LR-N05-0392, Request for Change to Technical Specifications Accident Monitoring Instrumentation

From kanterella
Jump to navigation Jump to search
Request for Change to Technical Specifications Accident Monitoring Instrumentation
ML052560221
Person / Time
Site: Salem  PSEG icon.png
Issue date: 08/31/2005
From: Joyce T
Public Service Enterprise Group
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
LCR S05-06, LR-N05-0392
Download: ML052560221 (37)
v • d • e


Text

Thomas P.Joyce PSEG Nuclear LLC Site Vice President - Salem P.O. Box 236, Hancocks Bridge, NJ 08038-1236 tel: 856.339.2086 fax: 856.339.2956 LCR S05-06 0 SuII AUG 3 1 2005 ANiclear- LLC United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 REQUEST FOR CHANGE TO TECHNICAL SPECIFICATIONS ACCIDENT MONITORING INSTRUMENTATION SALEM GENERATING STATION - UNIT I AND UNIT 2 DOCKET NOS. 50-272 AND 50-311 FACILITY OPERATING LICENSE NOS. DPR-70 AND DPR-75 In accordance with the provisions of 10 CFR 50.90, PSEG Nuclear, LLC (PSEG) hereby transmits a request for amendment of the Technical Specifications (TS) for Salem Generating Station Unit 1 and Unit 2. In accordance with 10 CFR 50.91 (b)(1), a copy of the transmittal has been sent to the State of New Jersey.

The proposed amendment moves the containment high range accident monitors (R44) from the radiation monitoring instrumentation technical specification (3.3.3.1) to the accident monitoring technical specification (3.3.3.7). Since TS 3.3.3.7 does not require source checks to demonstrate operability, this proposed change will remove the unnecessary requirement to expose the detectors to a source of radioactivity to comply with TS definition 1.31, SOURCE CHECKS.

The definition of SOURCE CHECKS will be modified to allow different methods to comply with a SOURCE CHECK requirement. This change will result in significant dose savings and remove the industrial safety hazard of containment entries at power by eliminating the requirement to source check the containment high range accident monitors.

This proposed change also corrects a typographical error in Salem Unit 1 Surveillance Requirement 4.2.2 contained in a previous amendment.

Attachment 1 provides a description of the proposed changes. Attachment 2 provides the existing TS pages marked up to show the proposed changes.

Attachment 3 provides the appropriate TS incorporating the proposed change.

PSEG requests a 60-day implementation period after amendment approval.

Should you have any questions regarding this request, please contact Mr. Justin Wearne at (856) 339-5081.

Orol

Document Control Desk LR-N05-0392 AUG 3 1 2005 I declare under penalty of perjury that the foregoing is true and correct.

Executed on ____ ____

(bate)

Sincerely, Thomas P. Jo/ce Site Vice President Salem Generating Station Attachments (3)

C: Mr. S. Collins, Administrator - Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. S. Bailey, Licensing Project Manager - Salem U. S. Nuclear Regulatory Commission Mail Stop 08B1 Washington, DC 20555 USNRC Senior Resident Inspector - Salem (X24)

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

DOCUMENT CONTROL DESK Attachment I LR-N05-0392 REQUEST FOR CHANGE TO TECHNICAL SPECIFICATIONS ACCIDENT MONITORING INSTRUMENTATION Table of Contents

1. Description ........................................ 2
2. Proposed Change ........................................ 2
3. Background ........................................ 2
4. Technical Analysis ........................................ 4
5. Regulatory Safety Analysis ....................................... 5 5.1 No Significant Hazards Consideration ...................................... 5 5.2 Applicable Regulatory Requirements/Criteria ............................. 6
6. Environmental Considerations ....................................... 7
7. References ........................................ 7 1

DOCUMENT CONTROL DESK Attachment I LR-N05-0392 CHANGES TO TECHNICAL SPECIFICATIONS

1. DESCRIPTION The purpose of this amendment is to revise Limiting Condition for Operation 3.3.3.1 and 3.3.3.7 by moving the requirement for a high range accident monitor from the radiation monitoring specification (3.3.3.1) to the accident monitoring specification (3.3.3.7).

The definition of SOURCE CHECK in the Definitions Section of Technical Specifications (Section 1) is revised to remove the requirement for exposing the detector to a source of radioactivity.

2. PROPOSED CHANGE Salem Technical Specifications (TS) require two high range accident radiation monitors to monitor conditions post accident inside the containment. The requested change moves the LCO from the radiation monitor specification to the accident monitoring specification. There will be no change in the required number of channels or actions for inoperability. There are two changes as the result of this administrative move in TS. The first change is that the containment high range accident radiation monitor will be required in modes 1-3 as opposed to the currently specified modes 1-4. The other change is the elimination of a monthly surveillance requirement of a SOURCE CHECK in accordance with TS definitions.

The definition of SOURCE CHECK in section 1.31 of Technical Specifications is revised to allow for determining proper operation of various detectors that do not require exposing the detector to a source of radioactivity.

A typographical error in Salem Unit 1 TS is also corrected. The error was inadvertently introduced in Salem Unit 1 Amendment 201, which was issued November 26, 1997. The error is an incorrect minus sign in the equation for Fxy located in surveillance requirement 4.2.2.2.c.2.

3. BACKGROUND It was recently realized that the current design and method for performing the source check on the containment high range accident monitors is not in compliance with the present definition of SOURCE CHECK. Upon discovery, corrective actions were taken that changed the testing methodology to be in compliance with the definition of SOURCE CHECKS. The revised testing methodology consists of containment entries to expose the containment high 2

DOCUMENT CONTROL DESK Attachment 1 LR-N05-0392 range accident radiation monitor to a source of radioactivity to comply with the SOURCE CHECK surveillance requirement. This action results in additional occupational dose and exposure to other industrial safety issues (e.g., heat stress, climbing, etc.) without any increased assurance of detector operability.

The proposed amendment will remove the requirement for the monthly SOURCE CHECK surveillance.

The goal of the accident monitoring Technical Specification is to ensure that sufficient information is available on selected plant parameters to monitor and assess these variables following an accident. This capability is consistent with the recommendations of Regulatory Guide 1.97, "Instrumentation for Light Water Cooled Nuclear Power Plants to Asses Plant Conditions During and Following an Accident." This proposed change is consistent with the Westinghouse Standard Technical Specifications (STS), NUREG-1431. The purpose of the radiation monitors contained in Technical Specification 3.3.3.1, per the TS basis is to ensure that:'

1. The radiation levels are continuously measured in the areas served by the individual channels, and
2. The alarm or automatic action setpoint is initiated when the radiation level trip setpoint is reached.

The containment environment will be still be monitored through redundant radiation monitoring channels. There are no changes to setpoints as a result of this change.

Regulatory Guide 1.97, Table 3 UPWR Instrumentation," requires two containment high range radiation monitors, located at widely separated locations capable of detecting gamma radiation within an energy range of 60 keV to 3 MeV between 1 R/hr and 10,000 R/hr. The currently installed containment high range accident monitors meet this requirement.

The proposed change eliminates the significant amount of occupational radiation exposure that is incurred by performing SOURCE CHECKS monthly by exposing local detectors inside the containment, as opposed to utilizing the current built-in equivalent electronic check source (ECS), which ionizes the gas in the detector by remote electrical means. Ionizing the detector gas generates the same instrument and detector response as exposing the detector to a source or radiation.

The Westinghouse STS, NUREG 1431, do not contain a definition for SOURCE CHECK. The STS requires a monthly CHANNEL CHECK and no source check on radiation detectors in the post accident monitoring specification. The proposed specification will maintain the present CHANNEL CHECK performed every shift.

3

DOCUMENT CONTROL DESK Attachment I LR-N05-0392

4. TECHNICAL ANALYSIS This proposal removes the Surveillance Requirement to perform a monthly SOURCE CHECK on the containment high range accident radiation detector, in accordance with the TS definition. The TS definition of SOURCE CHECK requires exposing the detector to a source of radioactivity and monitoring for proper detector response. The proposed change is to allow for the performance of an electronic source check (ECS), which consists of raising the voltage across the ion chamber. This ionizes the gas in the detector, which has an equivalent effect as exposing it to a source of radioactivity. The resulting instrument upscale validates that the ECS circuit has the same result as exposing the detector to a source of radioactivity.

The other change associated with this proposal is to change the mode applicability from modes 1-4 to modes 1-3. There is no adverse consequence to changing the mode applicability from modes 1-4 to modes 1-3 since the following are the instances that the containment high range accident radiation monitors are used:

The containment high range accident radiation monitor is used in the Fission Product Barrier Determination in the Salem Event Classification Guide (ECG) table 3.1, (applicable in modes 1-4). Multiple independent parameters, such as core exit thermocouples, containment pressure and reactor vessel water level remain available to determine the status of the.

fission product barriers.

These instruments are used in Salem's emergency operating procedures (EOPs) directly via the containment high range accident detectors and indirectly through the Sub-Cooling Margin Monitor (SCMM). The containment high range accident detectors provide an input into the SCMM to determine if adverse environmental conditions exist in the containment building. The Salem EOPs are applicable in modes 1-3 and the SCMM are required under TS 3.3.3.7 in modes 1-3.

The SCMM can be used in mode 4 to determine if safety injection (SI) flow reduction can proceed following a shutdown LOCA. A shutdown LOCA is defined as a LOCA in mode 3 or 4. The containment high range accident detectors provide an input to the SCMM to switch to an "adverse" mode of display when high radiation is detected in containment. Not having the automatic switch to adverse readings on high radiation levels has minimal consequence in mode 4 since mode 4 is a transitional mode, in that mode 4 is only entered as a result of transitioning between mode 3 and mode 5.

It is not current operating practice to maintain the plant in mode 4 for a significant period of time. However, if a shutdown LOCA were to occur 4

DOCUMENT CONTROL DESK Affachment I LR-N05-0392 while in mode 4, a procedure exists to calculate sub-cooling margin manually in the event that the SCMM is unavailable, and contingencies exist in this procedure to account for adverse conditions in the containment. The SCMM are required to be operable in modes 1-3 per the TS 3.3.3.7.

5. REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration
1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed change presents no change in the probability of a previously evaluated accident.

The proposed change presents no change in the consequence of an accident, since the containment high range accident monitors are used post-accident to determine the amount of core damage and status of the fission product barriers.

The containment high range accident monitors are used post accident to assess the conditions inside containment. They have an automatic function to switch the subcooling margin monitor (SCMM) to "adverse" mode (i.e., it displays a more conservative indication of the amount of subcooling in the RCS). Additionally, the containment high range accident monitors provide an indication that is used post accident in determining the status of the fission product barriers. There will be no change in the operation or use of the containment high range accident monitors.

The remaining change is editorial in nature and does not impact the accident analysis in any manner.

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

Response: No The proposed change is a minor change that is administrative in nature.

No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed changes. No new hardware is added, existing hardware is not modified and no significant changes in 5

DOCUMENT CONTROL DESK Attachment 1 LR-N05-0392 operations are implemented. Post accident monitoring instrumentation is not associated with the initiation of an accident.

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

Response: No The proposed change does not alter the manner in which safety limits, limiting safety systems settings or limiting conditions for operation are determined. The proposed change will not alter any assumptions, initial conditions or results specified in any accident analysis.

There is no change in the containment high range accident monitor high level alarm setpoint. The ECS is functionally equivalent to the TS definition of SOURCE CHECK.

Based on the above, PSEG concludes that the proposed change presents no significant hazards under the standards set forth in 10 CFR 50.92(c),

and accordingly, a finding of 'no significant hazards consideration' is justified.

5.2 Applicable Regulatory Requirements/Criteria Relocating the containment high range accident monitors to the Accident Monitoring Technical Specification is in alignment with the technical specification basis and is consistent with Regulatory Guide 1.97

'Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident" since one of the parameters that is recommended to be monitored post accident is containment radiation levels. The proposed required surveillance requirements, mode applicability, and actions for inoperability are equivalent to the specification detailed in for Westinghouse standard technical specifications, NUREG-1431.

In conclusion, based on the considerations discussed above:

1) There is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner,
2) Such activities will be conducted in compliance with the Commissions' regulations; and
3) Issuance of the amendment will not be inimical to the common defense and security or the health and safety of the public.

6

DOCUMENT CONTROL DESK Attachment I LR-N05-0392

6. ENVIROMENTAL CONSIDERATIONS There will be no physical change to the facility. The only changes will be requiring the containment high range accident monitors to be operable in modes 1-3, as opposed to modes 1-4 and eliminating the need for a SOURCE CHECK in compliance with section 1.31 of the technical specifications. This change provides no adverse impact to off site radiological dose, as these are area radiation monitors, not effluent release monitors.

PSEG has determined the proposed amendment relates to changes in a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or relates to changes in an inspection or a surveillance requirement. The proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released off site, or (iii) a significant increase in individual or cumulative occupational exposure. Accordingly, the proposed amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22 (c) (9). Therefore, pursuant to 10 CFR 51.22(b), an environmental impact statement or environmental assessment of the proposed change is not required.

7. REFERENCES 7.1 Regulatory Guide 1.97 "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident."

7.2 NUREG-1431, Revision 3 "Standard Technical Specifications Westinghouse Plants."

7

DOCUMENT CONTROL DESK Attachment 2 LR-N05-0392 SALEM GENERATING STATION UNIT 1 and UNIT 2 FACILITY OPERATING LICENSE NO. DPR-70 and NO. DPR-75 DOCKET NO. 50-272 and NO. 50-311 REVISIONS TO THE TECHNICAL SPECIFICATIONS TECHNICAL SPECIFICATION PAGES WITH PROPOSED CHANGES The following Technical Specifications for Facility Operating License DPR-70 are affected by this change request:

Technical Specification Page 1.0 1-6 3/4.2.2.1 3/4 2-6 3/4.3.3.1 314 3-36 and 3-38 3/4.3.3.7 3/4 3-55, 3-56A and 3-57A TECHNICAL SPECIFICATION PAGES WITH PROPOSED CHANGES The following Technical Specifications for Facility Operating License DPR-75 are affected by this change request:

Technical Specification Page 1.0 1-6 3/4 3.3.1 3/4 3-39 and 3-41 3/4 3.3.7 3/4 3-51A, 3-51C and 3-52A

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.

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

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 lengthrod cluster assemblies (shutdown-and control) are fully inserted except for the single. rod .cluster assembly of highest reactivity worth which is asumed to be FULLY WITiDRAWN.

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 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 I 1-6 Amendment No. 265t%-

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIR.ETS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 Fxy shall be evaluated to determine if FQCZ) is within its limit by:

a. Using the movable incore detectors to obtain a power distribution map:
1. When THERMAL POWER is < 25%. bit > 5t of RATED THERMAL POWER, or
2. When the Power Distribution Monitoring System (PDMS) is inoperable; and increasing the Measured Fq(Z) by the applicable manufacturing and measurement uncertainties at specified in the COLR.
b. Using the PDMS orv-the moveable ificore detectors when THERMAL POWER is > 25% of RATED THERMAL POWER. and increasing the measured F0 (Z) by the applicable manufacturing and measurement uncertainties as specified in the COL.R.

C. Comparing the Fxy computed (Fxy) obtained in b, above to:

1. The Fxy limits for RATED THERMAL POWER (FRT y) for the (43 appropriate measured core planes given in a and f below, and
2. The relationship: ckte ro -

sF L- _ RTP lT" g1PR._j b F j

,Y -Y XYll Y /

where F"xy is the limit for fractional THERMAL POWER operation expressed as a function of FR PXy rPxy is the power factor multiplier for FYY in the COLR. and P is the fraction of RATED THERMAL POWER at which Fxy was measured.

d. Remeacuring FXY according to the following schedule:.
1. When FCxy is greater then the FxTp limit for the appropriate measured core plane but less than the FLxy relationship, additional core power distribution measurements shall be taken and FCxy compared to F tyl and F;Xy:

a) Either within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding by 20% of RATED THERMAL POWER or greater, the THERMAL POWER at which FCty was last determined, or SA-LEM - UNIT 1 3 /4 2- 6 Amendment No. 237

TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION MINIMUM CHANNELS APPLICABLE ALARM/TRIP MEASUREMENT INSTRUMENT OPERABLE MODES SETPOINT RANGE ACTION

1. AREA MONITORS
a. Fuel Storage Area - 5mR
2. PROCESS MONITORS
a. Containment
1) Gaseous Activity 1# 1,2,3,4&5 per ODCM Control 3.3.3.9 . 101-105 cpm 23 a) Purge & Pressure -

Vacuum Relief Isolation b) RCS Leakage 1 1j2,3&4 N/A 101- 106 cpm 20 Detection

2) Air Particulate Activity a) (NOT USED) b) RCS Leakage 1 1,2,3&4 N/A . 101-106 cpm 20 Detection With fuel in the storage pool or building.
  1. The plant vent noble gas monitor may also function In this capacity when the purge/pressure-vacuum relief Isolation valves are open.

SALEM - UNIT 1 3/4 3-36 Amendment No. 263

. TABLE 4.3-3

. RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES IN WHICH CHANNELS SOURCE CHANNEL FUNCTIONAL SURVEILLANCE INSTRUMENT CHECKS CHECKS CALIBRATION TEST REQUIRED

1. AREA MONITORS
a. Fuel Storage Area S M R 0
b. Confairm'nfAron RM . 1, 2 ,a&4
2. PROCESS MONITORS
a. Containment Monitors
1) Gaseous Activity a) Purge & Pressure S M R 0 1,2,3,4 & 5 Vacuum Relief I Isolation b) RCS Leakage S M R Q 1,2,3&4 Detection
2) Air Particulate Activity.

a) (NOT USED)

I b) RCS Leakage . S M R a 1, 2, 3 & 4 Detection

'With fuel in the storage pool or building.

SALEM - UNIT i 3/4 3-36 Amendment No. 263 I

TABLE 3.3-11 (CONTINUED)

ACCIDENT MONITORING INSTRUMENTATION REQUIRED MINIMUM NO. OF NO. OF INSTRUMENT CHANNELS CHANNELS ACTION

13. PORV Block Valve Position Indicator 2/valve** 1 1, 2
14. Pressurizer Safety Valve Position 2/valve** 1 1, 2 Indicator
15. Containment Pressure - Narrow Range 2 11, 2
16. Containment Pressure - Wide Range 2 1 7, 2
17. Containment Water Level - 2 1 7, 2 Wide Range
18. Core Exit Thermocouples 4/core quadrant 2/core quadrant 1, 2
19. Reactor Vessel Level Instrumentation 2 1 8, 9 J System (RVLIS) P 4
20. Co~Ao t~ e-'r Kil /2f '-of ff,

(*) Total number of channels is considered to be two (2) with one (1) of the channels being any one (1) of the following alternate means of determining PORV, PORV Block, or Safety Valve position: Tailpipe Temperatures for the valves, Pressurizer Relief Tank Temperature Pressurizer Relief Tank Level OPERABLE.

SALEM - UNIT 1- 3/4 3-55 Amendment No. 191 l

TABLE 3.3-11 (continued)

TABLE NOTATION 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 SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ACTION 7 With the number of OPERABLE channels-one less than the Required Number of Channels shown in Table 3.3-11, operation may proceed until the next CHANNEL CALIBRATION (which shall be performed upon the next entry into MODE 5, COLD SHUTDOWN).

ACTION 8 With one RVLIS channel inoperable, restore the RVLIS channel to OPERABLE status within 30 days, or submit a special report in accordance with Specification 6.9.4.

ACTION 9 With both RVLIS channels inoperable, restore one channel to OPERABLE status within 7-days or submit a special report in accordance with Specification 6.9.4.

r re k&Ig. /ess 7-n ACT'4OA 10 W.% 'e A L^ er C4 Sac A.I ACo -an? C 7 1-,< /S oPaSdv -N A ZE e2 i: k; 4 by1 j,  ;#- ./ re 0 ?7r "t-%

'I.- .e, M e 7-ev o- /I,f cl.,? 74c- erprlvol;f.-ft-e WI n': -72 4dv5 Jt I)  ;

fr4T e /*-c_ I4,,(5J Tt.) $'

5 T--4 z, - w:r.-- 7 day&, DP Thc eve.?r / , c5 Z) ere P qre- ad ' *5A!J*^'- -

Gape c.-' 112C feo-7rr)

G. °(. 2-GA (OM el'4 .5 ,'o.1 K CO a4 7^ 7 7- Sfp-c ;-,r A-r e ae-I-

/ z/drYS -o /,

A ,407

-S .t 11 1r en , ede47 LCfw r

nex-(TheX /IC e/ ¶5 A,' i I"n ,e 1aL ; t1'/

C,4 -li/e-4r- I- s 77-,'7 Sc~e~de-5tains.

SALEM - UNIT I 3/4 3-56a Amendment No. 191 l

TABLR 4.3-11 (Continued)

SURVEILLANCE REQUIRMIENTS FOR ACCIDENW MHNITORING INSTRUMENTATION CHANNEL CKAMNL CHANNEL FUNCTIONAL INSTRUMENT CHECK CAIATION TEST

12. PORV Position Indicator M A.

N.A.

13. PORV Block Valve Position Indicator M N.A.
14. Pressurizer Safety Valve Position H N.A. R Indicator
15. Containment Pressure - Narrow Range . M R N.A.
16. Containment Pressure - Wide Range H R N.A;
17. Containment Water Level - Wido Range R** N.A.
18. Core Exit Theraccouplea R N.A.
19. Reactor Vessel Level Instrumentation H N.A.

Syston (RVLIS.

Zo CoA,:+teit /j.'k 4.rv ,GC" *t g

Unless the block valVe is closed in order to meet the requirezants of Action b, or c in specification 3.4.3.

  • i A one-time extension to this surveillance requirement is granted during fuel cycle thirteen allowing Unit 1 operations to continue to the thirteenth refueling outage (1R13). The surveillance is to be completed at the appropriate time during the 1R13 outage, prior to the unit returning to Hode 4 upon outage completion.

SALEM - UNIT 1 3/4 3-S7a Amendment No. 22Z

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 chainel sensor until loss of stationary gripper-coil voltage-.

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-aid control) are fully inserted except for the single rod cluster assenibly 6o highest reactivity worth which is assumed to be FOLLY WITsDRAWN.--

SITE BOUNDARY

-, 1r -'

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 litative assessment of channel response, 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 2 1-6 Amendment No. 215

TABLE 3.3-6

. RADIATION MONITORING INSTRUMENTATION MINIMUM CHANNELS APPLICABLE ALARM/TRIP MEASUREMENT.

INlCTCl IRACKIT Hi1IJ 1ll ,I . _1 v..JI

  • nlrPFAnI 1 __

MlnDp5 .SFTPOINT RANGE ACTION 1

1. AREA MONITORS
a. Fuel Storage Area 1slrR/hr 10 -104 mR/hr 23 93_q -d i l" 1
2. PROCESS MONITORS
a. Containment
1) Gaseous Activity 1# . 1,2,3,4&5 per ODCM Control 3.3.3.9 1 0 cpm 26 a) Purge & Pressure -

Vacuum Relief Isolation b) RCS Leakage 1 1,2,3&4 .. N/A 10I 1_16 CpM 24 Detection

2) Air Particulate Activity a) (NOT USED)

I b) RCS Leakage 1 1,2,3&4 N/A 101-1 06 cpm 24 Detection With fuel in the storage pool or building. .

  1. The plant vent noble gas monitor may also function In this capacity when the purge/pressure-vacuum relief isolation valves are open.

'. I1 I .

SALEM - UNIT 2 . 314 3-19 ' . Amendment No. 245

.i I

TABLE 4.3-3.

RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES IN WHICH CCHANNELS , SOURCE

  • CHANNEL FUNCTIONAL SURVEILLANCE INSTRUMENT . CHECKS* CHECKS CALIBRATION TEST REQUIRED
1. AREA MONITORS
a. Fuel Storage Area S M R t ntimraS. M R._Q1, 2, 3 &4
2. PROCESS MONITORS
a. Containment Monitors
1) Gaseous Activity
    • a) Purge & Pressure . S ' M . R 0 1,2,3, 4 & 5 Vacuum Relief Isolation b) RCS Leakage M F1 Q 1,2,3 & 4 Detection . .
2) Air Particulate Activity a) (NOT USED)'

I b) RCS Leakage S M Q 1, 2, 3 & 4 Detectioh

'With fuel in the storage pool or building.

SALEM - UNyFT 2 3/4 3-41

  • Amendment No. 245 I

TABLE 3.3-11 (Continued)

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

13. PORV Block Valve Position Indicator 2/valve** 1 1, 2
14. Pressurizer Safety Valve Position 2/valve** 1 1, 2 Indicator
15. Containment Pressure - Narrow Range 2 1 1, 2
16. Containment Pressure -. Wide Range 2 1 7, 2
17. Containment Water Level - Wide Range 2 1 7, 2
18. Core Exit Thermocouples 4/core quadrant 2/core quadrant .1, 2
19. Reactor Vessel Level Instrumentation 2 1 8, 9 System (RVLIS) tO I 20 CrjqJ7 .f6 £ '2<iS ?4b

(**) Total number of channels is considered to be two (2) with one (1) of the channels being any one (1) of the following alternate means of determining PORV, PORV Block, or Safety Valve position: Tailpipe Temperatures for the valves, Pressurizer Relief Tank Temperature Pressurizer Relief Tank Level OPERABLE.

SALEM - UNIT 2 3/4 3-51a Amendment No 174 I

TABLE 3.3-11 (continued)

TABLE NOTATION 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 SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ACTION 7 With the number of OPERABLE channels one less than the Required Number of Channels shown in Table 3.3-11, operation may proceed until the next CHANNEL CALIBRATION (which shall be performed upon the next entry into MODE 5, COLD SHUTDOWN).

ACTION 8 With one RVLIS channel inoperable, restore the RVLIS channel to OPERABLE status within 30 days, or submit a special report in accordance with Specification 6.9.4.

ACTION 9 With both RVLIS channels inoperable, restore one channel to OPERAB:E status within 7 days or subniit a;special report in accordance with Specification 6.9.4.

Chac'7ilef /' r:j Agq-4 C.TiA 10

w. reuuRi) -bye re d- .- ? -4 "7 c4I 7vlIs A4 0A, TV/& t .} ~e e 0 fA1 C(Jijoat'e-- e I>°v.1W7lc .e'ic - Ws/, ~i' 77 2 1 o-a s CIA)  :

rj61a42e 1 (sJ 1 I- e_

I) E. t- fe-4 7 ). ys o{ A e- e or/

cPvZ? LI .57';-tra- A,1`

z)  ? ee -e A) -4.#5 ta rep_ Co." / 7p

- _.2e 4-C' 6 7I.S Co ' .ss.3e I--- -

fr' 0o1 7 /. Hi>

h4/A,.? /1I o a yj C

.r C.-I V

. to c~se

_kc, co4 -C- IA've f e,6:1.

-qr0iIte P/R".s 1J ScrAe-Ld e. G- res70-. As SaysF- , v c>prm8CF 5-cs, S7£ SALEM - UNIT 2 3/4 3-51c Amendment No. 174 l

TABLE 4.3-11 (Continued)

SURVEILLANCE REQUIREMENTS FOR ACCIDENT MONITORING INSTRUMENTATION CUUNNEL CHANNEL CHANNEL FUz CTIONAL INSTRUMENT CHECKS CALIBRATION TEST

12. PORV Position Indicator M N.A. _.-
13. PORV Block Valve Position Indicator M N.A. 2*
14. Pressurizer Safety Valve Position M N.A. R Indicator
15. Containment Pressure - Narrow Range M R. N.A.
16. Containment Pressure - Wide Range M R N.A.
17. Containment Water Level - Wide Range M R N.A.
18. Core Exit Thermocouples M R N.A.
19. Reactor Vessel Level Instrumentation M R N.A.

System (RVLIS) 7o. Ct,' f7H~rc- 1?,fC6, j*Cckbf,, 5 /2%

,2A~2*7711"dA-r1vvL-

  • Unless the block valve is closed in order to meet the requirements of Action b, or c in specification 3.4.5.

SALEM - UNIT 2 3/4 3-52a Amendment No. 206

DOCUMENT CONTROL DESK Attachment 3 LR-N05-0392 SALEM GENERATING STATION UNIT I and UNIT 2 FACILITY OPERATING LICENSE NOS. DPR-70 and DPR-75 DOCKET NOS. 50-272 and NO. 50-311 REVISIONS TO THE TECHNICAL SPECIFICATIONS TECHNICAL SPECIFICATION PAGES WITH PROPOSED CHANGES The following Technical Specifications for Facility Operating License DPR-70 are affected by this change request:

Technical Specification Page 1.0 1-6 3/4.2.2.1 314 2-6 3/4.3.3.1 3/4 3-36 and 3-38 3/4.3.3.7 3/4 3-55, 3-56A and 3-57A TECHNICAL SPECIFICATION PAGES WITH PROPOSED CHANGES The following Technical Specifications for Facility Operating License DPR-75 are affected by this change request:

Technical Specification Page 1.0 1-6 3/4 3.3.1 3/4 3-39 and 3-41 3/4 3.3.7 3/4 3-51A, 3-51C and 3-52A

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.

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.

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 I 1-6 Amendment No. 234

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 FXY shall be evaluated to determine if FQ(Z) is within its limit by:

a. Using the movable incore dietectors to obtain a power distribution map:
1. When THERMAL POWER is
  • 25%, but > 5% of RATED THERMAL POWER, or
2. When the Power Distribution Monitoring System (PDMS) is inoperable; and increasing the Measured F0 (Z) by the applicable manufacturing and measurement uncertainties as specified in the COLR.
b. Using the PDMS or the moveable incore detectors when THERMAL POWER is > 25% of RATED THERMAL POWER, and increasing the measured FQ(Z) by the applicable manufacturing and measurement uncertainties as specified in the COLR.
c. Comparing the Fy computed (Fcxy) obtained in b, above to:
1. The Fy limits for RATED THERMAL POWER (F.,R>) for the appropriate measured core planes given in e and f below, and
2. The relationship:

FX' FXYRT [1+PFXY (-P)]

where F,,L is the limit for fractional THERMAL POWER operation expressed as a function of FXJrp PF'y is the power factor multiplier for FXy in the COLR, and P is the fraction of RATED THERMAL POWER at which FXy was measured.

d. Remeasuring FXY according to the following schedule:
1. When FCx, is greater than the F":X1 limit for the appropriate measured core plane but less than the FLXY relationship, additional core power distribution measurements shall be taken and F'.Y compared to FRTPX and FLxy:

a) Either within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding by 20% of RATED THERMAL POWER or greater, the THERMAL POWER at which FcXY was last determined, or SALEM - UNIT I 3/4 2-6 Amendment 237

TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION MINIMUM CHANNELS APPLICABLE ALARM/TRIP MEASUREMENT TNSTPTRMENT OPPRABTL MODES SETPOINT RANGE ACTTION

1. AREA MONITORS
a. Fuel Storage Area 1 * *15 mR/hr 1 0 -l-104 mR/hr 19
2. PROCESS MONITORS
a. Containment
1) Gaseous Activity 1 1,2,3,4&5 per ODCM 101-106 cpm 23 a) Purge & Pressure - Control 3.3.3.9 Vacuum Relief Isolation b)RCS Leakage 1 1,2, 3&4 N/A 101-106 cpm 20 Detection
2) Air Particulate Activity a) (NOT USED) b) RCS Leakage 1 , 1,2,3&4 N/A 101-106 cpm 20 Detection
  • With fuel in the storage pool or building.

SALEM - UNIT 1 3/4 3-36 Amendment No. 263

TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES IN WHICH CHANNELS SOURCE CHANNEL FUNCTIONAL SURVEILLANCE INSTRUMENT CHECKS CHECKS CALIBRATION TEST REQUIRED

1. AREA MONITORS M *
a. Fuel Storage Area S R Q
2. PROCESS MONITORS
a. Containment Monitors
1) Gaseous Activity a) Purge & Pressure S M R Q 1, 2, 3, 4 & 5 Vacuum Relief Isolation b) RCS Leakage S M R Q 1, 2, 3 & 4 Detection
2) Air Particulate Activity a) (NOT USED) b) RCS Leakage S M R Q 1, 2, 3 & 4 Detection
  • With fuel in the storage pool or building.

SALEM - UNIT 1 3/4 3-38 Amendment No. 263

TABLE 3.3-11 (CONTINUED)

ACCIDENT MONITORING INSTRUMENTATION REQUIRED MINIMUM NO. OF NO. OF INSTRUMENT CHANNELS CHANNELS ACTION

13. PORV Block Valve Position Indicator 2/valve** 1 1, 2
14. Pressurizer Safety Valve Position 2/valve** 1 1, 2 Indicator
15. Containment Pressure - Narrow Range 2 1 1, 2
16. Containment Pressure - Wide Range 2 1 7, 2
17. Containment Water Level - 2 1 7, 2 Wide Range
18. Core Exit Thermocouples 4/core quadrant 2/core quadrant 1, 2
19. Reactor Vessel Level Instrumentation 2 -1 8, 9 System (RVLIS)
20. Containment High Range Accident 2 2 10 Radiation Monitor

(**) Total number of channels is considered to be two (2) with one (1) of the channels being any one (1) of-the following alternate means of determining PORV, PORV Block, or Safety Valve position: Tailpipe Temperatures for the valves, Pressurizer Relief Tank Temperature Pressurizer Relief Tank Level OPERABLE.

SALEM - UNIT 1 3/4 3-55 Amendment No. 191

TABLE 3.3-11 (continued)

TABLE NOTATION 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.SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ACTION 7 With the number of OPERABLE channels one less than the Required Number of Channels shown in Table 3.3-11, operation may proceed until the next CHANNEL CALIBRATION (which shall be performed upon the next entry into MODE 5, COLD SHUTDOWN).

ACTION 8 With one RVLIS channel inoperable, restore the RVLIS channel to OPERABLE status within 30 days, or submit a special report in accordance with Specification 6.9.4.

ACTION 9 With both RVLIS channels inoperable, restore one channel to OPERABLE status within 7 days or submit a special report in accordance with Specification 6.9.4.

ACTION 10 With the number of OPERABLE Channels less than required by the minimum channels OPERABLE requirements, initiate the preplanned alternate method of monitoring the appropriate parameter within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and:

1) either restore the inoperable Channel(s) to OPERABLE status within 7 days of the event, or
2) prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 days following the event outlining the actions taken, the cause of the inoperability and the plans and schedule for resotring the system to OPERABLE status.

SALEM - UNIT I 3/4 3-56a Amendment No. 191

TAB] CiE 4.3-11 (Continued)

SURVEI'LLANCE REQUIREMENTS FOR ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHANNEL CHANNEL FUNCTIONAL INSTRUMENT CHECK CALI1BRATION TEST

12. PORV Position Indicator M N.A. R
13. PORV Block Valve Position Indicator M N.A. Q*
14. Pressurizer Safety Valve Position M N.A. R Indicator
15. Containment Pressure - Narrow Range M N.A. N.A.
16. Containment Pressure - Wide Range M R N.A.
17. Containment Water Level - Wide Range M R N.A.
18. Core Exit Thermocouples M R N.A.
19. Reactor Vessel Level Instrumentation M R N.A.

System (RVLIS)

20. Containment High Range Accident Radiation S R Q Monitor
  • Unless the block valve is closed in order to meet the requirements of Action b, or c in specification 3.4.3.
    • A one-time extension to this surveillance requirement is.granted during fuel cycle thirteen allowing Unit 1 operations to continue to the thirteenth refueling outage (1R13). The surveillance is to be completed at the appropriate time during the lR13 outage, prior to the unit returning to Mode 4 upon outage completion.

SALEM - UNIT 1 3/4 3-57a' Amendment No. 222

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.

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.

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 2 1-6 Amendment No. 215

TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION MINIMUM CHANNELS APPLICABLE ALARM/TRIP MEASUREMENT T1NrTMrWPM adz - o PDETvDART MODrE D*VLL 1qW.TPOTNT R ANGE v CTTION SLL

1. AREA MONITORS 4
a. Fuel Storage Area 1
  • 515 mR/hr 10--1_0 mR/hr 23
2. PROCESS MONITORS
a. Containment
1) Gaseous Activity 1,2,3,4&5 per ODCM 101 -10r,cpm 26 a) Purge & Pressure Control 3.3.3.9 Vacuum Relief Isolation b) RCS Leakage 1 1,2,3&4 N/A l01-106, cpm 24 Detection
2) Air Particulate Activity a) (NOT USED) b) RCS Leakage 1 1,2,3 &4 N/A 101-10s cpm 24 Detection
  • With fuel in the storage pool or building.
  1. The plant vent noble gas monitor may also function in this capacity when the purge/pressure-vacuum relief isolation valves are open.

SALEM - UNIT 2 3/4 3 -39 Amendment No. 245

TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS

1. AREA MONITORS
a. S M R
  • Fuel Storage Area Q
2. PROCESS MONITORS
a. Containment Monitors
1) Gaseous Activity a)Purge & Pressure S M R Q 1, 2, 3, 4 & 5 Vacuum Relief Isolation b) RCS Leakage S M R Q 1, 2, 3 & 4 Detection
2) Air Particulate Activity a) (NOT USED) b) RCS Leakage S M RI Q 1, 2, 3 & 4 Detection
  • With fuel in the storage pool or building.

SALEM - UNIT 2 3/4 3-41 Amendment No. 245

TABLE 3.3-11 (Continued)

ACCIDENT MONITORING INSTRUMENTATION REQUIRED MINIMUM NO. OF NO. OF INSTRUMENT CHANNELS CHANNELS ACTION

13. PORV Block Valve Position Indicator 2/valve** 1 1, 2
14. Pressurizer Safety Valve Position 2/valve** 1 1, 2 Indicator
15. Containment Pressure - Narrow Range 2 1 1, 2.-
16. Containment Pressure - Wide Range 2 1 7, 2
17. Containment Water Level - Wide Range 2 1 7, 2
18. Core Exit Thermocouples 4/core quadrant 2/core quadrant 1, 2
19. Reactor Vessel Level Instrumentation 2 1 8, 9 System (RVLIS)
20. Containment High Range Accident 2 2 10 Radiation Monitor

(**) Total number of channels is considered to be two (2) with one (1) of the channels being any one (1) of the following alternate means of determining PORV, PORV Block, or Safety Valve position: Tailpipe Temperatures for the valves, Pressurizer Relief Tank Temperature Pressurizer Relief Tank Level OPERABLE.

SALEM - UNIT 2 3/4 3-51a Amendment No. 174

TABLE 3.3-11 (continued)

TABLE NOTATION 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 SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ACTION 7 With the number of OPERABLE channels one less than the Required Number of Channels shown in Table 3.3-11, operation may proceed until the next CHANNEL CALIBRATION (which shall be performed upon the next entry into MODE 5, COLD SHUTDOWN).

ACTION 8 With one RVLIS channel inoperable, restore the RVLIS channel to OPERABLE status within 30 days, or submit a special report in accordance with Specification 6.9.4.

ACTION 9 With both RVLIS channels inoperable, restore one channel to OPERABLE status within 7 days or submit a special report in accordance with Specification 6.9.4.

ACTION 10 With the number of OPERABLE Channels less than required by the minimum channels OPERABLE requirements, initiate the preplanned alternate method of monitoring the appropriate parameter within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and:

1) either restore the inoperable Channel(s) to OPERABLE status within 7 days of the event, or
2) prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 days of the event outlining the actions taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

SALEM - UNIT 2 3/4 3-51c Amendment No. 174 I

4 TABLE 4.3-11 (Continued)

SURVEILLANCE REQUIREMENTS FOR ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHANNEL CHANNEL FUNCTIONAL INSTRUMENT CHECKS C]'ALIBRAT ION TEST

12. PORV Position Indicator M N.A. R
13. PORV Block Valve Position Indicat:or M N.A. Q*
14. Pressurizer Safety Valve PositioiI M N.A. R Indicator
15. Containment Pressure - Narrow Raiige M R N.A.
16. Containment Pressure - Wide Range M R N.A.
17. Containment Water Level - Wide Riinge M R N.A.
18. Core Exit Thermocouples M R N.A.
19. Reactor Vessel Level Instrumentat :ion M R N.A.

System (RVLIS)

20. Containment High Range Accident S R Q Radiation Monitor
  • Unless the block valve is closed in order to meet the requirements of Action b, or c in specification 3.4.5.

SALEM - UNIT 2 3/4 3-52a Amendment No. 206

Retrieved from "https://kanterella.com/w/index.php?title=LR-N05-0392,_Request_for_Change_to_Technical_Specifications_Accident_Monitoring_Instrumentation&oldid=2590476"