE 1.15 IDENTIFIED LEAKAGE shall be:

a. Leakage (except Reactor Coolant Pump Seal Water Injection) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or SALEM - UNIT 1 1-3 Amendment No. 3II"

DEFINITIONS REACTOR TRIP SYSTEM RESPONSE TIME 1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until loss of stationary gripper coil-voltage.

In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.

REPORTABLE EVENT 1.27 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10CFR Part 50.

SHUTDOWN MARGIN 1.28 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which.the reactor is subcritical or would be subcritical from its present condition assuming all full length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be FULLY WITHDRAWN.

SITE BOUNDARY 1.29 The SITE BOUNDARY shall be that line beyond which the land is not owned, leased, or otherwise controlled by the licensee, as shown in Figure 5.1-3, and which defines the exclusion area as shown in Figure 5.1-1.

SOLIDIFICATION 1.30 Not Used SOURCE CHECK 1.31 SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.

STAGGERED TEST BASIS 1.32 A STAGGERED TEST BASIS shall consist of:

a. A test schedule for (n) systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into (n) equal subintervals.

SALEM - UNIT 1 1-6 Amendment No.

3/4.3 INSTRUMENTATION 3/4 .3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1.1 As a minimum, the reactor trip system instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE, ith RDEPONJCD TIFan a_ Thu in TahZc-'

APPLICABILITY: As shown in Tabie 3.3-1.

ACTION:

As shown in Table 3.3-1.

SURVEILLANCE REQUIREMENTS 4.3.1.1.1 Each reactor trip system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the MODES and at the frequencies shown in Table 4.3-1.

4.3.1.1.2 The logic for the interlocks shall be demonstrated OPERABLE prior to each reactor startup unless performed during the preceding 92 days. The total interlock function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by interlock operation.

4.3.1.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip function shall be verified to be within its limit at least once per 18 months.PEacy--

AV verification shall include at least one logic train such that both logic trains are verified at least once per 36 months and one channel per function such that all channels are verified at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function as shown in the "Total No. of Channels" column of Table 3.3-1.

maPE Ae dee --- ekiY- %

+ 0 .... .................................. .....

SALEM - UNIT-I 3/4 3-1 Amendment No. No-

( ~~~REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE ITEMSA FUNCTIONAL UNIT RESPONSE TIME

1. Manual Reactor Trip NOT APPLI

2. Power Range, Neutron Flux S O ds*

3. Power Range, Neutron Flux, PPLICABLE High Positive Rate

4. Power Range, Neutron Flux, / 0.5 seconds*

High Negative Rate

5. Intermediate Range, Neutron Flux NOT APPLICABLE

6. Source Range, Neutron Flux NOT APPLICABLE

7. Overtemperature AT / 5.75 seconds*

8. Overpower AT NOT APPLICABLE

9. Pressurizer Pressure--Low / 2.0 seconds

10. Pressurizer Pressu -High

2.0 seconds

11. Pressurizer W r Level--High NOT APPLICABLE

Neutro etectors are exempt from response time testing. Response time of the neutron flux signa portion of the channel shall be measured from detector output or input of first ele ronic component in channel.

-ALMPASTLEST L43W ltrnaei No.Au

- SALEM - UNIT. 1 3/4 3-9 Amendment No.

{ ~REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES/

FUNCTIONAL UNIT RESPONSE TIME

12. Loss of Flow - Single Loop (Above P-8)

1.0 seconds

13. Loss of Flow - Two Loops (Above P-7 and below P-8) S 1.0 s s

14. Steam Generator Water Level-- Low-Low conds

15. Deleted

16. Undervoltage-Reactor Coolant Pumps / 1.2 seconds

17. Underfrequency-Reactor Coolant Pumps / 0.6 seconds

18. Turbine Trip A. Low Fluid Oil Pressure NOT APPLICABLE B. Turbine Stop Valve NOT APPLICABLE

19. Safety Injection Input f SF NOT APPLICABLE

20. Reactor Coolant Pu reaker Position Trip NOT APPLICABLE

21. Reactor Trip B kers NOT APPLICABLE

22. Automat ip Logic NOT APPLICABLE SALEM -UNIT 1 ax - 3/4 3 Amendment No.

INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2.1 The Engineered Safety Feature Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4,z2-w -h P r 5'"IC T. Ib . T+/-1_. i+ C APPLICABILITY: As shown in Table 3.3-3. \ ^A ACTION:

a. With an ESFAS instrumentation channel trip setpoint less.conserva-tive than the value shown in the Allowable Values column of Table 3.3-4, declare the channel inoperable and apply the applicable ACTION requirement of Table 3.3-3 until the channel is restored to OPERABLE status with the trip setpoint adjusted consistent with the Trip Setpoint value.

b. With an ESFAS instrumentation channel inoperable, take the ACTION shown in Table 3.3-3.

SURVEILLANCE REQUIREMENTS 4.3.2.1.1 Each ESFAS instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the MODES and at the frequencies shown in Table 4.3-2.

4.3.2.1.2 The logic for the interlocks shall be demonstrated OPERABLE during the automatic actuation logic test. The total interlock function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by interlock operation.

4.3.2.1.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be verified to be within the limit at least once per 18 months. Each verification shall include at least one logic train such that both logic trains are verified at least 6nce per 36 months-&nd one channel per function such that all channels are verified at least once per N times 18 months where N is the total number of redundant channels in a specific ESFAS function as shown in the "Total No. of Channels" Column of Table 3.3-3. The provisions of Specification I

4.0.4 are not applicable to MSIV closure time testing. The provisions of Specification 4.0.4 are not applicable to the turbine driven auxiliary feedwater pump provided the surveillance is performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the secondary steam generator pressure is greater than 680 psig.

SALEM - UNIT 1 3/4 3-14 Amendment No.

/ ~TABLE 3.3-5 t ~ENGINEERED SAFETY FEATURES RESPONSE ITEMS/

F INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SEOND

1. Manual

a. Safety Injection (ECCS) Not Ap icable Feedwater Isolation Not plicable Reactor Trip (SI) No Applicable Containment Isolation-Phase "A" t Applicable Containment Ventilation Isolation Not Applicable Auxiliary Feedwater Pumps Not Applicable Service Water System Not Applicable Containment Fan Cooler Not Applicable

b. Containment Spray Not Applicable Containment Isolation-Phase "B" Not Applicable Containment Ventilation Isolat n Not Applicable

c. Containment Isolation-Phas "A" Not Applicable Containment Ventilation solation Not Applicable

d. Steam Line Isolatio Not Applicable

/ 2. Containment PressurHg

- a. Safety I ection (ECCS) . _27.0(1)

b. React Trip (from SI) *2.0 e terIsolation 10.0

d. ontainment Isolation-Phase "A" *17.0(2)/27.0(3)

e. Containment Ventilation Isolation Not Applicable Auxiliary Feedwater Pumps *60

g. Service Water System *13.0(2)/45.0(3)

h. Containment Fan Coolers *60.0 (7)

SALEM - UNIT 1 3/4 3-27 Amendment No.

EPA LEFT BLAA4I 1tAIgTltTrALLY I

/ TABLE 3.3-5 (Continued)

/ ~~ENGINEERED SAFETY FEATURES RESPONSE-TIMES VO INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SE

3. Pressurizer Pressure-Low

a. Safety Injection (ECCS) S 27.01"l

b. Reactor Trip (from SI) S 2.0 C. Feedwater Isolation < 10.0

d. Containment Isolation - Phase "A" S I t2

e. Containment Ventilation Isolation t Applicable

f. Auxiliary Feedwater Pumps /

\ . Service Water System 4 .0tU'/13 .0121

4. Differential Pressure Between Steam ines ih

a. Safety Injection (ECCS) / 12 .012)/ 22 .0*3o

b. Reactor Trip (from SI) , 2.0

c. Feedwater Isolation / 10.0

d. Containment Isolation Phase "A"

17.012./27.0I3

e. Containment Ventila on Isolation Not Applicable

f. Auxiliary Feedwat Pumps

60

g. Service Water stem

13.0(2)/48.013)

5. Steam Flow in Two eam Lines -High Coincident wgith Tas - Iw-Low'K

a. Safet Injection (ECCS) 5 15.75(2)/25.75(31

b. Re tor Trip (from SI) , 5.75/

c. Xedwater Isolation S 15.0/

d./ Containment Isolation -Phase "A" S20.75 12)/30.715g" (

l Containment Ventilation'Isolation. Not Applicable t . Auxiliary Feedwater Pumps S5 61.75l

> . Service Water System S 15.75121/50.75"13 VteamLine

h. solaion 10.75/

SALEM - UNIT 1 3/4 3-28 Amendment No. J pA&VE LeFr 3L64K LkSTEMTLOWALLI

TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

6. Steam Flow in Two Steam Lines-High Coincident with Steam Line Pressure-Low

a. Safety Injection (ECCS)

12.0'

b. Reactor Trip (from SI)

C. Feedwater Isolation 1.0

d. Containment Isolation-Phase "A"

e. Containment Ventilation Isolation Not Applicable

f. Auxiliary Feedwater Pumps

60

g. Service Water System

14.0"/48.0'3'

h. Steam Line Isolation < 8.0 I

7.

a.

33.0

b. "B" Not Applicable

c.

7.0 I

8.

Trip 5 2.5

10.0

a. Motor-Driven Auxiliary Feedwater Pumps(4) S 60.0

b. Turbine-Driven Auxiliary Feedwater Pumps(5)

60.0 SALEM - UNIT 1 3/4 3-29 Amendment No.

PAt LEF VTLA,A I TEtVI - hLL Y

/ ~ENGINEERED SAFETY FEATURES RESPONSE TIMES/

IIITNG SIGNAL AND FUNCTION RESPONSE TIME INSE S

10. Undervoltage RCP Bus

a. Turbine-Driven Auxiliary Feedwater S 6 Pumps

11. Containment Radioactivity - High

a. Purge and Pressure Vacuum Relief 5.0

12. Trip of Feedwater Pumps

a. Auxiliary Feedwater Pumps Not Applicable

13. Undervoltage. Vital Bus

a. Loss of Voltage / *4.0

14. Station Blackout

a. Motor Driven Auxiliary / 60.0 Feedwater Pumps SALEM - UNIT 1 3/4 3-30 Amendment No. .4-"

PA&S LE FT 3L^K lr C -rL t AkL LY

TABlLE 3.3-5 (Continued)

TABLE NOTATION (1) Diesel generator starting and sequence loading delays included. Response time limit includes opening of valves to establish SI path an attainment of discharge pressure for centrifugal charging pumps, SI an RHR pumps.

(2) Diesel generator starting and sequence loading delays included.

Offsite power available. Response time limit includes ening of valves to establish SI path and attainment of discharge pressu for centrifugal charging pumps.

(3) Diesel generator starting and sequence loading lays included. Response time limit includes opening of valves to esta ish SI path and attainment of discharge pressure for centrifugal charg g pumps.

(4) On 2/3 in any steam generator.

(5) On 2/3 in 2/4 steam generators.

(6) The response time is the time t isolation circuitry input reaches the isolation setpoint to the time he Isolation Valves are fully shut.

(7) The response time includes he time to automatically align the service water flow to the CFCUs lowing an accident coincident with a loss of offsite power, and also ncludes the time delays associated with isolation of the Turbine Genera r Area service water header.

SALEM - UNIT 1 3/4 3-31 Amendment No.

CLESFT XLPt4 0I-mNlb<ItALLY,

EVALUATION OF REVISIONS TO THE TECHNICAL SPECIFICATIONS RELOCATION OF RESPONSE TIME TESTING TIME LIMITS TO THE UPDATED FINAL SAFETY ANALYSIS REPORT Attachment 2 Salem Unit 2 Affected Pages 1-3 Definition 1.12 1-6 Definition 1.26 3/4 3-1 Reactor Trip System Instrumentation 314 3-9 Table 3.3-2 3/4 3-10 Table 3.3-2 3/4 3-14 Engineered Safety Feature Actuation System Instrumentation 3/4 3-28 Table 3.3-5 3/4 3-29 Table 3.3-5 3/4 3-30 Table 3.3-5 3/4 3-31 Table 3.3-5 3/4 3-32 Table 3.3-5 Notations

DEFINITIONS thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844 "Calculation of Distance Factors for Power and Test Reactor Sites."

E - AVERAGE DISINTEGRATION ENERGY 1.11 E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half-lives greater than 15 minutes, making up at least 95t of the total non-iodine activity in the coolant.

ENGINEERED SAFETY FEATURE RESPONSE TIME 1.12 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.

D lIn lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed-and approved by the NRC.

FREQUENCY NOTATION 1.13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the-intervals defined in Table 1.2.

FULLY WITHDRAWN 1.13a FULLY WITHDRAWN shall be the condition where control and/or shutdown banks are at a position which is within the interval of 222 to 228 steps withdrawn, inclusive. FULLY WITHDRAWN will be established by the current reload analysis.

GASEOUS RADWASTE TREATMENT SYSTEM 1.14 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

IDENTIFIED LEAKAGE 1.15 IDENTIFIED LEAKAGE shall be:

a. Leakage (except Reactor Coolant Pump Seal Water Injection) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or SALEM - UNIT 2 1-3 Amendment to No.

DEFINITIONS I ; .n.. ..

REACTOR TRIP SYSTEM RESPONSE TIME 1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until loss of stationary gripper coil voltage.

A-pp In lieu of measurement, response time may be verified for selected

'1___T components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.

REPORTABLE EVENT 1.27 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10CFR Part 50.

SHUTDOWN MARGIN 1.28 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all full length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be FULLY WITHDRAWN.

SITE BOUNDARY 1.29 The SITE BOUNDARY shall be that line beyond which the land is not owned, leased, or otherwise controlled by the licensee, as shown in Figure 5.1-3, and which defines the exclusion area as shown in Figure 5.1-1.

SOLIDIFICATION 1.30 Not Used SOURCE CHECK 1.31 SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.

STAGGERED TEST BASIS 1.32 A STAGGERED TEST BASIS shall consist of:

a. A test schedule for (ni systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into (n) equal subintervals.

SALEM - UNIT 2 1-6 Amendment No.

3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1.1 As a minimum, the reactor trip system instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE,4th fRGCFfr ,. D APPLICABILITY: As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1.

SURVEILLANCE REQUIREMENTS 4.3.1.1.1 Each reactor trip system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies shown in Table 4.3-1.

4.3.1.1.2 The logic for the interlocks shall be demonstrated OPERABLE prior to each reactor startup unless performed during the preceding 92 days.

The total interlock function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by interlock operation.

4.3.1.1.3 The REACTOR TRIP.SYSTEM RESPONSE TIME of each reactor trip function shall be verified to be within its limit at least once per 18 months. Eah verification shall include at least one logic train such that both logic trains are verified at least once per 36 months and one channel per.function such that all channels are verified at least once every N times 18 months.

where N is the total number of redundant channels in.a specific reactor trip function as shown in the "Total No. of Channels" column of Table 3.3-1.

a viri Zi ao .e A esDv SLM-UNIT 2 3 e4 31Am ndment Nov.

SALEM - UNIT 2 3/4 3-1 Amendment NNo.

TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES FUNCTIONAL UNIT RESPONSE TIME

1. Manual Reactor Trip NOT APPLICABLE

2. Power Range, Neutron Flux ; 0.5 seconds*

3. Power Range, Neutron Flux, NOT APPLICABLE High Positive Rate

4. Power Range, Neutron Flux, < 0.5 se High Negative Rate

5. Intermediate Range, Neutron Flux APPLICABLE

6. Source Range, Neutron Flux NOT APPLICABLE

7. Overtemperature AT / 5.75 seconds*

8. Overpower AT NOT APPLICABLE

9. Pressurizer Pressure--Low / 2.0 seconds

10. Pressurizer Pressur High

2.0 seconds

11. Pressurizer r Level--High NOT APPLICABLE XNeutron tectors are exempt from response time testing. Response time o he neutron flux signal portion of the channel shall be measured fro etector output or input of first electronic component in channel.

CASE LEFT 'L AtA4< i tET X1 I4DJ AL Y SALEM - UNIT 2 3/4 3-9 Amendment No.1-a

TABLE 3.3-2 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES FUNCTIONAL UNIT RESPONSE TIME

12. Loss of Flow - Single Loop

1.0 seconds (Above P-8)

13. Loss of Flow - Two Loops

1.0 seconds (Above P-7 and below P-8)

14. Steam Generator Water Level--

2.0 seconds Low-Low

15. Deleted

16. Undervoltage-Reactor Coolant Pumps

17. Underfrequency-Reactor Coolant Pumps 0.6 seconds

18. Turbine Trip A. Low Fluid Oil Pressure NOT APPLICABLE B. Turbine Stop Valve NOT APPLICABLE

19. Safety Injection Input f ESF NOT APPLICABLE

20. Reactor Coolant P Breaker Position Trip NOT APPLICABLE

21. Reactor Tri akers NOT APPLICABLE

22. Automati rip Logic NOT APPLICABLE L LA^ 1WT&e. ETr4O IALLY SALEM - UNIT 2 3/4 3-10 Amendment No.

INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM'INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2.1 The Engineered Safety Feature Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4. -A '-4' piI sTMare a- g~io i- Tc+/-1: . E r APPLICABILITY: As shown in Table 3.3-3.

ACTION:

a. With an ESFAS instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3-4, declare the channel inoperable and apply the applicable ACTION requirement of Table 3.3-3 until the channel is restored to OPERABLE status with the trip setpoint adjusted consistent with the Trip Setpoint value.

b. With an ESFAS instrumentation channel inoperable, take the ACTION shown in Table 3.3-3.

SURVEILLANCE REQUIREMENTS 4.3.2.1.1 Each ESFAS instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES and at frequencies shown in Table 4.3-2.

4.3.2.1.2 The logic for the interlocks shall be demonstrated OPERABLE during the automatic actuation logic test. The total interlock function shall be demonstrated OPERABLE at least once per' 18 months during CHANNEL CALIBRATION testing of each channel affected by interlock operation.

4.3.2.1.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be verified to be within the limit at least once per 18 months. Each verification shall include at least one logic train such that both logic trains are verified at least once per 36 months and one channel per function such that all channels are verified at least once per N times 18 months where N is the total number of redundant channels in a specific ESFAS function as shown in the "Total No. of Channels" Column of Table 3.3-3. The provisions of Specification 4.0.4 are not applicable to MSIV closure time testing.

The provisions of Specification 4.0.4 are not applicable to the turbine driven auxiliary feedwater pump provided the surveillance is performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the secondary steam generator pressure is greater than 680 psig.

SALEM - UNIT 2 3/4 3-14 Amendment No. ;-/

Reactor TrpBSIE

{ ~~ENGINEERED SAFETY FEATURES RESPONSE TIMES /)

/ INITIATING SIGNAL AND FUNCTION RESPONSE TIME SECONDS

/ 1. Manual/)

(a. Safety Injection (ECCS) Not plcable(

\Feedwater Isolation Ns Applicable Reactor Trip (SI) / t Applicable)

Containment Isolation-Phase "A" Not Applicable Containment Ventilation Isolation Not Applicable Auxiliary Feedwater Pumps Not Applicable Service Water System Not Applicable Containment Fan Cooler Not Applicable

b. Containment Spray Not Applicable Containment Isolation-Phase "B" Not Applicable Containment Ventilation Isol ion Not Applicable

c. Containment Isolation-Phas "A" Not Applicable Containment Ventilation solation Not Applicable

d. Steam Line Isolation Not Applicable

2. Containment Pressure-X h

a. Safety Injecti (ECCS)

27.0(1)

b. Reactor Trip from SI) < 2.0

c. Feedwater solation S 10.0

d. Contai ent Isolation-Phase "A" < 17.0121/27.c031

e. Cont nmerit VentilationhIsolation - Not Applicable"

f. A .iliary Feedwater Pumps < 60

g. Service Water System

13.0(2 /45.0(3)

Containment Fan Coolers S 60.0171 SE UN L2/4F3T -2L84 AmmEen4 N.L SALEM - UNIT 2 3/4 3-28 Amendment No. ;/;,

/ ~ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN ECONDS

3. Pressurizer Pressure-Low

a. Safety Injection (ECCS) S 27. /12.0(2 /

b. Reactor Trip (from SI) S 2

c. Feedwater Isolation 0.0

d. Containment Isolation-Phase "A" 18.0 21

e. Containment Ventilation Isolation Not Applicable

f. Auxiliary Feedwater Pumps / 60

/g. Service Water System / 49.0(1)/13. 02)1

4. Differential Pressure Between Steam Lin -Hi h

\a. Safety Injection (ECCS) / S 12.0(2)/22.0(3) /

b. Reactor Trip (from SI) / 2.0

c. Feedwater Isolation / 10.0

d. Containment Isolation Ph "A" "e

17.0(2)/27.0(3)

e. Containment Ventilatio Isolation Not Applicable

f. Auxiliary Feedwater umps

60 ag. Service Water S / m 13.0(2)1/48.0(3)

5. Steam Flow in two team Lines High-Coincident

/ ~with TV, --Low-vgw

a. .SafetyIn'ection .(ECCS)

15.75(2)125.75(3)

b. Reactor rip (from SI)

5.75

c. Feedw ter Isolation

15.0

d. Co ainment Isolation-Phase "A"

20.75(2)/30.75(3)

e. ntainment Ventilation Isolation Not Applicable

f. Auxiliary Feedwater Pumps

61.75 Service Water System

15.75(2) /50.75 (3)

h. Steam Line Isolation

10.75 SALEM - UNIT 2 3/4 3-29 Amendment No.

/ - TABLE 3.3-5 (Continued)v ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SEC DS

6. Steam Flow in Two Steam Lines-High Coincident with Steam Line Pressure-Low

a. Safety Injection (ECCS) < 12.0(2) 0

b. Reactor Trip (from SI) S 2.0

c. Feedwater Isolation S

d. Containment Isolation-Phase "A" 7.0X2) /27.0")

e. Containment Ventilation Isolation Not Applicable

f. Auxiliary Feedwater Pumps / 60

g. Service Water System / 14.O 2) /48.0X3)

h. Steam Line Isolation / 8.0

7. Containment Pressure--Hi h-Hi

/a. Containment Spray / 33.0

b. Containment Isolation hase "B" Not Applicable

c. Steam Line Isolat

7.0

8. Steam Generator Wat Level--Hi h-High

a. Turbine Trip/ 2.5

b. Feedwater Xsolation 10.0

9. Steam Gene tor Water Level --Low-Low

a. Mot -Driven Auxiliary Feedwater S 60.0

- PD ps(4) - -- -

b. Turbine-Driven Auxiliary Feedwater < 60.0 Pumps (5)

\LLY ErE.- Lt V T "5 hAK 147- r-tA OTt 0 SALEM - UNIT 2 3/4 3-30 Amendment No.e

~~TABLE 3.3-5 (Continued)><

l ~ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

10. Undervoltage RCP Bus

a. Turbine-Driven Auxiliary Feedwater

60.0 Pumps

11. Containment Radioactivity - High

a. Purge and Pressure Vacuum Relief / 0(6/

12. Trip of Feedwater Pumps

a. Auxiliary Feedwater Pumps Not Applicable

13. Undervoltage, Vital Bus

a. Loss of Voltage / 4.0

14. Station Blackout

a. Motor'Driven il lary Feed Pumps, S 60.0/

15. Semiautomati ransfer to Recirculation 4

\ a. EC Xa aves 21SJ44, 22SJ44, 21RH4, H

R4, 21CC16, 22CC16, 21SJ113, 22SJ113 Not Applicable 9LEV L 1- eLht4K itETetTtc)PALLY SALEM - UNIT 2 3/4 3-31 Amendment No.

1

TABLE 3.3-5 (Continued)

TABLE NOTATION (1) Diesel generator starting and sequence loading delays inclu Response time limit includes opening of valves to establi SI path and attainment of discharge pressure for centrifugal chargi pumps, SI and RHR pumps.

(2) Diesel generator starting and sequence loading elays not included.

Offsite power available. Response time limi ncludes opening of valves to establish SI path and attainment of di narge pressure for centrifugal charging pumps.

(3) Diesel generator starting and se ence loading delays included. Response time limit includes opening of alves to establish SI path and attainment of discharge pressure for ce rifugal charging pumps.

(4) On 2/3 in any steam gen ator.

(5) On 2/3 in 2/4 steam nerators.

(6) The response t X is the time the isolation circuitry input reaches the isolation se oint to the time the Isolation Valves are fully shut.

(7) The res se time includes the time to automatically align the service water ow to the CFCUs following an accident coincident with a loss of off ce power, and also includes the time delays associated with i ation of the Turbine Generator Area service water header.

'AEL T -5L h 4 I tTE Wr I 0U PtL L SALEM - UNIT 2 3/4 3-32 AmendmenttNo..

EVALUATION OF REVISIONS TO THE TECHNICAL SPECIFICATIONS RELOCATION OF RESPONSE TIME TESTING TIME LIMITS TO THE UPDATED FINAL SAFETY ANALYSIS REPORT SALEM NUCLEAR GENERATING STATION UNITS 1 AND 2 FACILITY OPERATING LICENSES DPR-70 AND DPR-75 DOCKET NOs. 50-272 AND 50-311 ATTACHMENT 3 UPDATED FINAL SAFETY ANALYSIS REPORT MARKED-UP CHANGES

TABLE 7.2-4 SALEM UNIT-1 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES FUNCTIONAL UNIT RESPONSE TIME

1. Manual Reactor Trip NOT APPLICABLE

2. Power Range, Neutron Flux

0.5 seconds*

3. Power Range, Neutron Flux, NOT APPLICABLE High Positive Rate

4. Power Range, Neutron Flux,

0.5 seconds*

High Negative Rate

5. Intermediate Range, Neutron Flux NOT APPLICABLE

6. Source Range, Neutron Flux NOT APPLICABLE

7. Overtemperature AT

5.75 seconds*

8. Overpower AT NOT APPLICABLE

9. Pressurizer Pressure--Low

2.0 seconds

10. Pressurizer Pressure--High : 2.0 seconds

11. Pressurizer Water Level--High NOT APPLICABLE

Neutron detectors are exempt from response time testing. Response time of the neutron flux signal portion of the channel shall be measured from detector output or input of first electronic component in channel.

1

TABLE 7.2-4(Continued)

SALEM UNIT-1 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES FUNCTIONAL UNIT RESPONSE TIME

12. Loss of Flow - Single Loop (Above P-8)

1.0 seconds

13. Loss of Flow - Two Loops (Above P-7 and below P-8)

1.0 seconds

14. Steam Generator Water Level-- Low-Low

2.0 seconds

15. Deleted

16. Undervoltage-Reactor Coolant Pumps

1.2 seconds

17. Underfrequency-Reactor Coolant Pumps

0.6 seconds

18. Turbine Trip A. Low Fluid Oil Pressure NOT APPLICABLE B. Turbine Stop Valve NOT APPLICABLE

19. Safety Injection Input from ESF NOT APPLICABLE

20. Reactor Coolant Pump Breaker Position Trip NOT APPLICABLE

21. Reactor Trip Breakers NOT APPLICABLE

22. Automatic Trip Logic NOT APPLICABLE 2

TABLE 7.2.5 SALEM UNIT-1 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME

1. Manual

a. Safety Injection (ECCS) Not Applicable Feedwater Isolation Not Applicable Reactor Trip (SI) Not Applicable Containment Isolation-Phase "A" Not Applicable Containment Ventilation Isolation Not Applicable Auxiliary Feedwater Pumps Not Applicable Service Water System Not Applicable Containment Fan Cooler Not Applicable

b. Containment Spray Not Applicable Containment Isolation-Phase "B" Not Applicable Containment Ventilation Isolation Not Applicable

c. Containment Isolation-Phase "A" Not Applicable Containment Ventilation Isolation Not Applicable

d. Steam Line Isolation Not Applicable 1

TABLE 7.2-5 (Continued)

SALEM UNIT-1 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

2. Containment Pressure-High

a. Safety Injection (ECCS)

27.0(1)

b. Reactor Trip (from SI)

2.0 C. Feedwater Isolation

10.0

d. Containment Isolation-Phase "A"

17.0(2)/27.0(3)

e. Containment Ventilation Isolation Not Applicable

f. Auxiliary Feedwater Pumps

60

g. Service Water System

13.0(2)/45.0(3)

h. Containment Fan Coolers

60.0 (7)

3. Pressurizer Pressure-Low

a. Safety Injection (ECCS)

27.0(1)/12 .(2)

b. Reactor Trip (from SI)

2.0

c. Feedwater Isolation

10.0

d. Containment Isolation - Phase "A" < 18.0(21

e. Containment Ventilation Isolation Not Applicable

f. Auxiliary Feedwater Pumps

60

g. Service Water System

49.01)/13.0(2}

2

TABLE 7.2-5 (Continued)

SALEM UNIT-1 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

4. Differential Pressure Between Steam Lines-High

a. Safety Injection (ECCS)

12. 02)/22.0(3)

b. Reactor Trip (from SI) 5 2.0 C. Feedwater Isolation

10.0

d. Containment Isolation - Phase "A"

17.0(2) /27. 0(3)

e. Containment Ventilation Isolation Not Applicable

f. Auxiliary Feedwater Pumps

60

g. Service Water System

13.0(2) /48. 0(3)

5. Steam Flow in Two Steam Lines - High Coincident with Tavg -- Low-Low

a. Safety Injection (ECCS) < 15.75 (2)/25.75 (3)

b. Reactor Trip (from SI)

5.75

c. Feedwater Isolation

15.0

d. Containment Isolation - Phase "A"

20.75(2)/30.75(3)

e. Containment Ventilation Isol. NOT APPLICABLE

f. Auxiliary Feedwater Pumps

61.75

g. Service Water System

15.75 (2) /50.75 (3)

h. Steam Line Isolation

10.75 3

TABLE 7.2-5 (Continued)

SALEM UNIT-1 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

6. Steam Flow in Two Steam Lines-High Coincident with Steam Line Pressure-Low

a. Safety Injection (ECCS)

12.0(2) /22. 0(3)

b. Reactor Trip (from SI)

2.0

c. Feedwater Isolation

10.0

d. Containment Isolation-Phase "A"

17.0(21/27. 0(3)

e. Containment Ventilation Isolation Not Applicable

f. Auxiliary Feedwater Pumps

60

g. Service Water System

14.0(2/48. 0(3)

h. Steam Line Isolation < 8.0

7. Containment Pressure--High-High

a. Containment Spray

33.0

b. Containment Isolation-Phase "B" Not Applicable

c. Steam Line Isolation

7.0

8. Steam Generator Water Level--High High

a. Turbine Trip

2.5

b. Feedwater Isolation

10.0 4

TABLE 7.2-5 (Continued)

SALEM UNIT-1 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

9. Steam Generator Water Level--Low-Low

a. Motor-Driven Auxiliary Feedwater Pumps(4) 5 60.0

b. Turbine-Driven Auxiliary Feedwater Pumps(5)

60.0

10. Undervoltage RCP Bus

a. Turbine-Driven Auxiliary Feedwater Pumps

60.0

11. Containment Radioactivity - High

a. Purge and Pressure Vacuum Relief

5.0 (6)

12. Trip of Feedwater Pumps

a. Auxiliary Feedwater Pumps Not Applicable

13. Undervoltage, Vital Bus

a. Loss of Voltage < 4.0

14. Station Blaeckut Loss of Offsite Power I

a. Motor Driven Auxiliary Feedwater Pumps

60.0 5

TABLE 7.2-5 (Continued)

SALEM UNIT-1 TABLE NOTATIONS (1) Diesel generator starting and sequence loading delays included. Response time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps, SI and RHR pumps.

(2) Diesel generator starting and sequence loading delays not included. Offsite power available.

Response time limit includes opening of valves-to establish SI path and attainment of discharge pressure for centrifugal charging pumps.

(3) Diesel generator starting and sequence loading delays included. Response time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps.

(4) On 2/3 in any steam generator.

(5) On 2/3 in 2/4 steam generators.

(6) The response time is the time the isolation circuitry input reaches the isolation setpoint to the time the Isolation Valves are fully shut.

(7) The response time includes the time to automatically align the service water flow to the CFCUs following an accident coincident with a loss of offsite power, and also includes the time delays associated with isolation of the Turbine Generator Area service water header.

6

TABLE 7.2.4A SALEM UNIT-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES FUNCTIONAL UNIT RESPONSE TIME

1. Manual Reactor Trip NOT APPLICABLE

2. Power Range, Neutron Flux < 0.5 seconds*

3. Power Range, Neutron Flux, High Positive Rate NOT APPLICABLE

4. Power Range, Neutron Flux, High Negative Rate < 0.5 seconds*

5. Intermediate Range, Neutron Flux NOT APPLICABLE

6. Source Range, Neutron Flux NOT APPLICABLE

7. Overtemperature AT < 5.75 seconds*

8. Overpower AT NOT APPLICABLE

9. Pressurizer Pressure--Low < 2.0 seconds

10. Pressurizer Pressure--High < 2.0 seconds

11. Pressurizer Water Level--High NOT APPLICABLE

Neutron detectors are exempt from response time testing. Response time of the neutron flux signal portion of the channel shall be measured from detector output or input of first electronic component in channel.

1

TABLE 7.2-4A (Continued)

SALEM UNIT-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES FUNCTIONAL UNIT RESPONSE TIME

12. Loss of Flow - Single Loop (Above P-8)

1.0 seconds

13. Loss of Flow - Two Loops (Above P-7 and below P-8) < 1.0 seconds

14. Steam Generator Water Level--Low-Low < 2.0 seconds

15. Deleted

16. Undervoltage-Reactor Coolant Pumps

1.2 seconds

17. Underfrequency-Reactor Coolant Pumps

0.6 seconds

18. Turbine Trip

a. Low Fluid Oil Pressure NOT APPLICABLE

b. Turbine Stop Valve NOT APPLICABLE

19. Safety Injection Input from ESF NOT APPLICABLE

20. Reactor Coolant Pump Breaker Position Trip NOT APPLICABLE

21. Reactor Trip Breakers NOT APPLICABLE

22. Automatic Trip Logic NOT APPLICABLE 2

TABLE 7.2-5A SALEM UNIT-2 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

1. Manual

a. Safety Injection (ECCS) Not Applicable Feedwater Isolation Not Applicable Reactor Trip (SI) Not Applicable Containment Isolation-Phase "A" Not Applicable Containment Ventilation Isolation Not Applicable Auxiliary Feedwater Pumps Not Applicable Service Water System Not Applicable Containment Fan Cooler Not Applicable

b. Containment Spray Not Applicable Containment Isolation-Phase "B" Not Applicable Containment Ventilation Isolation Not Applicable

c. Containment Isolation-Phase "A" Not Applicable Containment Ventilation Isolation Not Applicable

d. Steam Line Isolation Not Applicable 1

TABLE 7.2.5A (Continued)

SALEM UNIT-2 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

2. Containment Pressure-High

a. Safety Injection (ECCS)

27.0w

b. Reactor Trip (from SI) < 2.0 C. Feedwater Isolation < 10.0

d. Containment Isolation-Phase "A" < 17.012) /27. 0(3)

e. Containment Ventilation Isolation Not Applicable

f. Auxiliary Feedwater Pumps < 60

g. Service Water System < 13.0(2)/45.0(3)

h. Containment Fan Coolers < 60.0171 2

TABLE 7.2-5A (Continued)

SALEM UNIT-2 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

3. Pressurizer Pressure-Low

a. Safety Injection (ECCS)

27.0(')/12.0(2)

b. Reactor Trip (from SI) < 2.0

c. Feedwater Isolation < 10.0

d. Containment Isolation-Phase "A"

18.0(2)

e. Containment Ventilation Isolation Not Applicable

f. Auxiliary Feedwater Pumps < 60

g. Service Water System

49.0(1)/13 .0(2)

4. Differential Pressure Between Steam Lines-High

a. Safety Injection (ECCS) < 12 .02)/22. 03)

b. Reactor Trip (from SI)

2.0

c. Feedwater Isolation < 10.0 3

TABLE 7.2-5A (Continued)

SALEM UNIT-2 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

d. Containment Isolation Phase "A" < 17.0(2)/27. 0(3

e. Containment Ventilation Isolation Not Applicable

f. Auxiliary Feedwater Pumps

60

g. Service Water System < 13.0(21 /48.0(31

5. Steam Flow in two Steam Lines High-Coincident with Tav --Low-Low

a. Safety Injection (ECCS) < 15.75(2)/25.75(3)

b. Reactor Trip (from SI) < 5.75

c. Feedwater Isolation

15.0

d. Containment Isolation-Phase "A" < 20.75(2)/30.75(3)

e. Containment Ventilation Isolation Not Applicable

f. Auxiliary Feedwater Pumps

61.75

g. Service Water System < 15.75(2)/50.75(3)

h. Steam Line Isolation

10.75 4

9 TABLE 7.2-5A (Continued)

SALEM UNIT-2 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

6. Steam Flow in Two Steam Lines-High Coincident with Steam Line Pressure-Low

a. Safety Injection (ECCS) < 12 .0(2) /22 .0(3)

b. Reactor Trip (from SI) < 2.0

c. Feedwater Isolation

10.0

d. Containment Isolation-Phase "A" < 17.0 12) /27.0(3)

e. Containment Ventilation Isolation Not Applicable

f. Auxiliary Feedwater Pumps

60

g. Service Water System < 14. 0(2/48. 0(3)

h. Steam Line Isolation

8.0

7. Containment Pressure--High-High

a. Containment Spray

33.0

b. Containment Isolation-Phase "B" Not Applicable

c. Steam Line Isolation

7.0 5

0 TABLE 7.2-5A (Continued)

SALEM UNIT-2 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

8. Steam Generator Water Level-High-High

a. Turbine Trip

2.5

b. Feedwater Isolation < 10.0

9. Steam Generator Water Level --Low-Low

a. Motor-Driven Auxiliary Feedwater Pumps(4) < 60.0

b. Turbine-Driven Auxiliary Feedwater Pumps(5) < 60.0 10.Undervoltage RCP Bus

a. Turbine-Driven Auxiliary Feedwater Pumps < 60.0 11.Containment Radioactivity - High

a. Purge and Pressure Vacuum Relief

5. 061 12.Trip of Feedwater Pumps

a. Auxiliary Feedwater Pumps Not Applicable 13.Undervoltage, Vital Bus

a. Loss of Voltage

4.0 14.Statien Blza~cout Loss of Offsite Power I

a. Motor Driven Auxiliary Feed Pumps < 60.0 15.Semi-automatic Transfer to Recirculation

a. ECCS valves 21SJ44, 22SJ44, 21RH4, 22RH4, 21CC16, 22CC16, 21SJ113, 22SJ113 Not Applicable 6

TABLE 7.2-5A (Continued)

SALEM UNIT-2 TABLE NOTATIONS (1)Diesel generator starting and sequence loading delays included. Response time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps, SI and RHR pumps.

(2)Diesel generator starting and sequence loading delays not included. Offsite power available.

Response time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps.

(3)Diesel generator starting and sequence loading delays included. Response time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps.

(4)On 2/3 in any steam generator.

(5)On 2/3 in 2/4 steam generators.

(6)The response time is the time the isolation circuitry input reaches the isolation setpoint to the time the Isolation Valves are fully shut.

(7)The response time includes the time to automatically align the service water flow to the CFCUs following an accident coincident with a loss of offsite power, and also includes the time delays associated with isolation of the Turbine Generator Area service water header.

7

The method described provides capability for checking from the process signal to the logic cabinets and from there to the individual field equipment including all field cabling actually used in the circuitry. For those devices whose operation could have an effect on plant stability, the procedure provides for checking from the process signal to the logic rack and continuity determination for output cables and field devices; however, the actuated equipment will be manually initiated as plant conditions permit.

The SEC units have the following test capability during power operation:

1. Check the operational capability of each bus undervoltage sensor and its input to the logic.

2. Check the operational capability of the LOCA signal, "SO, from the SSPS logics.

3. Check that the logic combinations of input signals result in proper operation of the various functions, including automatic load sequencing, without actuation of any motors and a verification of the timed loading sequence.

4. Check the output relay capability to actuate the driven equipment.

The SEC units can also be checked for complete system operability from sensor to actuated equipment during plant shutdowns.

Reactor Trip System and ESF actuation system response time tests are required by and will be performed in accordance with the Technical Specifications. The Technical Specifications Tables containing the response time limits were relocated to UFSAR Tables 7.2.4 and 7.2.5 for Unit 1 and 7.2.4A and 7.2.5A for Unit 2. The relocation of these tables to this document was approved by the NRC in Amendments for Unit 1 and - for Unit 2.

7.3-26 SGS-UFSAR Revision 6 February 15, 1987

LIST OF TABLES Table Title 7.2-1 List of Reactor Trips, Engineered Safety Features, Containment and Steam Line Isolation and Auxiliary Feedwater 7.2-2 Interlock Circuits 7.2-3 Legend of Analog Symbols 7.2-4 Salem Unit 1 - Reactor Trip System Instrumentation Response Times.

7.2.4A Salem Unit 2 - Reactor Trip System Instrumentation Response Times.

7.2-5 Salem Unit 1 - Engineered Safety Features Response Times.

7.2-5A Salem Unit 2 - Engineered Safety Features Response Times 7.3-1 Process Instrumentation for RPS and ESF Actuation 7.3-2 Post-Accident Equipment (Inside Containment) Operational and Testing Requirements 7.3-3 Postulated Submerged Electrical Components in the Containment Following a LOCA 7.3-4 Safety Evaluation - Electrical Components and Circuits That are Affected by the Flooding of Components Within the Containment During Post-LOCA Conditions 7.3-5 Safety Evaluation - Submerged Electrical Components in the Containment During Post-LOCA Conditions 7.3-6 Safety Evaluation - Electrical Components and Circuits That are Affected by the Flooding of Components Within The Containment During Post-LOCA Conditions 7-vi SGS-UFSAR Revision