View source for AECM-83-0565, Forwards marked-up Proposed Changes to Tech Specs Per 821020 Confirmation of Action Ltr.Changes Intended to Enhance Clarity or Provide Consistency W/Plant Design & Operation. Justification for NSHC Encl

{{Adams
| number = ML20080F505
| issue date = 09/09/1983
| title = Forwards marked-up Proposed Changes to Tech Specs Per 821020 Confirmation of Action Ltr.Changes Intended to Enhance Clarity or Provide Consistency W/Plant Design & Operation. Justification for NSHC Encl
| author name = Dale L
| author affiliation = MISSISSIPPI POWER &amp; LIGHT CO.
| addressee name = Denton H
| addressee affiliation = NRC OFFICE OF NUCLEAR REACTOR REGULATION (NRR)
| docket = 05000416
| license number = 
| contact person = 
| document report number = AECM-83-0565, AECM-83-565, NUDOCS 8309190313
| document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE, UTILITY TO NRC
| page count = 137
}}

=Text=
{{#Wiki_filter:'
MISSISSIPPI POWER & LIGHT COMPANY
                        ]                          Helping Build Mississippi adMAME             P. O. B O X 16 4 0 J A C K S O N. MISSISSIPPI 39205 NUCLEAR PRODUCTION DEPARTMENT September 9, 1983 U. S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, D.C.         20555 Attention:       Mr. Harold R. Denton, Director                                           ;

==Dear Mr. Denton:==

==SUBJECT:==
Grand Gulf Nuclear Station Unit 1 Docket No. 50-416 License No. NPF-13 File 0260/L-860.0 Transmittal of Proposed Changes to Grand Gulf Technical Specifications AECM-83/0565 Consistent with NRC direction provided to Mississippi Power & Light (MP&L) in the " Confirmation of Action" letter, Mr. J. P. O'Reilly to Mr. D. C.
Lutken, dated October 20, 1982 (MAEC-82/242), MP&L is providing for your review and approval proposed changes to the Grand Gulf Nuclear Station Technical Specifications. These changes result primarily from MP&L's review of the technical specifications and surveillance procedures and are intended, in general, to enhance clarity or provile consistency with the plant design and operation. These changes have been J-viaved crd accepted by the Plant Safety Review Committee (PSRC) and the Safety Review Ccmmittee (SRC).
The subject changes have been evaluated by MP&L in accordance with the guidelines of 10 CFR 50.92. On the basis of that evaluation, MP&L has determined that the subject changes involve no "significant hazard consideration." Justification for this determination is provided in the attached information.
Following your review and authorization to incorporate the requested changes, MP&L will implement the affect specifications. Immediate implementation following NRC review and approval is MP&L's intention except in cases where the subject changes must await (1) the implementation of a design modification or (2) the development of implementing surveillance procedures.
In general, every effort will be made to accomplish the expeditious implementation of the requested changes following URC review and approval.
Yours truly,
  ' 309190313 830909 DR ADOCK 05000 h                                \
L.  . Dale Manager of Nuclear Services GWD/JGC: sap                                                                         !  .
Attachment                                                                              I cc:     (See Next Page)
Member Middle South Utilities System

                      .                                                                                                                                                                                                              AECM-83/0565 MICCl;21PPI POWER O LIENT COMPANY                                                   #88 cc: Mr. J. B. Richard (w/o)
Mr. G. B. Taylor (w/o)
                                                       .Mr. R. B. McGehee (w/o)
Mr. T. B. Conner (w/o)
Mr. J. P. O'Reilly (w/a)
Regional Administrator Region II 101 Marietta Street, N.W.,                                                                           Suite 2900 Atlanta, Georgia 30303 Mr. R. C. DeYoung, Director (w/a)
Office of Inspection & Enforcement                                                                                                                                                       c 4                                                            U. S. Nuclear Regulatory Commission Washington, D.C. 20555 Dr. Alton B. Cobb (w/4.)
State Health Officer State Board of IIealth                                                                                                                                                                 <
Box 1700 Jackson, Mississippi 39205 i
)
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TRANSMITTAL OF PROPOSED CHANGES TO GRAND GULF TECHNICAL SPECIFICATIONS
: 1. (GGNS - 687) (Resubmittal of Item 1 AECM-83/0411)

==SUBJECT:==
Technical Specification Table 3.3.7.9-1 and Technical Specification 3.3.7.9, pages 3/4 3-76 through 3/4 3-80.
DISCUSSION:   Technical Specification Table 3.3.7.9-1 provides a listing of fire detection instrumentation. The present table is ordered by building and room numbers within a building whereas the Specification (3.3.7.9) governing the table is ordered by zones. One of the changes to Table 3.3.7.9-1 is to order the table by building, zone numbers within the building, and room numbers within the zone. The present listing of fire detection instruments does not distinguish them as either early warning or actuation of fire suppression systems devices. The proposed change adds Function A (early werning and notification) and Function B (actuation of fire suppression systems) to the table and to the ACTION statements of Technical Specification 3.3.7.9. Other changes to the Fire Detection Instrumentation Table 3.3.7.9-1 include the following:
: 1. Design Changes
: 2. Additions to the Table
: 3. Corrections to the Table The changes to the table are discussed in detail in the justification section below.
JUSTIFICATION: The following changes to Table 3.3.7.9-1 are the result of design changes made to the fire detection system:
: 1. In Zone 1-4 for the Control Building Area OC201 (stairwell) including one smoke detector was added to the table. Safety related cables pass through this area and appropriate fire detection was added to provide coverage.
: 2. In Zone 1-6 for the Control Building, OC216 (West Corridor) with two smoke detectors was added. Safety related cables pass through this area and appropriate fire detection was added.
: 3. In Zone 1-23 for the Control Building one smoke detector was added to provide additional coverage.
: 4. In Zone 2-9 for the Auxiliary Building, one smoke detector was added to provide additional coverage.
The following changes to Table 3.3.7.9-1 constitute additions of areas / zones (and their associated instrumentation) which were inadvertantly left out of the original Technical Specification.
: 1. Add Zone 1-3 at Elevation 93' in the Control Buf1 ding.
This zone consists of areas OC109 (Decontamination Area),
0C115 (corridor), OC116 (Hot Machine Shop), and OC117 G33sp5

O (Corridor). These areas have safety related cable and should be included in the table. A total of 12 smoke detectors is added with these areas.
: 2. Add area OC306 (Electrical Chase) in Zone 1-10 at Elevation 133' in the Control Building. Safety related cable passes through these areas.
: 3. Add area OC410 (Battery Room) in Zone 1-14 at Elevation 148' in the Control Building. Specification 3.3.7.9 covers all instrumentation in each Zone listed in Table 3.3.7.9-1 and this area is added to complete the listing for Zone 1-10.      The number of Smoke Detectors for Zone 1-10 increased from 7 to 9 as a result of this change.
: 4. Add Zone 1-12 at Elevation 133' in the Control Building.
This zone includes areas OC304 (Electrical Space) and OC305 (Electrical Space). Two additional smoke detectors are added with this change. Safety related cable passes through this zone.
: 5. Zone 1-13 at Elevation 133' in the Control Building is added since part of Unit 1 control room HVAC equipment is in this zone. Zone 1-13 includes area OC303 (HVAC Room) and adds 16 smoke detectors.
: 6. Areas OC401 (Corridor), OC408 (Corridor), and OC409 (Electrical Chase) are added to Zone 1-15 along with four additional smoke detectors. Specification 3.3.7.9 covers all instrumentation in each Zone listed in Table 3.3.7.9-1 and these areas (OC401, 408, 409) were added to complete the listing for Zone 1-15.
: 7. Zone 1-19 at Elevation 166' in the Control Building is added because safety related cable passes through this area. This zone includes area 0C514 (Locker Room) and i
adds 9 smoke detectors.
: 8. Zone 1-21 at Elevation 166' in the Control Building is added because safety related cable passes through this area. This zone includes area OC518 (Electrical Chase) and adds two stoke detectors.
: 9. Zone 1-22 at Elevation 177' in the Control Building is added because safety related cable passes through this area. This zone includes areas 0C601 (Viewing Gallery),
!                  OC603 (Emergency Dormitory), and OC608 (Technical Support) and also adds sixteen smoke detectors.
l
: 10. Areas 0C706 (West Corridor), OC709 (Electrical Chase) and OC712 (HVAC Room) are added to Zone 1-23 at Elevation 189' in the Control Building. Specification 3.3.7.9 covers all instrumentation in each zone listed in Table 3.3.7.9-1 and these areas are added to complete the listing for Zone G33sp6 i                                                         - - , _ , . , - . - . - - - . . ,

0 1-23. The number of smoke detectors in Zone 1-23 is increased from 15 to 21 with one being added as a design change and five added with areas OC706, OC709, and OC712.
: 11. Add areas 1A123 (RHR "A" Heat Ex Room), 1A129 (RHR "B" Heat Ex Room), and 1A223 (Passage) to Zone 2-4 in the Auxiliary Building. Areas 1A128 and 1A129 are separated by grating from 1A102 and 1A106, respectively, and as such smoke detectors in area 1A102 serve 1A128 and in 1A106 serve 1A129. Area 1A223 is an area already served by Zone 2-4 instrumentation and is included to complete the Zone 2-4 listing.
: 12. Add areas 1A524 (Platform) and 1A529 (FPC and CU Room) to Zone 2-9. These areas are added to complete the listing for Zone 2-9,
: 13. Add area 1A101 (Passage) to Zone 2-17. This area is currently served by Zone 2-17 instrumentation and should be included.
: 14. In Zone 2-10 for the Diesel Generator Building, add three smoke detectors due to the addition of the corridor between the Auxiliary Building and Diesel Generator Building.
: 15. The type of fire protection initiated has been added to the Heat detector column. This administrative change indicates that Halon, CO      r Deluge is actuated by the 2
heat detector.
: 16. The Control Room HVAC Intake Plenum Mounted Detectors have been added since they involve control room habitability.
: 17. PGCC Halon systems in the Control Building have been added as Section "g" of the Table. The list of Halon systems is broken down by room, elevation and also by Halon panels within the room. The panels are listed as underfloor modules /Halon panel (Example: 1H13-U713/1H13-P913) and the number of detectors associated with each Halon panel is
,                    shown.
: 18. The function designation of A or B (X or Y) is added to the Table to distinguish between those instruments that perform early warning fire detection and notification and those that also actuate fire suppression systems as well as give early warning and notification. The Function A and Function B designation is also added to the ACTION statements of Technical Specification 3.3.7.9 to provide consistency between the Technical Specification and its associated Table. The adding of the Function A and B designations follows the Standard Technical Specification format for Table 3.3.7.9-1. Due to Grand Gulf Fire Detection Instrumentation design the Standard Technical G33sp7

Specificatica ACTION statements do not apply. With the addition of function A or B designation, present footnote (2) on page 3/4 3-77 is not needed and is deleted.
Present footnote (2) is a duplication of the function A or B requirements.
The following changes to Taale 3.3.7.9-1 constitutec correction of errors in the original Technical Specification:
: 1. Area OC308 (Corridor) at Elevation 133' in the Control Building is moved from Zone 1-10 to Zone 1-11.
: 2. The number of heat detectors in OC403 - Computer Room of the Control Building is corrected from present 13 to 12.
This change does not reflect the deletion of a heat detector from the plant but only a correction to the table.
: 3. The number of Smoke Detectors in OC503 (Control Room) at Elevation 166' of the Control Building is changed from 17 to 16. This change reflects the temporary split of Unit 1 and Unit 2 Control Rooms. The detector deleted is on the Unit 2 side of the Control Room.
: 4. Area 1A211 [ North Corridor (Partial)] is added to Zone 2-2 of the Auxiliary Building. This area overlaps into Zone 2-2 and also appears in Zone 2-18 as an overlap or interface area.
: 5. Area 1A314 [ South Corridor (Partial)] is added to Zone 2-6 of the Auxiliary Building. This area overlaps into Zone 2-6 and also appears in Zone 2-19 as an overlap or interface area.
: 6. Area 1A424 [ Set Down Area (Partial)] is added to Zone 2-8.
This area overlaps into Zone 2-8 and also appears in Zone 2-7 as an overlap or interface area. The number of smoke detectors in Zone 2-7 goes from 12 to 11 and the number in Zone 2-8 goes from 24 to 25 due to Zone assignment of instrumentation.
: 7. Areas 1A122 [ South Corridor (Partial)] and 1A123 [ North Corridor (Partial)] are added to Zone 2-14 of the Auxiliary Building. These areas overlap into Zone 2-14 and also appear in Zone 2-17 as an overlap or interface area.
: 8. Zones 6-9A, 6-9B, and 6-9C for the Diesel Generator Building are corrected to 2-10, 2-11, and 2-12 respectively. Added Corridor between Diesel Generator and Auxiliary Building.
: 9. Added area numbers for Standby Service Water Pump House.
These area numbers were inadvertently omitted.
G33sp8

The format change is proposed so that the Technical Specification will more accurately reflect that the operability of each individual smoke detector affects the entire zone, not i
just the area in which it is installed.
SIGNIFICANT HAZARDS CONSIDERATION:
The changes to the Fire Detection Instrumentation Table 3.3.7.9-1 constitute additions, corrections, and changes due to design changes to the plant. The design changes add additional equipment and enhance fire detection capability. The additions to the table also enhance fire detection capability. The corrections to the table do not decrease fire detection capability but reflect actual plant systems and instrumentation arrangements. This change does not involve a reduction of safety margins and no significant increase in the probability or consequences of an accident previously evaluated is involved nor is the possibility of a new or different kind of accident from any accident previously evaluated created. Thus the proposed change to the Technical Specifications does not involve any significant hazards considerations.
G33sp8.1

L.(GGNS-697)
INSTRUMENTATION i.
FIRE DETECTION INSTRUNENTATION LIMITING CONDITIGN FOR OPERATION i
                                                                                                                                                       ~
l 3.3.7.9 - As'a minimum, the fire detection instrumentation for each fire                                                                                   f detection zone shown in Table 3.3.7.9-1 shall be OPERABLE.
APPLICABILITY: Whenever equipment protected by the fire detection instrument is required to be OPERABLE.
ACTION:                                                                                g ,, Fu J.'*      aB~
With the number of OPERABLEAfire detection instruments lessA,~                                                                    thanrothe.iMinimum   l Instruments OPERABLE requirement of Table 3.3.7.9-1:
Within 1 hour, establish a fire watch patrol to inspect t                                                           zone (s) a.
with the inoperable instrument (s) at least once per hour, unless the                                                                  ,   p instrument (s) is located inside the containment,-ee drywell, then                                                                     4  g inspect the primary containment at least once per 8 hcurs or monitor                                                                  "g the containment,and/or d                                                        air temperature at least once per hour at the locations listed i                                                  Specification  4.6.1.8 and 4.6.2.6.
or c re a ~ w e.aJ
: b. Restore the minimum number of instruments to OPERABLE status within 14 days or, in lieu of any other report required by Specification 6.9.1, prepare and submit a Special Report to the Commission pur-suant to Specification 6.9.2 within 30 days outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the instrument (s) to OPERABLE status.
: c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.7.9.1 Each of the above required fire detection instruments which are accessible during unit operation shall be demonstrated OPERABLE                                                     at least once Fire detectors        which l
per 6 months by performance of a CHANNEL FUNCTIONAL TEST.ar performance of a CHANNEL FUNCTIONAL TEST during each COLD SHUTDOWN excee 24 hours unless performed in the previous 6 months.
4.3.7.9.2 The NFPA Standard 720 supervised circuits supervision associ.ated with the detector alarms of each of the above required fire detection instruments shall be demonstrated OPERABLE at least once per 6 months.                                                                      .
GRAND GULF-UNIT 1 3/4 3-76

   '                                                                                                                                                                   i.scwr-6 e 7)
TABLE 3.3.7.9-1                                                                                       .
FIRE DETECTION INSTRUMENTATION                                                             ,
MINIMUMINSTRUdNTSOPERABLE*                      /
         \ INSTRLMENT LOCATION                                                                                                                              ZONE II) HEAT (2)   FLAME                  SM0KE I
: o.                     ntainment Building
: 1. Return Duct Mounted NA          NA        NA tectors ROOM     .        ELEV.                         ROOM NAME
: b.                Control      11 ding 1-4        6                                4
: 1. OC202         111'                          DIV I SWGR RM DIV I BATTERY RM                                                                 1-4       NA        NA                     1
: 2. OC207             l' DIV II REMOTE                                                                    1-27      1         NA                     1
: 3. OC208         111 SHUTDOWN PANEL ROOM IV I REMOTE                                                                1-27                NA                     1
: 4. OC208A      111' DOWN PANEL ROOM DIV          I BATTERY RM                                                       1-5       NA         NA                     1
: 5. OC209       111' 1-5       4          NA                     2
: 6. OC210         111'                           DIV II SWGR RM 6      NA         NA                     1
: 7. OC211         111'                            DIV II BA ERY RM 1-6        7         NA                     4
: 8. OC215         111'                            DIV II SWGR 1-10       NA         NA                    1
: 9. OC307         133'                            ELECTRICAL CHA 1-10       NA         NA                    1            -
: 10. OC308         133'                            ELECTRICAL CHASE 1-11       NA         NA                    13
: 11. OC302          133'                           HVAL EQUIP. ROOM RM                                       1-15       7          NA                    10
: 12. OC402          148'                            CABLE SPREADI 1-14      13         NA                    7
: 13. OC403          148'                            COMPUTER 1 5       2          NA                    1
: 14. OC407          148'                            INSTR.               OR GEN ROOM
: 15. OC503                                                                                                                                              NA                    17 OC504       166'                            CONT L ROOM                                                                    1-18      NA 1-23      12         NA                    14
: 15. OC702          189'                            C LE SPREADING RM 1-24                 NA                     6
: 17. OC703          189'                                ONTROL CAB. ROOM 1-23       NA        NA                     1
: 18. OC707          189'                             IMSTR MOTOR GEN. RM nt are not
* The fire de etion instruments located within the primary contai required        be OPERABLE during the performance of Type A Containmen Leakage Rate Tes    .
(1) Zones              ply only to smoke detectors.
(2) Heat              tectors provide warning and activation of automatic extinguishing sys    s.
(3)              ke detectors provide early warning capability.                                                                                                                                g (4)             eur thermocoopit.s whick men. tor a.d:ed ot.c bprodere. Will                                                                                                                     5 Pr* vide. Early warning esp 4;(g, GRAND GULF-UNIT 1                                                                    3/4 3-77 i

3 (c,sas-s s,)
TABLE 3.3.7.9-1 (Continued)
FIRE DETECTION INSTRUMENTATION                                                                 i STRUMENT LOCATION MINIMUM INSTRUMENTS OPERABLE * /
II)               SM05[3)
M No.                     ELEV.
ROOM NAME                                    ZONE (1)                  HEAT                      FLAME
: c.               iliary Building                                                                                                                 ~
2-4                        NA                       NA         1
: 1.         102    93'         RHR 'A' HT EX RM 2
: 2. IA 3       93'          RHR ' A' PLMP RM                           2-4                        NA                       NA 2-4                        NA                                  2
: 3. IA104      93'          RCIC PUMP RM 2-4                         NA                      NA         2
: 4. 1A105         3'        RHR 'B'      PUMP RM 2-4                         NA                      NA         1
: 5. IA106          9'           RHR 'B' HT EX RM 2-17                        NA                      NA         2
: 6. 1A109          93'          HPCS PUMP RM 2-17                                                 NA        1
: 7. IA111          93'          PIPING PENETRATION RM 2-14                        NA                      NA        4
: 8. IA114          93'         F       COIL AREA 2-14                        NA                      NA         1
: 9. IA115          93'         PIP         PENETRATION RM 2-                           NA                     NA         1
: 10. IA116            93'        PIPING ENETRATION RM
                                                                                             -14                        NA                     NA         4 1              11. IA117            93'        MISC. E            AREA 2-14                         NA                     NA         2
: 12.      IA118       93'        RHR     'C'   PUM ROOM 2-14                         NA                     NA         2
: 13.      1A119       93'         LPCS PUMP R00M 2-14                          NA                    NA         3
: 14.      1A120       93'         CCW PUMP AND HX A A 2-17                          NA                     NA        5
: 15.      1A121       103'        EAST CORRIDOR 2-17                         NA                     NA         3
: 16.       IA122      103'        SOUTH CORRI
                                                                                              -17                        NA                     NA         5 l              17.       1A123      103'        h0RTH COR           R 2 8                           NA                    NA         6
: 18.       IA201      119'        EAST C0 IDOR.
2-4                           NA                    NA         1
: 19. 1A202           119'        RHR ' ' HX RM 2-4                           NA                    NA         2 l               20.       1A203      119'       P       NG PEMETRATION RM 2-4                                 A               NA         2~
: 21.       1A204      119'         IPING PENETRATIOM RM 2-4                                                  NA        2
: 22. IA205            119'        PIPING PENETRATION RM 2-4                             NA                   NA         1
;                23.      1A206      l' '        RHR    'B'   HX RM 2-4                            3                    NA         2
: 24.      IA207        19'       ELECT. SWGR ROOM 2-4                             3                              2
: 25. IA208          119'        ELECT. SWGR ROOM 1
2-4                            NA                   NA
: 26.                115'        RWCU RECIRC PUMP 'A' RM NA         1, RWCU RECIRC PUMP 'B' RM                     2-4                             NA 115'
: 27.         10 NA           4 1A211                 NORTH CORRIDOR                              2-18                            NA 28                 }19' 2-2                             NA                  NA          5
                     . 1A215      119'       SOUTH CORRIDOR 2-3                              2                   NA         2
: 30.      1A219      119'        ELECT, SWGR RM GRAND GULF-UNIT 1                                     3/4 3-78
_ . _ . _ _ _              __2_..___...__________                   _ _ _ _ . . . _ .         . _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ . _ _

j,,(64Nf-$87)
TABLE 3.3.7.9-1 (Continued)
FIRE DETECTION INSTRUMENTATION kNSTRUMENTLOCATION NINIMUMINSTRUkENTSOPERABLE*  /
ELEV. ROOM NAME                      ZONE (1) HEAT (2)  FLAME  SMO[)
               \R60M'NO.
: c.      iliary Building (Continued) 2-3       NA       NA I
: 31.             20          119'    PIPING PENETRATION RM 2-3       2        NA     2
: 32.               1         119'    ELECT. SWGR RM 18,
: 33.        1A22)            119'    WEST CORRIDOR                  2-2       NA 2      NA       NA     2
: 34. IA301                   139'    NORTHEAST CORRIDOR
                                                   '                                 2-6       NA       NA     1-
: 35.        IA302            1       SOUTHEAST CORRIDOR RHR 'A' HX RM                  2-6       NA       NA     l'
: 36. IA303                   139' 2-6                NA    1.
: 37.        IA304            139'     PIPING PENETRATION RM
: 38.        IA305             139'    S AM TUNNEL                    2-20     NA        NA    4 WA   lR d
2-6       NA       NA     1-
: 39.         1A306            139'    PIP      PENETRATION RM 2-        NA       NA     1:
: 40.        1A307            139'    RHR 'B     X RM
                                                                                        -6      3        NA     2
: 41.        IA308            139'    ELECT PE TRATION RM TION RM      2-6       3 NA     2
: 42.        1A309            139'    ELECT. PENE                                              3 2-6 2-19      NA       NA     3
: 43.        1A314            139'    SOUTH CORRIDOR 2-6      NA        NA    13
: 44.         IA316           139'    NORTH CORRIDOR 2        NA     2 139'    ELECT. PENET TION RM           2 ,5
: 45.         IA318 2-5       NA       NA     1 139'    RW INSTR       EST RM
: 46. '1A319 2         2        NA     2
: 47.        3A320            139'    ELECT.      NETRATION RM l                                                                 ~
3 2-1       NA       NA
!                   48.         3A321           139'    MCC    EA 2-19      NA       NA     4
: 49.         1A322           139'    C TRIFUGAL CHILLER REA 2-19                NA     1
: 50.         IA323           139'    SGTS AREA 2-19      NA       NA     1
: 51.         IA324           1       HVAC EQUIP AREA 2-19      NA       NA     1
: 52.         IA326             39'   SGTS AREA 2
166'   NORTHEAST CORRIDOR              2-8       NA
: 53. 1A401                                                                              NA      1 2-8       NA
: 54. SA 2                    166'    STEAM TUNNEL ROOF NA      2 166'    SOUTHEAST CORRIDOR             2-8       NA
: 55.               03 2-8        NA      NA 56          1A404           166'    UNASSIGNED AREA 2-8        NA      NA      1
: 7.         3A405           166'    CNTMT VENT. EQUIP RM 2-8       NA       NA      1
: 58.         IA406           166'    CNTMT EXHAUST FILTER ANO VENT ROOM GRANO GULF-UNIT 1 3/4 3-79

j,,(6 6 NS - 4 9 7)
TABLE 3.3.7.9-1 (Continued)
FIRE DETECTION INSTRUNENTATION MSTRUNENTLOCATION                                                                                                                     MININUM INSTRUNENTS OPERABLE *       /
II)    HEAT (2)    FLANE    SM0Kd3)
              \R00N NO..,                                         ELEV.                     ROOH NAME                                    ZONE
: c.                     liary Building (Continued)                                                                                                              ,
: 59.                        07 . 166'                                     MCC AREA                                     2-8          2           NA                 ,
: 60.                            0               166'                      MCC AREA                                     2-8          2           NA        1
: 61.             IA41                          166'                       NORTH CORRIDOR                              2-8          NA                    14
: 62.             IA420                         166'                       SOUTH CORRIDOR                              2-7          NA                    4
: 63. 1A424                                     1 e'                       feT DOWN AREA                               2-7          NA          NA        2
: 64. 1A428                                     166                        WEST CORRIDOR                               2-7          NA          NA        4
: 65. IA432                                      166'                      FPC AND CU PUNP RH                          2-7                      NA        1 2-7                      NA        1
: 66. IA434                                      166'                                 SAGE 2-9          NA          NA        4
               - 67.                 1A519                         185'                       ST0                 E AREA 2-9         NA          NA        5
: 68. IA527                                     185'                       LOAD                  TER AREA 185'                       CABLE C                                         5       NA           NA       1
: 69.             1A539                                                                            E 2-13        NA           NA       6
: 70.              1A602                     208'10" STORAGE AR 2-13         NA          NA        3
: 71.              IA603                     208'10" PASSAGE 2-13         NA          NA       13
: 72. 1A604                                  208'10" FUEL HANDLING AR 2-13         NA          NA        9
: 73. IA606                                      245'                      HVAC EQUIP AREA
: d.           Diesel Generator Building l
6-9A         7           6         NA
: 1.           Unit 1 E1. 158'-0" HPCS Generator 6-           7           6         NA
: 2.           Unit 1 El. 158'-0" Bus 8 Generator                                                                                                     .
1            ignit 1 F1                          1sn'-0" 8 s A                                                  6-9C         7           6         NA Generator
: o.            Standby Service Wat                                                      Pump House 2-1         NA           NA       1
: 1. Pump House A                                                                                                                                  '1
: 2. Valve Room A                                                                                                  2-1         NA           NA 2-1          NA          NA        1
: 3. Pump House                                                                                                                                       1
: 4. Valve R                                        8                                                              2-1          NA          NA
: f.           Charcoal F ter Trains                                                                                            MA           1           NA -      NA
: 1.           Sta                 y Gas Treatment                                                       -
                                                                                                                                                                        'sto'r Wire)
Sy em Filter Train                                                                                              (Allison The iliary' Building
: 1. 139'-W' NA           1           NA Control Room Standby Fresh Air System Filter                                                                                          (Allison Thermistor W e)
Train, Control Building l
El. 133'-0" GRAND GULF-UNIT 1                                                                                    3/4 3-80
: l. Cc4Ns-6 6 7)
TABLE 3.3.7.9-1 FIRE DETECTION INSTRUMENTATION MINIMUM INSTRUMENTS OPERABLE *     ,
HEAT
                                                                            )   SMOKE INSTRUMENT LOCATION (X/Y)      (X/Y)      (X/Y)
: a. CONTAINMENT BUILDING #
: 1. Return Duct Mounted                                             3/0 Detectors ROOM      ELEV    ROOM NAME
: b. CONTROL BUILDING
: 1. Zone 1-3                                                        12/0 OC109     93'     Decontamination Area                                ,
0C115     93'     Corridor 0C116     93'     Hot Machine Shop OC117     93'     Corridor
: 2. Zone 1-4                                                        6/0 OC201     111'    Stairwell OC202     111'    Div I Swgr Rm           0/6(CO2 )
OC207     lil'    Div I Battery Rm
     *     (X/Y): X - is number of Function A (early warning fire detection and notification only) instruments.
Y - is number of Function B (actuation of fire suppression systems and early warning and notification) instruments.
     #     The fire detection instruments located within the primary containment are not required to be OPERABLE during the performance of Type A Containment Leakage Rate Tests.
(1) Smoke and flame detectors provide only early warning capability with the exception of:
(a) Zone 1-27 detectors trip closed the door between the OC208/0C208A Remote Shutdown panel rooms.
(b) Containment building return duct mounted detectors' trip the containment cooler fans.
(c) Zone 1-11 and 1-13 detectors initiate the control building purge fan system.
(d) Control Room HVAC Intake Plenum Detectors trip the control room A/C units unless a control room emergency filtration system isolation mode automatic actuation signal is present.
GRAND GULF-UNIT 1                        3/4 3-77 G33sp9

                                                                       /.(seas-4r7)
TABLE 3.3.,7.9-1 FIRE DETECTION INSTRUMENTATION MINIMUM INSTRUMENTS OPERABLE
* ROOM     ELEV    ROOM NAME                 HEAT       FLAME (   SMOKE (X/Y)      -(X/Y)    (X/Y)
: 3. Zone 1-5                                                        3/0 OC209    111'    Div III Battery Rm OC210    111'    Div III Swgr Rm           0/4(CO2 )
: 4. Zone 1-6                                                        7/0 OC211    111'    Div II Battery Rm OC215    111'    Div II Swgr Rm            0/7(CO2 )
OC216    111'    West Corridor
: 5. Zone 1-10                                                       2/0 OC306    133'    Electrical Chase OC307    133'    Electrical Chase
: 6. Zone 1-11                                                        13/0 OC302     133'   HVAC Equipment Rm OC308     133'   Corridor
: 7. Zone 1-12                                                        2/0 OC304     133'    Electrical Space OC305     133'   Electrical Space
: 8. Zone 1-13                                                       16/0 7C303    1s3'    HVAC Room
: 9. Zone 1-14                                                       9/0 OC403    148'    Computer Room             0/12(Halon)
OC410    148'    Battery Room
: 10. Zone 1-15                                                         15/0 OC401    148'    Corridor OC402    148'    Lower Cable Spreading     0/7(CO3 )
Room OC407    148'    Instr. Motor Gen Rm       0/2(CO2 )
OC408    148'    Corridor OC409    148'    Electrical Chase
: 11. Zone 1-18                                                          16/0 OC503     166'   control Rm (Unit 1 Side)
OC504     166'   U-1 Inst Rack Area l     GRAND GULF-UNIT 1                        3/4 3-78 l     G33sp10

f,(&&NS-f t ?)
TABLE 3.3.7.9-1 FIRE DETECTION INSTRUMENTATION MINIMUM INSTRUMENTS OPERABLFZ ROOM    ELEV   ROOM NAME                  HEAT       FLAME ( ) SM0KE( }
(X/Y)      (X/Y[      (X/Y)
: 12. Zone 1-19                                                       9/0 OC514    166'  Locker Room
: 13. Zone 1-21                                                        2/0 OC518    166'  Electrical Chase
: 14. Zone 1-22                                                        16/0 0C601    177'  Viewing Callery 0C603    177'  Emergency Dormitory 0C608    177'  Technical Support
: 15. Zone 1-23                                                         21/0 OC702    189'  Upper Cable Spreading Rcom                       0/12(CO2 )
OC706    189'  West Corridor OC707    189'   Instr. Motor Gen Rm OC709    189'  Electrical Chase OC712    189'  HVAC Room
: 16. Zone 1-24                                                         6/0 OC703   189'   Control Cabinet Area       4/0(CO2 )
: 17. Zone 1-27                                                         2/0 OC208    111'  Div I Remote Shutdown Panel                      0/1(CO2 )
OC208A 111'    Div II Remote Shutdown Panel                      0/1(CO2 )
: 18. Control Roon HVAC Intake Plenum Mounted Detectors                                                2/0
: c. AUXILIARY BUILDING
: 1. Zone 2-2                                                         23/0 1A211    119'   North Corridor (Partial) 1A215    119'   South Corridor 1A222   119'   West Corridor GRAND GULF-UNIT 1                      3/4 3-79 G33sp11

c
: l. [64NS -6 T7)
TABLE 3.3.7.9-1 FIRE DETECTION INSTRUMENTATION MINIMUM INSTRUMENTS OPERABLE
* ROOM      ELEV   ROOM NAME                 HEAT        FLAME    SMOKE (X/Y)        (X/Y)   (X/Y)
: 2. Zone 2-3                                                        5/0 1A219    .119'   Electrical Swgr Rm        0/2(CO2 )
1A220     119'   Piping Penetration Rm 1A221    119'   Electrical Swgr Rm        0/2(CO2 )
: 3. Zone 2-4                                                        22/0 1A102     93'    RHR "A" Heat Ex Pm 1A103   93'    RHR "A" Pump Rm 1A104    93'    RCIC Pump Rm 1A105   93'    RHR "B" Pump Rm 1A106   93'    RHR "B" Heat Ex Rm 1A128    108'  RHR "A" Heat Ex Rm 1A129    108'  RHR "B" Heat Ex Rm 1A202    119'  RHR "A" Heat Ex Rm 1A203    119'  Piping Penetration Rm 1A204    119'  Piping Penetration Rm 1A205    119'  Piping Penetration Rm 1A206    119'  RHR "B" Heat Ex Rm 1A207    119'  Electrical Swgr Rm        0/3(C0 )
1A208    119'  Electrical Swgr Rm        0/3(C0 )
1A209    115'  RWCU Recire Pump "A" Rm 1A210    115'  RWCU Recire Pump "B" Rm 1A223    128'  Passage
: 4. Zone 2-5                                                      5/0 1A318    139'  Electrical Penetration Room                      0/2(CO2 )
1A319    139'  RPV Instr Test Rm 1A320    139'   Electrical Penetration Room                      0/2(CO2 )
: 5. Zone 2-6                                                       26/0
!               1A301    139'   East Corridor 1A302    139'   Southeast Corridor
,               1A303    139'   RHR "A" Heat Ex Rm l               1A304    139'   Piping Penetration Rm l
1A306    139'   Piping Penetration Rm 1A307    139'   RHR "B" Heat Ex Rm l
1A308    139'   Electrical Penetration Room                      0/3(CO2 )
Electrical Penetration l                1A309   139' Room                      0/3(CO2 )
l i                1A314    139'  South Corridor (Partial)
I                1A316    139'  North Corridor
!     GRAND GULF-UNIT 1                     3/4 3-80 G33sp12

b(6(rNS~$tl)
TABLE 3.3.7.9-1 FIRE DETECTION INSTRUMENTATION MINIMUM INSTRUMENTS OPERABLE
* ROOM      ELEV  ROOM NAME                HEAT       FLAME     SM0KE (X/Y)      (x/Y)     (x/Y)
: 6. Zone 2-7                                                     11/0 1A420    166'  South Corridor (Partial) 1A424    166'  Set Down Area (Partial) 1A428    166'  West Corridor 1A432    166'  FPC & CU Pump Rm 1A434    166'  South Passage
: 7. Zone 2-8                                                     25/0 1A401    166'  Northeast Corridor 1A402    166'  Steam Tunnel Roof 1A403    166'  Southeast Corridor 1A404    166'  Unassigned Area 1A405    166'  Containment Vent. Equip Room 1A406     166' Containment Exhaust Filter Rm 1A407     166' MCC Area                 0/2(C0 )
1A410     166' MCC Area                 0/2(C0 )
1A417     166' North Corridor (Partial) 1A424     166'  Set Down Area (Partial)
: 8. Zone 2-9                                                      10/0 1A519    185'  Storage Area 1A524    195'  Platform 1A527    185' Load Center Area 1A529    185'  FPC & CU Rm 9   Zone 2-13                                                     31/0 1A602    208'  Storage Area l               1A603    208'  Passage 1A604    208'  Fuel Handling Area 1A606    245'  HVAC Equip Area
: 10. Zone 2-14                                                     17/0 1A114    93'   Fan Coil Area 1A115    93'   Piping Penetration Rm 1A116    93'   Piping Penetration Rm 1A117    93'   Misc Equip Area 1A118    93'   RHR "C" Pump Room GRAND GULF-UNIT 1                    3/4 3-80a G33spl3

l.G,(,NS-6t7)
TABLE 3.3.7.9-1 FIRE DETECTION INSTRUMENTATION MINIMUM INSTRUMENTS OPERABLE
* ROOM    ELEV       ROOM NAME                       HEAT       FLAME (   SMOKE (x/Y)      (x/Y)     (x/Y) 1A119   93'        LPCS Pump Room 1A120    93'       CCW Pump & Heat Ex Rm 1A122    103'      South Corridor (Partial) 1A123    103'      North Corridor (Partial)
: 11. Zone 2-15                                                               1/0 1A539   185'      Cable Chase
: 12. Zone 2-17                                                               16/0 1A101   93'       Passage 1A109   93'       HPCS Pump Rm 1A111   93'       Piping Penetration Rm 1A121   103'      East Corridor 1A122   103'      South Corridor (Partial) 1A123   103'      North Corridor (Partial)
: 13. Zone 2-18                                                               20/0 1A201   119'      East Corridor 1A211    119'     North Corridor (Partial)
: 14. Zone 2-19                                                                13/0 1A314    139'      South Corridor (Partial) 1A321    139'      MCC Area 1A322    139'      Centrifugal Chiller Area 1A323   139'      SGTS Area 1A324    139'      HVAC Equip Area 1A326    139'      SGTS Area l           15. Zone 2-20                                                                2/0 1A305   139'      Steam Tunnel
: d. DIESEL GENERATOR BUILDING
: 1. Zone 2-10                                                     6/0       3/0 l                 1D301   133'      Corridor                        0/3 (Deluge) 1D306   133'      Div III Diesel Gen Room 1D401   158'      Div III Diesel Gen Room 0/7 (Deluge)
GRAND GULF-UNIT 1                               3/4 3-80b G33sp14

l,[66Mi~$ O )
TABLE 3.3.7.9-1 FIRE DETECTION INSTRUMENTATION MINIMUM INSTRUMENTS OPERABLE
* ROOM     ELEV   ROOM NAME                HEAT             FLAME    SMOKE (X/Y)            (x/Y)    (x/Y)
: 2. Zone 2-11                                                 6/0 1D308    133'   Div II Diesel Gen Room 1D402    158'   Div II Diesel Gen Room 0/7 (Deluge)
: 3. Zone 2-12                                                 6/0 1D310    133'   Div I Diesel Gen Room 1D403    158'   Div I Diesel Gen Room    0/7 (Deluge)
: e. STANDBY SERVICE WATER PUMP HOUSE
: 1. Zone 2-1                                                           4/0 1M110    133'   SSW Pump Rm A 1M112    133'   SSW Valve Rm A 2M110    133'   SSW Pump Rm B 2M112    133'   SSW Valve Rm B
: f.   ' CHARCOAL FILTER TRAINS
: 1. Standby Gas Treatment System Filter Train                              1/0 (Allison Thermistor Wire)
Auxiliary Building E1. 139'
: 2. Control Room Standby Fresh Air System filter Train                       1/0 (Allison Thermistor Wire)
(
l Control Building El. 133' l
: g. CONTROL BUILDING (PCCC HALON SYSTEMS)
OC503    166'   Control Room (Unit 1 side)
Module /Halon Panel 1H13-U700/1H13-P900      0/10                      10/0 1H13-U701/1H13-P901      0/10                      15/0 1H13-U702/1H13-P902      0/9                       14/0 1H13-U703/1H13-P903      0/11                      17/0 l
1H13-U720/1H13/P920      0/7                       13/0 SH13-U730/1H13-P930      0/11                      12/0 1H13-U738/1H13-P938      0/10                      12/0 SH13-U739/5H13-P939      0/5                       14/0 i
3/4 3-80c t
GRAND GULF-UNIT 1 l       G33sp15

O
: l. [46HS-4s 7)
TABLE 3.3.7.9-1 FIRE DETECTION INSTRUMENTATION MINIMUM INSTRUMENTS OPERABLE *              ,
ROOM            ELEV     ROOM NAME                HEAT          FLAME ( } SMOKE (X/Y)         (X/Y)           (X/Y)
OC504            166'    Unit 1 Instrument Rack Area Module /Halon Panel 1H13-U710/1H13-P910      0/8                           15/0 1H13-U711/1H13-P911      0/8                           14/0 1H13-U712/1H13-P912      0/8                           9/0 1H13-U714/1H13-P914      0/8                           13/0 1H13-U732/1H13-P932      0/8                           14/0 1H13-U733/1H13-P933      0/8                           13/0 1H13-U734/1H13-P934      0/8                           13/0 1H13-U735/1H13-P935     0/8                           11/0 OC703            189'    Unit 1 Instrument Rack Area Module /Halon Panel 1H13-U713/1H13-E913     0/9                            15/0 1H13-U715/1H13-P915     0/8                            10/0 1H13-U717/1H13-P917     0/8                            15/0 1H13-U736/1H13-P936     0/8                            14/0 1H13-U737/1H13-P937     0/8                            10/0 GRAND GULF-UNIT 1                             3/4 3-80d G33sp16
: 2. (GCNS - 697) (Resubmittal of item f 6, AECM-83/0370)

==SUBJECT:==
Technical Specification Tables 3.3.7.1-1, 3.3.7.12-1, 4.3.7.1-1 and Technical Specification 3.11.2.7; pages 3/4 3-56, 3/4 3-58, 3/4 3-59, 3/4 3-90, 3/4 3-91, and 3/4 11-17.
DISCUSSION:    The pre-treatment and post-treatment Offgas Radiation Monitoring Instrumentation appear in the four Technical Specification Tables listed below:
: 1. Table 3.3.7.1-1, Radiation Monitoring Instrumentation
: 2. Table 4.3.7.1-1, Radiation Monitoring Instrumentation Surveillance Requirements.
: 3. Table 3.3.7.12-1, Radioactive Gaseous Effluent Monitoring Instrumentation.
: 4. 4.3.7.12-1, Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements.
The Technical Specifications that apply to each instrument are not consistent among the tables. The proposed change deletes specifications for the pre-treatment and post-treatment Offgas Radiation Monitoring Instrumentation in Tables 3.3.7.1-1 and 4.3.7.1-1 and leaves the specifications for these instruments in Tables 3.3.7.12-1 and 4.3.7.12-1. The number of minimum channels operable for the Offgas Post-Treatment Monitor is corrected from 1 to 2 and an applicable ACTION statement is added.
The change to Technical Specification 3.11.2.7 corrects a typo.
JUSTIFICATION: In order to promote consistency in the Technical Specifications concerning the specifications on the Offgas Pre-treatment and Post-treatment Radiation Monitors, these instruments should only appear in Tables 3.3.7.12-1 and 4.3.7.12-1. By placing the instruments only in these tables, conflicting ACTION statements between Tables 3.3.7.1-1 and 3.3.7.12-1 for these instruments will be eliminated. Also conflicting applicability, minimum channels OPERABLE, and Surveillance Requirements among the tables will be eliminated.
The proposed changes to '.able 3.3.7.1-1 are as follows:
: 1. Delete items 3 and 4, the Offgas Pre-treatment and Post-treatment Radiation Monitors.
: 2. Delete ACTION 71 on page 3/4 3-58, since this only applies to the Offgas Post-treatment Radiation Monitor.
The proposed changes to Table 4.3.7.1-1 are to delete items 3 and 4, the Offgas Pre-treatment and Post- treatment Radiation Monitors.
G82sp1

The proposed changes to Table 3.3.7.12-1 are to change the Minimum Operable Channels from 1 to 2 for the Offgas Post-Treatment Monitor and to modify the ACT!DN statement for this Monitor for the 2 channels instead of the present 1 channel required OPERABLE. Present Action Statement 121 for the Offgas Post-Treatment Monitor is changed to ACTION 125.
ACTION 125 has two parts: (a) is for one channel inoperable and is the same as present ACTION 71 part a for this monitor in Table 3.3.7.1-1 and part (b) is for two channels inoperable and is taken from present ACTION 121 in Table 3.3.7.12-1. Two channels are required to be OPERABLE for the Offgas Post-Treatment Monitor to perform its function.
The radiation effluent monitoring requirements for the Offgas Pre-treatment and Post-treatment Radiation Monitors have not been relaxed by this change.
The ACTION statement for Technical Specification 3.11.2.7 incorrectly lists the metastable state of Kr as Kr-87m. The correct metastable state is Kr-85m as presently stated in the first paragraph of Technical Specification 3.11.2.7.
SIGNIFICANT HAZARDS CONSIDERATION:
This change is purely administrative in nature since it is made to promote consistency in the Technical Specifications, to avoid any unnecessary confusion and to reduce the possibility of mis-interpretation of conflicting statements. This change does not involve a reduction of safety margins and no significant increase in the probability or consequences of an accident previously evaluated is involved nor is the possibility of a new or different kind of accident from any accident previously evaluated created. Thus the proposed change to the Technical Specifications does not involve any significant hazards consideration.
I G82?p2

TAttE 3.3,7.1-1 i
RetAT10N IGNITORING INSTRISIENTAT10N NIN19ED OWINELS       APPLIOWLE          Al. ant /TWIP      IEASURDENT        ,
INSTRIBENTATIfRI                 OPtRABLE          CONDITIONS         SETPotNT           RANGE                ACT10H l
Y  1.      Campenent Costins
  !          Water Itodiation                                                      5
* Monitor                   1 At all times    $1 x 10 cpedIR       10 to 106 ,,,           y,
: 2.      Staney Service Iloter System Radiation                                                                           g Monitor                   1/ heat          1, 2, 3, anda 11 x 10 5cpudIR        10 to 10 cpu            70 exchan0er train Ows.arur03
: 3. fe, l    Radiat on                                  1, 2            55 x 1         sA/hr                         70
: 4. Offees Post-treatment       I*)
to 105cpm Radiation Monitor                            ,2            ia                                           71 3
 ,Y                                                                   -
g   5. Carten Bed Voelt
                                                                        < 2.x full peuer     1 to 105 all/hr         72 l            Radiation llenitor        1                1, 2 1                                                                       Eackground/Nm S. Centrol Seen Ventile-                                                              -2 to le tmR/hr tion Radiation Monitor     2 rar           1,2,3,5 and** 14 mR/hr/              10                      73 hs sysr *h
                                                ,-                                   y l     7. Containment and Drywell
:            Ventilation Exhaust                                                                -2 to le tmit/hr                        M
!             Radiation Monitor         3(h)             At all times   <2.0 mR/hr/          10                      74
                                                                        <4 alt /hr(bN l                                                                       -
e p
: 8.      Fuel Hend11n0 Aroe to 102alt /hr i                                                                                                2 3(h)             1,2,3,5 and** i 2ndt/hr/            10                      75   r-g l             Ventilation Exhaust
                                                                         <4 mR/hrI # N l
Radiation Monitor                                                                                                           i j
Fuel Handling Area Peel
: 9.                                                                                                                                 4
                                                                                                -2 to 102aft /hr l
Sweep Enheest Radiation 3(h}
Monitor                                    (c)            i 18 udt/hr/         10                      75                   $
!                                                                        <35 aft /hr(dW
                                                                         ~
l                                                                                                                                     .

    .                                                                                                t. C&&us- m )
INSTRUMENTATION TABLE 3.3.7.1-1 (Continued)                                                                                 l RADIATION MONITORING INSTRUMENTATION l
ACTION
        ' ACTION 70                         -
                                                 'With the required monitor inoperable, obtain and analyze at                                                                    l least one grab sample of the monitored parameter at least once per 24 hours.
ACTION 71                        -
CDELETEb]
inope                   el in the downscale trip e                                           within one hour.
: b. With                  ,
required monitors inope                         -
n at least ACTION 72-                             With the required monitor inoperable, perform area surveys of the monitored area with portable monitoring instrumentation at least once per 24 hours.
ACTION,73 -                                                                k r,lg spra-                                                                          g
: a. With one of the required monitorsjinoperable, place the inoperable I channel in the'downscale tripped condition within one hour; restore the inoperable channel to OPERABLE status within 7 days, or, within the next 6 hours, initiate and maintain operation o room emergency filtration system in the isolation so ofoperation control               b("
                                                                                            ,. a                       r       **                      *  **
: b. Withbothoftherequiredmonit: ors]:pinennoperable, initiate and main- l tain operation of e control room emergency filtration system in "I                                     the isolation mode of operation within one hour.
ACTION 74                        -
: c. With one of the required monitors inoperable, place the inoperable channel in the downscale tripped condition within one hour.
: b. With two of the required monitors inoperable, isolate the l                                                  containment and drywell purge and vent penetrations within 12 hours.
l           ACTION 75                         -
: a. With on.. a f the- required monitors inoperable, place the inoperu b d W 1 in the downscale tripped condition within i                                                  one houn
: b. With two of the required monitors inoperable, initiate and maintain operation of at least one standby gas treatment subsystem within 12 hours.
1                                                               :
1 GRAND GULF-UNIT 1                                                 3/4 3-58

TABLE 4.3.7.1-1                            '
cs RADIATION IGNITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS l      l               '                                                                                 *
  '                                                                                                              OPERATIONAL CHANNEL                         CONDITIONS FOR l    h                                                        CHANNEL       FUNCTIONAL    CHANNEL WHICH SURVElttANCE T                                                                                   CALIBRATION              REQUIRED g                 INSTRUE NTATION                         CHECK _         TEST U     1. Component Cooling Water Radiation e'          Monitor                                         S               M              R At all times
: 2. Standby Service Water System Radiation Monitor                               S               M              R               1, 2, 3, and*
l     g)serrs]                                                                     "                              ;, O
: 3. - ^''g: "c; tr: ^ :;t "-ftt t; .t M;;              5 l                                                                  ";              "              "               I, ?                     .
: 4. Offy; " eat i.ed _i.t C ftt ur. ";;;ter l                                                                                  M              R               1, 2 l            5. Carbon Bed Vault Radiation Monitor              S 6     Control Room Ventilation Radiation MI *)                          1, 2, 3, 5 and**
1 R
i                  Monitor                             ,           S     .
  .          7. Containment and Drywell Ventilation Exhaust Radiation Monitor                        S              M              R               At all times l
l
* 8. Fuel Handling Area Ventilation                                                                 1, 2, 3, 5 and**
M              R
,       if         Radiation Monitor                                S g    9. Fuel Handling Area Pool Sweep M             R               (b)
Exhaust Radiation Monitor                       S                                                                               ,
: 10. Area Monitors
: a. Fuel Handling Area Monitors l
i                         1)   New Fuel Storage Vault               S               M             R               (c)
:                         2)   Spent Fuel Storage Pool              S               M             R               (d)                                     '
: 3)   Dryer Storage Area                   S               M             R               (e) l
: b. Control Room Radiation Monitor            S               M             R               At all times l   ,
j
              "    With RHR; heat exchangers in operation.            r."-/ *
* h
              **   Whenirradiatedfuelisbeinghandledinthe[secondarycontainment.
(a) The CHANNEL FUNCTIONAL TEST shall demonstrate that control room annunciation occurs if any of the    " fo110 wing
                                                                                                                                               *e conditions exist.                                                                                                           D j                  1. Instrument indicates measured levels above the alam/ trip setpoint.                                                     "
: 2. Circuit failure.                                     .
l
;                  3. Instrument indicates a downscale failure.
di l                  4. Instrument controls not in Operate mode,                                                                               *
!              (b) With irradiated fuel in the spent fuel storage pool.                                                                         O (c) With fuel in the new fuel storage vault.
I
!              (d) With. fuel in the spent fuel storage pool.                                                                                  D
!              (e) With fuel in the dryer storage area.

  @                                           TABLE 3.3.7.12-1 (Continued)
E RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUM2NTATION E
G E                                                MINIMUM CHANNELS 55          INSTRUMENT                                OPERABLE                             ACTION a                                                                           -APPLICABILITY N
: 6. OFFGAS PRE-TREATMENT MONITOR
: a. Noble Gas Activity Monitor                      1                    ***         126
: 7. OFFGAS POST-TREATMENT MONITOR
: a.  ' Noble Gas Activity Monitor u,            Providing Alarm and Automatic              .
szS 3:            Termination of Release                    ^ - 4'                    **
                                                                                              -hPb Y
8 to i
15) 9 N
i 4
h 4
v
                                                                                                      \/

2.(fr6NS-69$
TA8LE 3.3.7.12-1 (Continued) i                          RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION TABLE NOTATION
* At all times.
               **    During main condenser offgas treatment system operation.
               *** During operation of the main condenser air ejector.
ACTION 121 -      With the number of channels OPERABLE less than required by the Minmum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are taken at least once per 8 hours and these samples are analyzed for gross activity within 24 hours.
ACTION 122 -      With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided samples are continuously collected with auxiliary sampling equipment as required by Table 4.11.2.1.2-1.
ACTION 123 -       With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent release via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 8 hoursh                                                                        -
ACTION 124 -      With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, operation of main condenser offgas treatment system may continue for up to 30 days provided grab samples are collected at least once per 4 hours and analyzed within the following 4 hours.
ACTION 125 -      [0ELETE03 ACTION 126 -      With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the SJAE effluent may be released to the environment for up to 72 hours provided:
: a.        The offgas system is not bypassed, except for filtration i                                            system bypass during plant startups, and
: b.        The offgas delay system noble gas activity effluent downstream monitor is OPERABLE; Otherwise, be in at least HOT STANDBY within 12 hours.
A C T I D M I2 5 a . M rs o ,. a     d ri, e          yas :v *) mon:%rs in */* r* EI" o t'"                                                         :
rk l opsrabia p            c. i.. . ., al               1,. he      d o.vn s cale                 tv; ppa) can)lt **
i                         w .rs .~ .. . s.,,r.
                                 ], . V; rt, Lark          r% pag l,o) ~ on: tors ln oy2 ra Ll*s s.07"t=T re /** ses- v:n ti.is p.1Avey -ay cann,ax .%r .7 r-w days prov;)2) graL sa g ler ara- te kam at la.as r i                             once           f A.o.ir.s                                                  A are analyme/
for a;psrvoss sc.t:v ry wirW 2 *t k s,y J rbs n.
 <           k                                                                                         ors.
GRAND GULF-UNIT 1                              3/4 3-91
                                                                       ,m  ,,-,,,---,-,-,---,,,wa            ----,,,,---,w,-- ,-,m--w----,v,m,--ry,-,,y-        -wm,

g.(ssus- m )
RADIDACTIVE EFFLUENTS MAIN CONDENSER i
                                                      ~      ''               ' ~ '                  '
                                                                                                                    ~                 '
LIMITING CONDITION FOR OPERATION l
3.11.2.7 Theghossradioactivity(gamma)rateofthenoblegasesXe-135m, Xe-133,'Xe-135, Xe-138, Kr-85m, Kr-87, Kr-88 measured at the off gas recombiner effluent shall be limited to less than or equal to 380 millicuries /second.
APPLICABILITY:      OPERATIONAL CONDITIONS 1, 2, and 3 ACTION:
Im With the gross radio tivity rate of the noble gases Xe-135m, Xe-133, Xe-135, Xe-138, Kr-8 , Kr-8         and Kr-88 at the off gas recombiner affluent exceeding                                                              l 380 mi11 curies /second, restore the gross radioactivity release rate to within its limit within 72 hours or be in at least HOT STANDBY within the next 12 hours.                                                                                                                                                           ,
SURVEILLANCE REQUIREMENTS 1
4.11.2.7.1 The radioactivity release rate of noble gases near the outlet of the main condenser air ejector shall be continuously monitored in accordance with Specification 3.3.7.12.
4.11.2.7.2 The gross radoactivity release rate of the noble gases Xe-135m, Xe-133, Xe-135, Xe-138, Kr-85m, Kr-87, and Kr-88 from the main condenser air ejector shall be determined to be within the limits of Specification 3.11.2.7 at the following frequencies by performing an isotopic analysis of a representative sample of gases taken at the discharge (prior to dilution and/or discharge) of the main condenser air ejector:
: a. At least once per 31 days.
: b. Within 4 hours following an increase, as indicated by the Condenser                                                                                      -
Air Ejector Noble Gas Activity Monitor, of greater than 50%, after factoring out increases due to changes in THERMAL POWER level, in the nominal steady state fission gas release from the primary coolant.
e GRAND GULF-UNIT 1                         3/4 11-17                                                                                                               .
                                                                 ,,-----,a,,-   ----- --.-----a-,,.-      .m,.,,--w        e.wn__,,--   - - - ,, , , - , - - , , - - - . - , - - - - , - -
: 3. (GGNS - X23) (Resubmittal of Item 2, AECM-83/0370)

==SUBJECT:==
Technical Specification 3.7.10, page 3/4 7-45.
DISCUSSION:    Technical Specification 3.0.3 and 3.0.4 are not applicable to the specificatior. on embankment stability and this change is proposed to except these provisions in the ACTION statement of Technical Specification 3.7.10.
JUSTIFICATION: Technical Specification 3.7.10 is applicable "at all times" and as such the provisions of Technical Specification 3.0.3 do not apply. Technical Specification 3.0.3 requires that wher a Limiting Condition for Operation is not met except as provided in the associated ACTION requirements, within one hour action shall be initiated to place the unit in an OPERATIONAL CONDITION in which the Technical Specification does not apply. Since 3.7.10 is applicable "at all times" Technical Specification 3.0.3 cannot be met.      Exceptions are granted in other areas of the Technical Specifications to the provisions of 3.0.3 where these provisions do not apply.
The intent of Technical Specification 3.0.4 is to ensure that unit operation is not initiated with either required equipment or systems inoperable or other limits being exceeded.
Exceptions to these provisions have been provided for a limited number of specifications when startup with inoperable equipment would not affect plant safety. Startup with slope embankment at culvert no. 1 not within specifications will not affect plant safety and as such the requested exception should be granted.
The exceptions to 3.0.3 and 3.0.4 were inadvertently omitted when new Technical Specification 3.7.10 was submitted to the NRC for approval under cover letter AECM-83/0370.
SIGNIFICANT HAZARDS CONSIDERATION:
I Bases 3.0.3 and 3.0.4 specify the reasons for Technical I
Specification 3.0.3 and 3.0.4 and gives examples of when the provisions of these specifications are required. Exempting Technical Specification 3.7.10 from the requirements of 3.0.3 and 3.0.4 follows the examples of other technical specifications presently with the exceptions. This change does not involve the reduction of safety margins and no significant increase in the probability or consequences of an accident l
previously evaluated is involved nor is the possibility of a new or different kind of accident from any accident previously l
evaluated created. Thus the proposed change to the Technical Specification does not involve any significant hazards consideration.
i i
I l
j        G85sp4 1

i                  .
: 3. (66MS - X 2. 3 )
i ptANT SYSTEMS 3/4.7.10 EM ANKMENT STABILITY i.
LINITING CO MITION FOR OPERATION 3.7.10 The downs'tream access road slope at Culvert No. I and the drainage basin slopes shall remain stable.
APPLICABILITY: At all times.
l                ACTION:"'If Culvert No. I has blockage exceeding 15K of its cross-sectional crea, the Culvert shall be cleaned and the slope embankments verified to be stable.
: b. T h e- j>rovis'en e of Cy acikTions 3,o.3 and 2,o.y ava rear gl'alla .
SURVEILLANCE REQUIREMEh75 4.7.10 The downstream access road slope at Culvert No. I and the drainage
;                basin slopes shall be confirmed to be stable by:
: c. At least once per year, performing a visual inspection of the embankments and Culvert No. 1.
: b. At least once per five years, performing a five year survey to confirm no significant degradation to the base-line data.
: c. Following the occurrence of earthquakes, hurricanes, tornados, or intense local rainfalls, a visual inspection of the embankments and Culvert No. I will be made. If this special inspecticn reveals evidence of change, a survey will be performed to confirm no significant degradation to the base-line data.                -
GRAND GULF-UNIT 1                       3/4 7-45
: 4. (GGNS - 394)

==SUBJECT:==
Technical Specification 3.4.1.4, page 3/4 4-4.
DISCUSSION:     Technical Specification 3.4.1.4 requires that the temperature differential between the reactor vessel steam space coolant and the bottom head drain line coolant should not exceed 100*F when an idle recirculation loop is started.
The temperature of the vessel steam space coolant is obtained from a pressure measurement signal which is fed into the process computer. The computer converts this signal into a temperature value by reference to a standard pressure / temperature saturation curve. This overall method of measurement thus covers the range from operating pressure to ambient conditions with an accuracy of 16*F. It follows that temperatures of steam space coolant below 212*F are not available.
By means of the available temperature indication it is not possible to measure the differential temperature of Technical Specification 3.4.1.4 properly when the reactor is at ambient pressure. A steam space coolant temperature of 212*F will be indicated when in fact the true temperature may be much lower.
Thus, the present Technical Specification places unnecessary restrictions on recirculation loop operation. The proposed change will make this Technical Specification applicable only when the vessel pressure is greater than 25 psig.
JUSTIFICATION: The 100*F differential temperature of Technical Specification 3.4.1.4 was established on the basis of the combined stress due to the full rated operating pressure of 1050 psig plus the thermal stress caused by the 100'F temperature difference. At 25 psig the stress due to pressure is negligible and the thermal stress due to a temperature differential of 100*F is well within the allowables foi nuclear boiler system component materials. At 25 psig in the reactor the pressure instrumentation is on scale enough for the operator to read and also inputs accurately to the computer for conversion to temperature.
SIGNIFICANT HAZARDS CONSIDERATION:
The above proposed change does not cause a significant increase in the probability or consequences of an accident previously evaluated nor does it create the possibility of a new or different from any previously evaluated. These changes do not constitute a significant hazards consideration.
1 G75sp2

4., (GGus - 3 e4)
REACTOR COOLANT SYSTEM IDLE RECIRCULATION LOOP STARTUP                                                                                             .,
LIMITING CONDITION FOR OPERATION                                                                                                                              I
                ~
3.4.1.4 An idle recirculation loop shall not be started unless the temperature differential between the reactor pressure vessel steam space coolant and the bottomheaddrainlinecoolantislessthanorequalto100*Ffand:                                                                                                     I 1
: a.            When both loops have been idle, unless the temperature differential between the reactor coolant within the idle loop to be started up and the coolant in the reactor pressure vessel is less than or equal to 50*F, or
: b.            When only one loop has been idle, unless the temperature differential between the reactor coolant within the idle and operating recirculation loops is less than or equal to 50*F and the operating loop flow rate is less than or equal to 50% of rated loop flow.
APPLICABILITY:                          OPERATIONAL CONDITIONS 1, 2, 3 and 4.
ACTION:
With temperature differences and/or flow rates exceeding the above limits, suspend startup of any idle recirculation loop.
i l
SURVEILLANCE REQUIREMENTS 4.4.1.4 The temperature differentials and flow rate shall be determined to be within the limits within 15 minutes prior to startup of an idle recirculation loop.
Y hloe 25 p Q ska f e.ukyd urat. 8d(ex M M 'es ed
                ' appl lcal>le-GRAND GULF-UNIT 1                                                                                 3/4 4-4
: 5. (CGNS - 321)

==SUBJECT:==
Technical Specification 3/4.3.7.1. Table 3.3.7.1-1, page! 3/4 3-57.
DISCUSSION:     -The addition of item 10.a.3 to the subject table by Amendment 7 did not include the MINIMUM CHANNELS OPERABLE for the Dryer Storage Area radiation monitors. An entry of "1" should be added to this column.
JUSTIFICATION: The proposed change is editorial in nature. The identification of minimum operable channels completes the table.
SIGNIFICANT HAZARDS CONSIDERATION:
The proposed change is editorial and does not cause a significant increase in the probability or consequences of an accident previously evaluated nor does it create the possibility of a new or different accident from any previously evaluated. This change does not constitute a significant hazards consideration.
G74sp3
,. _       . . - .     -  -__  ,       . _ - . . _ . - , . . - __ _ ___._,_ ___ . , - . ~ . _ . _ . _ . . _ _ _     ._ _

l TMLE 3.3.7.1-1 (Continued)
RABIAT10W MOW 110 RING IWSTIENEWfAff0R                           ,
!                                 NfR19WI :4WWlELS     APPLICABLE     AUWWI/TW17                     IEASWR9ENT RANGE          -     MT!WI ih   INSTRIBENTAT!WI                  OPER 4LE    _
CONDITIONS     _SETPOINT_
!T                                                                                                                   -
!E   10. Aree IIenitore M           a. Fuel Mandling Area lw                 Monitors t
10
                                                                                                         -2 to 103 aft /Nr M
!                  1)   IIow Feel       1              (e)            12.5 aft /hr/ IIA j                       Storege Vault
                                                                                                          -2 to 103aft /hr    M
: 2)   Spent Feel      1              (f)            12.5 aft /hr/ IIA                10 Storage Pool
                                                                                                          -2 to 103 sit /hr   M
: 3)   Dryer Storage Area 1           (g)            12.5 aft /hr/II4                 10                                  l
                                                                                                          -2 to 103 sit /hr   M                   -
l t&         b. Centeel Room         1              At all times   ~
                                                                       <0.5 mil /hr/II4                10
!*                 Radiation Monitor Y
: we                      s i                                      in operetten.
i me    when irradiated fuel is being handled in the primary or secondary conteisseent.-
i       f Initial setpoint. Final setpoint to be determined during startop test pmgram. Any ressefred change to j           this setpoint shell be submitted to Caumrission within 90 days after test completion.
4 (a) Trips system trith 2 channels upscale-Mi Hi Hi, or one channel upscale Mt Mi Hi and one channel dsunscate er i           2 channels downscela.
j      (b) Isolates containment /drywell purge penettettens.
(c) With irradiated fuel in spent fuel storage pool.
(d) Alse iselstes the Austillary Building and Fuel Mond1tng Aree Venttistion Systems.
(e) With fuel in the new fuel storece vault.
(f) With fuel in the spent fuel storage pool.
(g) With fuel in the dryer storage area.                                                                                      k*
j      (h) Two upscale Mt M1, one upscale Mi M1 and one duunstale, or too duunstale signals fue the same trip r system actuate the trip system and initiate isolation of the associated Isolation values.
4
                                                                                                                                     =tn 1                                                                                                                                        '
03 l                                                                                                                                      m
:                                                                                                                                      M i
D l                                                                         _ _ _ _ _ _ . _-__ _

9
: 6. (GGNS - 638)

==SUBJECT:==
Technical Specification 3.3.3, Table 3.3.3-3, page 3/4 3-30.
DISCUSSION:                  The low pressure coolant injection (LPCI) mode of the RHR System is noted in the subject table as having a pump response time of 45 seconds for trains A and B. In order to be consistent with the accident analysis assumptions as noted in FSAR Tables 6.3-1 and 6.3-2 the correct response time is 40 seconds for all trains.
The proposed change deletes the distinction between trains A and B and C and utilizes the most limiting response time consistent with the accident analysis.
JUSTIFICATION: The proposed change is somewhat more limiting than the original requirement by its shorter response time. This change brings the technical specifications to a basis consiste'nt with the plant design as represented in the accident analyses.
SIGNIFICANT HAZARDS CONSIDERATION:
The proposed change is more restrictive and is consistent with the existing accident analyses. It does not increase the probability or consequences of an accident previously evaluated nor does it create the possibility of a new or different accident from those previously evaluated. It does not constitute a significant hazards consideration.
G81sp4

          '      ~
6.ccus-sse) l TABLE 3.3.3-3                                                                 !
EMERGENCY CORE COOLING SYSTEM RESPONSE TIMES (SECONDS)                                                                                ;
: 1. LOW PRESSURE CORE SPRAY SYSTEM                                                                                                 i 40
: 2. LOW PRESSURE -COOLANT INJECTION ISDE
                             '0F RHR SYSTEM                 tumes A,B w c.                                                                                  po
: c. P7
* rd !                                                                                                              < 's e 77 C
                                                                                                                                                          ~
L.                                                                                                                        5 ^^ 1
: 3. AUTOMATIC DEPRESSURIZATION SYSTEM                                                                                                     NA
: 4. HIGH PRESSURE CORE SPRAY SYSTEM                                                                                                i 27
: 5. LOSS OF POWER                                                                                                                         NA P
d GRAND GULF-UNIT 1                                                              3/4 3-30

7.-(GGNS - 12)

==SUBJECT:==
Technical Specification 3.3.3, Table 3.3.3-1, page 3/4 3-26.
DISCUSSION:    Delete the current footnote (e). Footnote ** should be revised to indicate that this condition is applicable when the associated ESF equipment is required to be operable.
JUSTIFICATION: Footnote (e) is being deleted because it is not used in Table 3.3.3-1, nor does it clarify the intent or requirement of the Technical Specification. The revision to the ** footnote clarifies the intent of the applicable operational conditions.
SIGNIFICANT HAZARDS CONSIDERATION:
The proposed changes are editorial or administrative in nature.
These changes do not cause a significant increase in the probability or consequences of an accident previously evaluated nor do they create the possibility of a new or different kind of accident from that previously evaluated. These changes do not constitute a significant hazards consideration.
G74sp2

i                                     .
TABLE 3.3.3-1 (Continued)
                               -                        EMERGENCY CORE COOLING SYSTEM ACTUAY10N INSTRtSEWfATION f                                                                                               MINI 9tlM OPERABLE        APPLICABLE
\                                                                                                                         OPERATIONAL j     @                                                                                            CHANNELS PERg                                    ACTION TRIP FUNCTION ,)         CONDITIONS _
l     7,                   .
TRIP FINICTION                                                                                                                                         !
l     g I     U   C.            DIVISION 3 TRIP SYSTEM
: 1. HPCS SYSTEM                                                                  4               1,2,3 '4*,58              33 Reactir Vessel Water Level - Low, Low, Level 2                                      1,2,3                     33
: a.                                                                           4                                         31 l                             b.       Drywell Pressure - High                                             2               1, 2, 3, 4*, 58
: c.       Reactor Wessel Water Level-High, level 8                                           1, 2, 3, 4*, 5*            34 2(d)             1, 2, 3, 4* , 5*          34 l                              d.       Condensate Storage Tank Level-tow                                  2
: e.       Suppression Pool Wate/ Level-High                                  1/ system        1, 2, 3, 4*, 5*           32 i
: f.       Manual Initiation i
D.        LOSS OF POWER
: 1. Division 1 and 2                                                            4               1, 2, 3, 4**, 5**         30 l      t'
* a. 4.16 kV Bus undervoltage (Loss of Voltage)                                                   4               1, 2, 3, 4**, 5**          30 Y                               4.16 kV Bus Undervoltage l
:       M                         b.                                                                                        1, 2, 3,      4**,  5**    30     m (80P Load Shed)                                                   4 l
: c.      4.16 kV Bus Undervoltage j                                             (Degraded Voltage)                                                                                                R i
5**    30       4 Division 3                                                                 4               1, 2, 3,      4**,
2.
4.16 kV Bus Undervoltage                                                                                          .'
a.
(Loss of Voltage) i d D
 ,                 (a)     A channel may be placed in an inoperable status for up to B hou j                          other OPERABLE channel in the same trip system is monitoring that parameter.                                           r j
Also actuates the associated division diesel generator.
]                   (b) Provides signal to close HPCS pump discharge valve only.                                                                                   l
;                   (c)
(d) Provides signal to HPCS pump suction valves only.
i
(;;
* C; nt ;Ptz tir pappMccMe. Applicable when the Aystemual                                            is torequired 135 psig.
to be OPER
                     **     Required when 4SF equipment is required to be OPERA 8LE.                                                                           ~     R Not     required        to be  OPERABLE         when  reactor  steam    done       pressure    is  less    than    or  eq%    n oS
}                    #
that    The. prev: s; oms of St >ce.&.e.:t;oa 4.s.< pre. n.+ sypi:e %. for L. ps  r                                  i R%    .e D,.
p soa.mnee
                            .:n..       iu..es R.an*,r.
re9a;ces ao,su     , open:q pc.y:4.JafL._
v tv.                                . , s3arw:llac..
                                                                                                                        . w ,,, r , u ;<,     . .t.yee formed         e
: 8. (GGNS - 724)

==SUBJECT:==
Technical Specification Tables 3.4.3.2-1 and 3.4.3.2-2, page 3/4 4-10.
DISCUSSION:     The changes to the Technical Specification Tables 3.4.3.2-1 and 3.4.3.2-2 are proposed to clarify the valve numbers for Residual Heat Removal (RHR) system valves to indicate the valves associated with RHR subsystems A, B, and C; to remove from the Reactor Coolant System (RCS) Interface Valves Leakage Pressure Monitor listing the reference to a Reactor Core Isolation Cooling (RCIC) valve leakage pressure monitor; and to provide alarm setpoints based on the system design specifications.
JUSTIFICATION: The RHR injection valves E12-F042A, B, and C and the downstream check valves E12-F041A, B, and C each should be indicated to clearly identify the valves and valve leakage pressure monitors associated with the RHR A, B, and C subsystems. Also, RHR Shutdown Cooling Return to Feedwater Valves E12-F053A and B and the downstream check valves E12-F050A and B each should be indicated to clearly identify the valves associated with RHR A and B subsystems.
The valve leakage pressure monitor on the RHR suction line from the recirculation system senses the pressure in the common line between the E12-F008 valve and both the E12-F006A and E12-F006B valves. Table 3.4.3.2-2 should be changed to indicate both valves E12-F006A and E12-F006B.
I                        The monitor indicated in Table 3.4.3.2-2 to be between RHR and i                        RCIC system valves, E12-F052 and E51-F064 (E51-F064 is listed in the table as F51-F064), is not an interface valve leakage l                        pressure monitor and should not be included in this table. The l
E12-F052 A&B valves are in the parallel lines from the RCIC l                        steam supply line to the RHR A&B heat exchangers. These are the first closed valves in each line from the Reactor Pressure Vessel. The E51-F063 and E51-F064 valves are upstream of the E12-F052 A&B valves and are normally open. As a result, the l
pressure transmitter between the E51-F064 and the E12-F052 valves senses the pressure in the open steam line from the Reactor Pressure Vessel. (This transmitter provides an interlock with the E12-F087A&B valves downstream of the E12-F052A&B valves to prevent the E12-F087A&B valves from opening if the steam line pressure is greater than 480 psig.)
In the parallel lines to the heat exchangers, the E12-F051A&B and E12-F087A&B valves are normally closed valves downstream of the E12-F052A&B valves. There is no RCS interface valve leakage pressure monitor downstream of the E12-F052A&B valves.
The NSSS vendor, General Electric, has stated there is no I
requirement at the RCIC/RHR interface for a leakage monitor.
!                         Thus, the monitor indicated in the table should be deleted because: 1) the pressure transmitter between those two valves l
i C58sp1

senses pressure in the open steam line from the reactor vessel,
: 2) this transmitter provides an interlock to prevent the opening of the downstream isolation valves, 3) this transmitter does not have an alarm function and, 4) per General Electric, a leakage monitor is not required at the RCIC/RHR interface.
The Table 3.4.3.2-2 alarm setpoints for the LPCS (Valves E21-F005 to E21-F006) and RHR (Valves E12-F041A, B, & C to E12-F042 A, B. & C) monitors should be changed from less than or equal to 50 to less than or equal to 70 psig to be consistent with the design specification.
The RHR (Valves E12-F008 to E12-F006 A&B) monitor setpoint should be changed to less than or equal to 195 psig to be consistent with the design specification. The present alarm setpoint incorrectly indicates less than or equal to 135 psig which is a separate interlock setpoint on the E12-F008 valve.
SIGNIFICANT HAZARDS CONSIDERATION:
The proposed amendment to the Technical Specification makes several additions and one deletion to tables 3.4.3.2-1 and 3.4.3.2-2. All changes proposed by the rer.2est are made to achieve consistency between the Technical Specifications and the equipment design specification, make minor corrections, reflect actual plant design or provide clarification without changing the intent. This change to the Technical Specification does not involve a significant reduction in a margin of safety and it does not involve a significant increase in the probability or consequences of an accident which has been previously evaluated, nor does it create the possibility of a new or different kind of accident from any accident previously evaluated. Therefore, the proposed change does not involve any significant hazards considerations.
C58sp2

8, (6 6NS- 72*f)
TABLE 3.4.3.2-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES                                         -c VALVE NUMBER                    SYSTEM E21-F005                        LPCS E21-F006 E22-F004                        HPCS E2z-F005 E12-F008                         RHR E12-F009 E12-F023 E12-F041 A, B.L E12-F042 A ,~6,C                                                                                   '
               - E12-F050 A,8 E12-F053 A,0 E12-F308 E51-F063                        RCIC E51-F064 E51-F065 E51-F066 3                 E51-F076 i               E51-F013 TABLE 3.4.3.2-2 l
                  ,RP"iOR COOLANT SYSTEM INTERFACE VALVES LEAKAGE PRESSURE MONITORS ALARM SETPOINT VALVE NUMBER                     SYSTEM                                             (psic)
E21-F005 to E21-F006            LPCS                                              1 M 70 E12-F008 to E12-F005 A ed       RHR                                               -<3itf (95 t
i eit.Foo6s E12-F041^to E12-F042 A          RHR                                               1 NC to
                 -Els-Fess t: F51-Fest             ::::                                               g=I di2 - FC,4 B +c, El2- FC42B      14R                                               g yo E rl-FC4ic to Ei2- Ec42C         R4R,                                              g gc
(,
GRAND GULF-UNIT 1                    3/4 4-10

!                      9. (GGNS-- 691)

==SUBJECT:==
Technical Specification 3.7.1.1 and 4.7.1.1, pages 3/4 7-1 and 3/4 7-2.
DISCUSSION:               Technical Specification 3.7.1.1 presently requires that two independent Standby Service Water (SSW) subsystems shall be operable under Operational Conditions 1, 2, 3, 4, 5, and *.
Action Statements e, d, and e, however, imply that only one operable SSW subsystem is acceptable under certain circumstances in Operational Conditions 4, 5, and *. These special circumstances are defined in those Technical Specifications which identify the operability requirement for associated ESF equipment in Operational Conditions 4, 5, and *.
t These associated system Technical Specifications include:
3.4.9.1            Residual Heat Removal System (RHR) - Hot Shutdown 3.4.9.2 RHR - Cold Shutdown 3.5.2              ECCS - Shutdown 3.9.11.1 RHR - High Water Level 3.9.11.2 RHR - Low Water Level 3.8.1.2 A.C. Sources - Shutdown i                                                                                                                                      '
ACTION (e) is being revised to reference all of the applicable Operational Conditions given in the associated system Technical Specification 3.8.1.2.
Surveillance Requirement 4.7.1.1 is being revised to better reflect the requirements of the Limiting Condition for Operation.
4 JUSTIFICATION: The proposed changes will make the Limiting Conditions for Operation more consistent with those for the associated ESF l
systems. The changes do not impact the overall intent of l                                                 assuring plant safety and ensure that the requirements for the
'.                                                Standby Service Water System are just as stringent as those for j                                                 the ESF equipment which is serviced by it.
l SIGNIFICANT HAZARDS CONSIDERATION:
l l                                                 The proposed changes reflect the plant design and are l                                                  consistent with the technical specifications for the associated ESF systems. They do not significantly increase the probability or consequences of an accident previously evaluated nor do they create the possibility of a new or different accident from those previously evaluated. They do not constitute a significant hazards consideration.
G81sp2

3/4.7 PLANT SYSTEMS v.(sus 4u) 3/4.7.1 SERVICE WATER SYSTEMS STANDBY SERVICE WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.1 M ndependent standby service water (SSW) system subsystem
                      . -he-49ERADbE with each subsystem comprised of:                                       -l
                              , s. One OPERABLE SSW pump, and
: b. An OPERABLE flow path capable of taking suction from the associated SSW cooling tower basin and transferring the water through the RHR heat exchangers, ECCS pump room seal coolers, and associated coolers andpumpheatexchangersg                       g. g APPLICABILITY:     OPERATIONAL CONDITIONS 1, 2, 3, 4, 5 and *.                       s
:                        ACTION:
: a. In OPERATIONAL CONDITION 1, 2 or 3:
: 1. With one SSW subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 72 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the.following 24 hours.
: 2. With both SSW subsystems inoperable, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN ** within the following 24 hours.
        -                       b. In OPERATIONAL CONDITION 3 or 4 with the SSW subsystem, which is associated with an RHR loop required OPERABLE by Specification 3.4.9.1 e,       '
l          -
or 3.4.9.2, inoperable, declare the associated RHR loop inoperaole and take the ACTION required by Specification 3.4.9.1 or 3.4.9.2, as applicable.
: c. In OPERATIONAL CONDITION 4 or 5 with the SSW subsystem, which is associated with an ECCS pump required OPERABLE by Specification 3.5.2, inoperable, declare the associated ECCS pump inoperable and take the ACTION required by Specification 3.5.2.
: d. In OPERATIONAL CONDITION 5 with the SSW subsystem, which is associated with an RHR system required OPERABLE by Specification 3.9.11.1 or 3.9.11.2, inoperable, declare the associated RHR system l
inoperable and take the ACTION required by Specification 3.9.11.1 I                                    or 3.9.11.2, as applicable.
4fc.r                                            a
: e. In Operational Conditioh 3*, with the SSW subsystem, which is             l associated with a diest i generator     required OPERABLE by Specifica-tion 3.8.1.2, inoperable, declare the associated diesel generator inoperable and take the ACTION required by Specification 3.8.1.2.             .
The provisions of Specificat, ion 3.0.3 are not applicable.
phy er see,,Jary c Ja ..tr, Whenhandlingirradiatedfuelintheg.,,ili;rjC.ildin;;rE:I;:r;Sciidin;.'                 I Whenever both SSW subsystems are inoperable, if unable to attain COLD SHUTDOWN s-                       as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods.
GRAND GULF-UNIT 1                      3/4 7-1 e

i.(GC.N5-[o9I)
                 /Ns6 RT       .6
: 3. i.1.1
             ~
OLcdl  be OPER R3LE             o.s fei\cws :
U             o. . In  OPER AT tom ht Co 6blT)otOS                   1.,2. o d 3 :
lg                           h         su bsy s+tes ; od m                                                                4ad ,6 od * :
' =:isi .          b. In  o?tA ATioM AL tomb)TioMS     ssoda_+
tLe sasys4w                                 witL &
                                                                         & AS % lar s    3. 4. 9.1, h! 3 sks4c RABLE 3.4.92,        ms, by dS cowrpoAuds3.6.
: 3. 8.l. 2 , 3 . 9. 11. 1 m      2. ,
: 3. 9. i t. 2 .
  \
f I
l l                                               s

l
     . .-                                                                                         \
l
  *.                                                                             /
PLANT SYSTEMS kl64NS-b91 6
             $URVEILLANCE REQUIREMENTS                                             i (s)                       a 4.7.1.The     re9uiral 1 4 ash Itandby service water system subsystem shall be demonstrated     [
OPERABLE:     f                                         9
                  - a. ,
AtlEastonceper31daysby:
: 1. Verifying that each valve in the flow path that is not locked, sealed or otherwise secured in position, is in its correct position.
: 2. Verifying that the valves isolating service to the spent fuel storage pool cooler are locked closed.
: b. At least once per 18 months during shutdown by verifying that each automatic valve servicing safety related equipment actuates to its correct position on a actuation test signal.
l d
i GRAND GULF-UNIT 1                    3/4 7-2
: 10. (GGNS - 686)

==SUBJECT:==
Technical Specification Table 3.3.2-3, page 3/4 3-19.
DISCUSSION:    Generic Letter 83-02 (MAEC-83/0034) provided guidelines for some of the technical specification changes required by NUREG-0737, " Clarification of the TMI Action Plan." Item 9 of this letter states, in part:
The pipe break-detection circuitry should be modified so that pressure spikes .asulting from HPCI and RCIC system initiation will not cause inadvertent system isolation.
The plants using a time delay relay for thio modification should change their technical specification to include the time delay added by the relay in the isolation system instrumentation response time. The minimum and maximum expected response time should be provided as discussed in the sample Technical Specifications.
The minimum expected response time is a plant specific value. The maximum expected response time should not be higher than seven seconds unless the licensee provides proper justification for selecting a higher response time.
The proposed modification to Table 3.3.2-2 changes the "###"
                         . note to read:
                                 ### Includes time delay of 3 to 7 seconds.
JUSTIFICATION: Present item Sa, RCIC Steam Line Flow-High, on Table 3.3.2-3 has a response time of less than or equal to 13 seconds. Bases 3/4.3.2 explains that the 13 second time frame is considered for A.C. operated valves and assumes A.C. power loss followed by startup of the emergency diesel generator. A time of 13 seconds is assumed before the valves start to close. The RCIC Steam Line Flow-High timers do not depend on the diesel generators for a power source and thus are timing out during the 13 second time frame assumed above, before the valves start to close.
Since the timers are timed out within the 13 seconds presently in the Technical Specifications, it is appropriate to change the ### note on Table 3.3.2-3 to reflect this action. The i                          present ### note reflects the General Electric design criteria for a maximum setpoint of 13 seconds. The proposed change reflects the requirements of Ceneric Letter 83-02. The minimum
-                          setpoint of 3 seconds reflects instrumentation response time plus delay time to avoid spurious isolation of the system due to pressure spikes resulting from RCIC initiation. The maximum delay time of 7 seconds is in accordance with Generic Letter 83-02 requirements.
G85sp2

SIGNIFICANT HAZARDS CONSIDERATION:
The proposed change to the Technical Specifications is made to comply with the intent of Generic Letter 83-02. The time delay to prevent spurious isolations of RCIC on Steam Line Flow-high is added as a notation (###) to Table 3.3.2-3. This change does not involve the reduction of safety margins and no significant increase in the probability or consequences of an accident previously evaluated is involved nor is the possibility of a new or different kind of accident from any accident previously evaluated created. Thus the proposed change to the Technical Specification does not involve any significant hazards consideration.
G85sp3

INSTRUMENTATION TABLE 3.3.2-3 (Continued)
ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME           .,
TRIP FUNCTION                                                                           RESPONSE TIME (Seconds)#
: 5.       REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION RCIC Steam Line Flow - High                                         < 13 I ") "
a.-
: b.                   RCIC Steam Supply Pressure - Low                                    7 13(*)
: c.                    RCIC Turbine Exhaust Diaphragm Pressure - High                     liA
: d.                    RCIC Equipment Rook Ambient Temperature - High                     NA
: e.                    RCIC Equipment Room A Temp. - High                                 NA
;                            f.                    Main Steam Line Tunnel Ambient Temp. - High                        NA
: g.                    Main Steam Line Tunnel A Temp. - High                              NA
: h.                   Main Steam Line Tunnel Temperature Timer                           NA
: 1.                    RHR Equipment Room Ambient Temperature - High                      NA
: j.                    RHR Equipment Room a Temp. - High                                  NA
: k.                    RHR/RCIC Steam Line Flow - High                                    MA
: 1.                    Manual Initiation                                                  NA
: 6.        RHR SYSTEM ISOLATION 1
: a.                     RHR Equipment Room Ambient Temperature - High                     NA l
: b.                      RHR Equipment Room A Temp. - High                                NA Reactor Vessel Water Level - Low, Level 3                         < 13(,)
t                             c.                                                                              i
                                                                                                                       ~
i                             d.                      Reactor Vessel (RHR Cut-in Pamissive)
Pressure - High                                         NA
' ' (.                                                Drywell Pressure - High                                           NA e.
Manual Initiation                                                 NA f.
(a) The isolation system instrumentation response time shall be measured                                         and Isolation system recorded as a part of the ISOLATION SYSTEM RESPONSE TIME.
instrumentation response time specified includes the delay for diesel generator starting assimed in the accident analysis.,
(b) Radiation detectors are exempt from response time testing. Response time shall be measured from detector output or the input of the first electronic component in the channel.
* Isolation system instrumentation response time for MSIVs only.                                      No diesel l
L                            generator delays assumed.
                           ** Isolation system instrumentation response time for associated valves except MSIVs.
                            # Isolation system instrumentation response time specified for the Trip Function actuating each valve group'shall be added to isolation time shown in Tables 3.6.4-1 and 3.6.5.2-1 for valves in each valve group to obtain ISOLATION SYSTEM RESPONSE TIME for each valve.
fjuitM.;
halvdes                       : ;;;ead   +L tin:     d:?:3
: a. dalay of 3 ss 7 ,,,,Js ,
GRAND GULF-UNIT 1                                                    3/4 3-19
  ' ' ' '        '   *-   w     - - - - - . . , , . ~ , _ _ , . _ _ _       _  ,__   __ _
: 11. (GGNS - 128)

==SUBJECT:==
Technical Specification 3.8.1.2, page 3/4 8-9.
DISCUSSION:    In order to clarify the operability requirements of the subject specification, the use of "and/or" in the current specification is being revised to reflect the Limiting Condition for Operation:
: 1. an offsite circuit,
: 2. either Diesel Generator 11 or 12, and
: 3. Diesel Generator 13.
The following changes are being proposed:
: 1. Change the "and/or" in 3.8.1.2.b to "or".
: 2. Change the "and/or" in ACTION statement a.
line 2 to "and".
: 3. Delete the phrase "of the above required A.C. electrical power sources" from ACTION statement a.
JUSTIFICATION: The proposed change clarifies the specification. The revised limiting condition and action requirement is a more straight forward statement of the intended criteria. The intent of the changes is purely for clarity and in no way alters the LCO or the ACTION Statement.
SIGNIFICANT HAZARDS CONSIDERATION:
The proposed changes are editoral in nature. They do not significantly increase the probability or consequences of an accident previously evaluated nor do they create the possibility of a new or different accident from those previously evaluated. They do not constitute a significant hazards consideration.
G81sp3

Il Gm-ae)
ELECTRICAL POWER SYSTEMS A.C. SOURCES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.1.2 As a minimum, the following A.C. electrical power sources shall be OPERABLE:
: a.      One circuit between the offsite transmission network and the onsite Class lE distribution system, and I                                         b'. Diesel generator 11 end/or 12 and diesel generator 13 when the HPCS                                                                                                      l systemisrequiredtobeOPERdBLE,witheachdieselgeneratorhaving:
: 1. A day tank containing a minimum of 220 gallons of fuel.
: 2. A fuel storage system containing a minimum of:
a)                    48,000 gallons of fuel each for diesel generators 11 and 12.
b)                     39,000 gallons of fuel for diesel generator 13.
: 3. A fuel transfer pump.
APPLICABILITY:                 OPERATIONAL CONDITIONS 4, 5 and *.
ACTION:
: a. With all offsite circuits inoperable and/or with diesel generators                                                                                                    n f % X'
:(                                                11 and/ee-12 ef th: h : 7: :f r:d A.C. :1 :tric:1 p.;;r-settecer l i                              suspend CORE ALTSRATIONS, handling of irradiated fuel                                                                                    18 PM*V " derable,1i:[- Sildirs; :nd En:10::r:      tx;                                                                                                              ild;;;,            operations with a jk   d potential for raining the reactor vessel and crane operations over                                                                                                      g$
meu d'l cd.we"1                          the spent fuel storage pool when fuel assemblies are stored therein.
In addition, when in OPERATIONAL CONDITION 5 with the water level less than 23 feet above the reactor pressure vessel flange, immediately initiate corrective action to restore the required power sources to 0/ERABLE status as soon as practical.
:                                            b. With diesel generator 13 of the above required A.C. electrical power i
sources inoperable, restore the inoperable diesel generator 13 to OPERABLE statos within 72 hours or declare the HPCS system inoperable and take the ACTION required by Specification 3.5.2 and 3.5.3.
: c. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.8.1.2 At least the above required A.C. electrical power sources shall be demonstrated OPERABLE per Surveillance Requirements 4.8.1.1.1, 4.8.1.1.2 and 4.8.1.1.3, except for the requirement of 4.8.1.1.2.a.5.
pcWo.cy or secoud q toda' N"*"Y' po
                               ,                                                                                                                                                                                          8 E When handling irradiated fuel in theAxili:ry Silding cr En:10::r: kilding. N                                                                                                                 T:
3/4 8-9 b.
GRANC GULF-UNIT 1 l
* j
    . . .                                                                                                                     [
l
: 12. (GGNS - 729)

==SUBJECT:==
Technical Specification Basis 3/4.6.7 page B3/4 6-6.
DISCUSSION:       The proposed change updates the effective revision date of Regulatory Guide 1.7. The applicable revision according to the commitments made in the FSAR is Revision 1, dated September, 1976.
JUSTIFICATION: The proposed change is administrative in nature to reflect the design commitments made in Appendix 3A of the GGNS FSAR.
The September, 1976, revision replaces Safety Guide 7, dated March, 1971, and its' Supplement dated October, 1971.
SIGNIFICANT HAZARDS CONSIDERATION:
The proposed change is administrative in nature. It does not significantly increase the possibility or consequences of an accident previously identified nor does it create the possibility of a new or different accident from any previously evaluated. It    does not constitute a significant hazards consideration.
[
l l
l l
l l
l l            G77sp1

           . *                                               ;                    12,(fe6MS-7 2N CONTAINMENT SYSTEMS BASES
             'f  3/4.6.6 SECONDARY CONTAINMENT Secondary containment is designed to minimize any ground level release of
                - radioactive material which may result from an accident. The Auxiliary Building and Enclosure 8411 ding provide secondary containment during normal operation when the containment is sealed and in service. When the reactor is in COLD j                SHUTDOWN or REFUELING, the containment may be open and the Auxiliary Building and Enclosure Building then become the only containment.
 ;                       Establishing and maintaining a vacuum in the Auxiliary Building and Enclo-sure Building with the standby gas treatment system once per 18 months, along
,                with the surveillance of the doors, latches, dampers and valves, is adequate to ensure that there are no violations of the integrity of the secondary containment.
The OPERABILITY of the standby gas treatment systems ensures that suffi-cient iodine removal capability will be available in the event of a LOCA. The reduction in containment iodine inventory reduces the resulting site boundary radiation doses associated with containment leakage. The operation of this i                system and resultant iodine removal capacit.y are consistent with the assumptions used in the LOCA analyses. Cumulative operation of the system with the heaters OPERABLE for 10 hours over a 31 day period is sufficient to reduce the buildup of moisture on the absorbers and HEPA filters.
        -:        3/4.6.7 ATMOSPHERE CONTROJ The OPERABILITY of the systems required for the detection and control of hydrogen gas ensures that these systems will be available to maintain the hydro-gen concentration within the containment below its flammable limit during post-4 LOCA conditions. The hydrogen recombiner and the hydrogen ignition systems are capable of controlling the expected hydrogen generation associated with (1) zirconium-water reactions, (2) radiolytic decomposition of water and (3) corrosion of metals within containment.
Two 100% drywell purge systems are the primary means of H2 control within the drywell purging hydrogen produced following a LOCA into the containment volume. Hydrogen generated from the metal-water reaction and radiolysis is assumed to evolve to the drywell atmosphere and form a homogenous mixture through natural forces and mechanical turbulence (ECCS pipe break flow). The drywell purge system forces drywell atmosphere through the horizontal vents and into the containment and as a result no bypass path exists.
                       . The hydrogen control system is consistent with the recommendations of Regulatory Guide 1.7, " Control of Combustible Gas Concentrations in Containment Following a LOCA",4eceH90A hpie.mbe.r ,19%.                                       l
                        -The operability of at least 41 of 45 ignitors in either hydrogen ignition subsystem will maintain an effective c, overage throughout the containment and drywell. Each subsystem of ignitors will initiate combustion of any sig-nificant amount of hydrogen released after a degraded core accident. This system will ensure burning in a controlled manner as the hydrogen is released instead of allowing it to be ignited at high concentrations by a' random igni-        ,
tion source.                                                     -
GRAND GULF-UNIT 1                    B 3/4 6-6 1

L          .
13.1(GCNSi- 784)

==SUBJECT:==
Source Range Monitors, detector not full in rod. block in Technical-Specification Tables 3.3.6-1 and 4.3.6-1, pages 3/4.3-50, 3/4.3-51, 3/4 3-53, and 3/4 3-54.
DISCUSSION:-     When entering' Operational Condition 5, Technical Specification 3/4.3.6 requires that-the Source Range Monitor " detector not full in" interlock be operable (see Table 3.3.6-1) . At this
                                   ' time, the detectors are full in, but the. drive motor module (including the interlocks) may be removed for maintenance which is acceptable in Operational Condition 4. The safety function of-the interlock is to generate a rod block when the detectors are not full in. This safety function is satisfied even with the interlock hardware removed since the rod block is also generated by the absence of the interlock hardware.
Operability of the interlock is required prior to entry into Operational Condition'5 in order to meet Specification 3.0.4.
The proposed ch'ange to.the Technical Specifications consists of adding a note (e) to Table 3.3.6-1 to state that the provisions of Technical Specification 3.0.4 are not applicable for entering Operational Condition 5.. A note (**) is added to the surveillance. requirements in Table 4.3.6.1 to require verification that the-SRM detectors are " full in" at least once
                                                   ~
per 24 hours when a detector drive motor module is removed.
JUSTIFICATION: Operability of the Source Range Monitor " detector not full in" rod block is satisfied even with the drive motor module removed. The presence of the rod block allows entJy into Operational Condition 5 and the-requirement of Specification 3.0.4 to be met. A letter (AECM-83/0015) from Mr. L. F. Dale to Mr. J. P. O'Reilly, Regional Administrator, dated January 10, 1983, requested NRC endorsement of this interpretation of the Technical Specifications. This' endorsement for only a one-time interpretation was received in a letter (MAEC-83/0016) from Mr. R. C. Lewis, Director Division of Project and Resident Programs, to Mr. N. L. Stampley dated January 13, 1983. The.
NRC considered-this a one-time interpretation which would be
                                   -corrected by a permanent change to the Technical
                                   ~ Specifications.
SIGNIFICANT HAZARDS CONSIDERATION:
This proposed change provides operational flexibility while retaining the intent of the Action Statement for inoperability
                                   -of the Source Range Monitor " detector not full in" interlock since the rod block is satisfied regardless of the presence of the drive motor module.. Through the design feature which establishes a rod block when a drive motor module is removed and the additional requirement of verifying SRM detectors are
                                     " full'in" every 24 hours when a detector drive motor module is removed, no reduction in
                     =
1FTSHide2'

safety margin is established. This change does t.ot significantly increase the probability or consequences of an accident previously evaluated nor does it create the possibility of a new or different accident from those previously evaluated. It does not constitute a significant hazard consideration.
r I
I l
l i
l l
FTSHide2.1

                                                                                                                                     ~
I                                                                 TA8tE 3.3.6-1 ss                                                                                                                     *                      .
                                         .               CONTROL ROD BLOCK INSTR M NTATION                          ,
O MINIMUN              APPLICABLE             .
OPERABtE CilANNELS         OPERATIONAL                        ,
.G.                                                                                        CONDITIONS      ACTION                -
PER TRIP FUNCTION i             TRIP FUNCTION 5                                                                                                                  .
!             1. R00 PATTEllN CONTROL SYSTEM                                                               60 Low Powstr Setpoint                             2                    1, 2 4                  a.
j                  b.       Intermediate Rod Withdrawal                                                      60 Limiter Setpoint                           2                    1, 2                                               .
: 2. APRM
: a.       Flow 81ased Neutron Flux-                                                        61 6                    1 Upscale                       .
61
: b.       Inoperative                                     6                    1,2,5 6                    1           61
: c.       Downscale 2, 5        61
: d.       Neutron Flux - Upscale, Startup                 6
!R   s                                                 s                                                                                '
!Y             3. SOURCE RANGE MONITORS                                                                     61 l
2, 5                                             .
E          . a.       Detector not full in(* 1                        4 4                    2, 5        61 Upscale (b)
)                  b.
2, 5        61
: c.        Inoperative (b)                                4 l                                                                                                            61 4                    2, 5
: d.      Downscale(C) 1 l              4. INTERNEDIATE RANGE MONITORS 6                   2,  5       61 l                   a.       Detector not full in (d)                        6                   2,  5       61                              %
l                   b.       Upscale 6                   2,  5       61                              .W '
i                   c.        InoperatigI                                                                     61 A
6                   2,  5                      F                %
i                    d.      Downscale i                                                                                                                                            O I               5. SCRAM OISCHARGE VOLUNE
: a. Mater Level-High                               2                    1, 2, 5*    62                             g'~
i               6. REACTOR C00LANT SYSTEM RECIRCULATION FLOW 1           62 Upscale                                        3 a.
U+

l3.(&frN.5- ?t*t)
INSTRUMENTATION TABLE 3.3.6-1 (Continued)
CONTROL ROD RLOCK INSTRUMENTATION f
ACTION ACTION 60                            -                 Declare the RPCS inoperable and take the ACTION required by Specification 3.1.4.2.
ACTION 61                            -
With the number of OPERABLE Channels:
: a. One less than required by the Minimum 0PERABLE Channels per Trip Function requirement, restore the inoperable channel to OPERABLE status within 7 days or place the inoperable channel in the tripped condition within the next hour,
: b. Two or more less than required by the Minimum OPERABLE Channels per Trip Function requirement, place at least one inoperable channel in the tripped condition within one hour.
ACTION 62                              -                With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place the inoperable channel in the tripped condition within one hour.
t-NOTES
* With more than one control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
ia) ' This function shall be automatically bypassed if detector count rate is
                                     > 100 cps or the IRM channels are on range 3 or higher.
'                           (b) This function shall be automatically bypassed when the associated IRM channels are on range 8 or higher.
i (c) This function shall be automatically bypassed when the IRM channels are on range 3 or higher.
(d) This function shall be automatically bypassed when the IRM channels are l
l on range 1.
(L') % prov;.sion.s of S y,w% tion 3. 0. '+ are m T~
I ayyli ca W % a.mte.rs y opraional Con) Mon E.
l l
GRAND GULF-UNIT 1                                                                    3/4 3-51
     . . ~ _ _ _ . _ _ _          _ . . _ _ _ . - . , - _ _ _ _ _ . _ _ . . _ ,                    . . _ . _ _ . . _ _ . . . . _ _ . . _ . _ , . . . _ , _ . . . _ . . - . _ . . - . _ _ . _ _ -                      _ , . _ .

TABLE 4.3.5-1                                   .
CONTR)t R00 SLOCK INSTRINENTATION SURVEILLANCE REQUIRE 9ENTS CMf01EL            ,                OPEMTI0llAL
   @                                                              FUNCTIONAL         CM191EL         COMITIONS FOR WilCH G;                                               CHANNEL TEST          CALIRRATION  I ,I SURVEILLANCE REQUIRED 1 RIP FUNCTIOlt                               CHECK
   !   1. ROD PATTFM CONTIl0L SYSTEM
: a. Low Power S tpoint                  NA     g)I    'I,CICII'I,        Q                  Ic 2
[
b)(e) ,DICII'I,                          1, 2
: b. Intermediate Rod Withdrawal         NA    Sg(II'I                    Q Lloiter Setpoint                        M                                                                            ,
: 2. APflM
!            a. Flow Blased Neutron Flux-                          .5/U(b) y          y(f)(g), 34       3
!                    Upscale                     . IIA                    ,
!           b. . Inoperative                          NA              S/U    ,W         g)
W    , SA 1, 2, 5 1
w        c. Downscale                           M               S/U(b),W l                 Neutron Flum - Upscale, Startup      NA             S/U    .N          Q                2, 5 l  )         d.
    *f  3. SOUllCE RANGE MONITORS l
5                                                  M              S/U$,W          NA                  2, 5 N           l l            a. Detector not full in                                         W                         2, 5
: b. Upscale                             NA                                Q
!                                                                     S/U(b),W                             2, 5 i            c. Inoperative                         M              S/U(b),W NA S/U                Q                 2, 5
: d. Downscale                           NA                     ,
l       4. INTEllMEDIATE RANGE MONITORS M                    2, 5 f'
l a.
b.
c.
Detector not full in Upscale Inoperative MA NA M
S/U S/U(b),W S/U(b),W
                                                                              ,W W    Q NA 2, 5 2, 5                          l l                                                      NA             S/U      ,      Q                    2, 5                          !
: d. Downscale                                                                                                      "
:                                                                                                                                 y
: 5. SCRAM SISCMilGE VOLWIE i
Water Level-Nigh                     NA               M               R                 1, 2, 5* "             h; I             a.                                                                                                                  O i        6. REACTOR COOLANT SYSTEM RECIRCULAT10ll FLOW                                                                         O I                                                       NA             S/U(b) M           q                1                       g     .
j              a. Upscale                                                      .
M
!                                                                                                                                    I i
1                                                                                                                                 M e
i i
h*
: u. (scas -,s4)
INSTRUMENTATION TABLE 4.3.6-1 (Continued)
CONTROL R0D BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS NOTES:
: a. Neutron detectors may be excluded from CHANNEL CALIBRATION.
: b. Within 24 hours prior to startup, if not performed within the previous 7 days.                                                                                                                                                    *                                       \
j
: c. Within one hour prior to control rod movement, unless performed within                                                                                                                              '
the previous 24 hours, and as each power range above the RPCS Tow power setpoint is entered for the first time during any 24 hour period during
-             power increase or decrease.
: d. At least once per 31 days while operation continues within a given power range above the RPCS low power setpoint.
: e. Includes reactor manual control sultiplexing system input.
f.
This calibration shall consist of the adjustment of the APRM channel to conform to the power values calculated by a he t balance during.-
OPERATIONAL CONDITION 1 when THERMAL POWER is greater than or equal to 2 M of RATED THERMAL POWER.
Adjust the APRM channel if the absoluta Any APRM Channel difference is greater than 25 of RATED THERMAL POWER.
gain adjustment made in compliance with Specification 3.2.2 shall not be included in determining the absolute difference.
: g. This calibration shall consist of the adjustment of the APRM flow biased channel to confore to a calibrated flow signal.
: h. This calibration shall consist of verifying the trip setpoint only.
* With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
         "                 TMs S R M ocrearm o mve .yo.ra. nooom NITH RENoVECh VERIFY THc   S R M ocrecron.s ^ ^e 2 4- H OU R S *
                   #FutL IN ' AT L EAST                                                                  ONc6 P6A GRAND GULF-UNIT 1 3/4 3-54
                                                                                                                                                                     ~
                               - - - . . . - - , - - - - - , . - - , . - _ . _ , , . - - . . , , - .                        ,--.--,----,,,,,,,,.m,-,--,w,-.,..ww,r-,
                                                                                                                                                                                  ,ww-wy- w---c-e--,--=--i.wvm

i
* i-   , .
: 14. (GGNS - X14)

==SUBJECT:==
Technical Specification Basis 3/4.6.1.7, page B 3/4 6-2.
DISCUSSION:     The proposed change corrects the range limits for the initial containment-to-Auxiliary and Enclosure Building differential pressure. The current limit given in the Basis is (-)2.0 to 0.0 psid. This range was apparently copied from the BWR
* Standard Technical Specification prior to verifying the plant specific data.
The correct range for GGNS is (-)0.1 to 1.0 psid. The correct range is used in the Technical Specification 3.6.1.7.                                                       l JUSTIFICATION: The above correction brings the basis for the technical specification into agreement with the specification itself and with plant design.
SIGNIFICANT HAZARDS CONSIDERATION:
The proposed change does not significantly increase the probability or consequences of an accident previously evaluated nor does it create the possibility of a new or different accident from those previously evaluated.                                       It does not constitute a significant hazards consideration.
i l
l G88sp1

I.4. (6 SNS - XI9)
CONTAINMENT SYSTEMS BASES                                                                                                                                                                        1 3/4.6.1.5 FEE 0 WATER LEAKAGE CONTROL SYSTEM                                                                                                                                 l l
The feedwater leakage control system consists of two independent subsystems designed to eliminate through-line leakage in the feedwater piping by pres-curizing the feedwater lines to a higher pressure than the containment and drywell pressure. This ensures that no release of radioactivity through the feedwater line isolation valves will occur following a loss of all offsite power coincident with the postulated design basis loss-of-coolant accident.
3/4.6.1.6 CONTAINMENT STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment will be maintained comparable to the original design standards for the life of the unit. Structural integrity is required to ensure that the containment                                                                         will A visual c:ithstand the maximum pressure of 11.5 psig in the event of a LOCA.
inspection in conjunction with Type A leakage tests is sufficient to demonstrate this capability.
l 3/4.6.1.7 CONTAINMENT INTERNAL PRESSURE The limitations on containment-to-Auxiliary Building and Enclosure Building differential pressure ensure that the containment peak pressure of 11.5 psig l                   does not exceed the design pressure of 15.0 psig during LOC $ conditions or that the external pressure differentidl does not exceed the design maximum external pressure differential of 3.0 psid. The limit of 2.0 6 0.C psid for initial                                                                            l containment-to-Auxiliary Building and Enclosure Building differential pressure will limit the containment pressure to 11.5 psid which is less than the design pressure and is consistent with the safety analysis.
3/4.6.1.8 CONTAINMENT AVERAGE AIR TEMPERATURE The limitation on containment average air temperature ensures that the containment peak air temperature does not exceed the design temperature of 185*F during LOCA conditions and is consistent with the safety analysis.
3/4.6.1.9 CONTAINMENT PURGE SYSTEM The continuous use of the containment purge lines during all operational conditions is restricted to the 6-inch purge supply and exhaust isolation valves; whereas, continuous containment purge using the 20-inch purge system is limited to only OPERATIONAL CONDITIONS 4 and 5. Intemittent use of the 20-inch purge system during OPERATIONAL CONDITIONS 1, 2 and 3 is allowed only to reduce airborne c''.ivity levels and shall not exceed 1000 hours of use per 365 days.                                                                      .
The design of the 6-inch purge supply and exhaust isolation valves meets the requirements of Branch Technical Position CSB 6-4, " Containment Purging During Normal Plant Operations."
GRAND GULI-UNIT 1                                          8 3/4 6-2 w    -n, - . -       ,e e  4   -n,,e--,,    -,-,,m- w .-- -,-- - e,, - - , - -  -,----_,,------n   ,----,---,,-,-,,,,,,,,-,,--,-e,--,-,---                 - , -        - - - - , , , - - -
: 15. (GGNS - 670)

==SUBJECT:==
Technical Specification 4.5.3.1, 3.6.3.1, and Bases 3/4.6.3, pages 3/4 5-9, 3/4 6-20, and B 3/4 6-4.
DISCUSSION:    The suppression pool is required to be OPERABLE as part of the ECCS to ensure that a sufficient supply of water is available to the HPCS, LPCS and LPCI systems in the event of a LOCA.
The suppression pool water volume must also absorb the associated decay and structural sensible heat released during a reactor blowdown from 1089 psig. Surveillance Requirement 4.5.3.1.a.2 presently requires the suppression pool water level to be verified greater than or equal to 12'5" at least once per 12 hours in OPERATIONAL CONDITION 4 or 5.         The proposed change corrects the present 12'5" to 12'8" to agree with Specifications 4.5.3.1.b. 3.5.3.b and ACTION Statement 3.5.3.c and d. Specification 3.6.3.1.a presently lists the suppression pool high water level as 18'30" corresponding to a volume of 138,851 ft8       The proposed change corrects the volume to 138,701 ft8 to agree with the FSAR. The presently stated 8 maximum and minimum suppression pool volumes of 138,851 ft and 136,146 ft 8in Bases 3/4.6.3 would also be changed to 138,701 ft8 and 135,291 ft 8 tu agree with the FSAR.
JUSTIFICATION: The present 12'5" listed in Surveillance Requirement 4.5.3.1.a.2 is a " typo" and should be corrected to read 12'8" as in Specifications 4.5.3.1.b. 3.5.3.b and ACTION Statements 3.5.3.c and d. The change to Specification 3.6.3.1.a corrects the present suppression pool high water volume of 138,851 ft8 to 138,701 ft*. These changes are made to agree with the values listed in FSAR Tables 6.2-1 and 6.2-50. The changes to the maximum and minimum supprecsion pool volume presently listed in Bases 3/4.6.3 as 138,851 ft       8 and 136,146 ft8 to 138,701 ft8 and 135,291 ft8, respectively, are made to agree with the values as listed in FSAR Tables 6.2-1 and 6.2-50.
SIGNIFICANT HAZARDS CONSIDERATION:
The following changes are considered purely administrative to promote consistency between the FSAR and the Technical Specifications or to correct typographical errors.
l
: 1. Suppression pool low water level in OPERATIONAL CONDITION 4 or 5 change from 12'5" to 12'8".
l 8
l                           2. Presentsuppressiongoolhighwatervolumeof 138,851 ft l
change to 138,701 ft
: 3. Change in Bases from present maximum and minimum suppression pool volumes of 138,8518ft and 136,146 8ft to l
138,701 ft8 and 135,291 ft8, respectively.
l l            C58sp3

I i
These changes do not involve the reduction of safety margins and no significant increase in the probability or consequences of an accident previously evaluated is involved nor is the possibility of a new or different kind of accident from any accident previously evaluated created. Thus the proposed change to the Technical Specification does not involve any significant hazards considerations.
c58sp4
                        . _ . . . . . _ _    . _ _     __. _._ _ . _ . _ _ _ _ . _ _ , . .._.-._. ~__. __
: 15. (GG NS-(gTC)
EERGENCY CORE COOLING SYSTEMS 1.
LIMITING CO W ITION FOR OPE UTION (Continued)
M: (Continued)
: c.      With one suppression pool water level instrumentation division
                   ' inoperable, restore the inoperable division to OPERABLE status within 7 days or verify the suppression pool water level to be ater than or equal to 1884-3/4" or 12'8", as applicable, at reast once per 12 hours by an alternate indicator.
: d.      With both suppression pool water level instroentation divisions
:                     inoperable, restore at least one inoperable division to OPERABLE                  1 status within 8 hours or be in at least WT SWTDOWN within the next 12 hours and in COLD $NUTDOWN within the following 24 hours and verify the suppression pool water level to be greater than er equal
                   . to 18'4-3/4" or 12'8", as applicable, at least once per 12 hours by at least one alternate indicator.                                                 .
SL'WEILLANCE REQUIREMENTS 4.5.3.1        The suppression pool shall be detemined OPERABLE by verifying:
: a.      The water level to be greater than er equal to, as applicable:
: 1. 38'4-3/4" at least once per 24 hours.                                       i
: 2. 4825' at least once per 12 hours.                                      l 12'8"
: 6.       Two sgpression pool water level instroentation divisions, with I channel per division OPERABLE with the low water level alarm setpoint 138'51a" or 22'8", as applicable, by perfomance of a:
: 1. CHANNEL CHECK at least once per 24 hours,
: 2. CHANNEL FUNCTIONAL TEST at least once per 31 days, and
: 3. CHANNEL CAlltRATION at least once per 38 months.
4.5.3.2 With the suppression pool level less than the above limit or drained in OPERATIONAL CONDITION 4 er 5*, at least once per 22 hours:
I              a.      Verify the required conditions of Specification 3.5.3.b to be
                  -    satisfied, or
: b.      Verify footnote condttions
* to be satisfied.
GRAND SULF-L811T 1                       3/4 5-9

15.(GG N5- M @
CONTA! MENT SYSTEMS 3/4.6.3 DEPRESSURIZATION SYSTEMS
                             #                                                     i SUPPRESSION POOL LIMITING CONDITION FOR OPERATION 3.6.3.1-    The suppression pool shall be OgAg with the pool water:
: a. Volume between 135,291 fts and 468i461 fts, equivalent to a level between 18'4-3/4" and 18'10", and a
: b. Maximum everage temperature of 95'F during OPERATIONAL CONDITION 1 or 2, except that the maximum average temperature may be permitted l
I                     to increase to:
: 1. 105'F during testing which adds heat to the suppression pool.
: 2. 110*F with THERMAL POWER less than or equal to IX of RATED          !
THERMAL POWER.
: 3. 120*F with the main steam line isolation valves closed following a scram.
APPLICA8ILITY: OPERATIONAL CONDITIONS 1, 2 and 3.
I ACTION:
: a. With the suppression pool water level outr.ide the above limits, restore the water level to within the limits within 1 hour or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
: b. In OPERATIONAL CONDITION 1 or 2 with the suopression pool average water temperature greater than 95'F, restore the average tesaperature to less than or equal to 95'F within 24 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours, except, as pemitted above:
: 1. With the suppression pool average water temperature greater than 105'F during testing which adds neat to the suppression pool, stop all testing which adds heat to the suppression pool and restore the average temperature to less than 95'F within 24 hours or be in at least HOT $HUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
: 2. With the suppression pool average water temperature grea,ter than 110*F, place the reactor mode switch in the Shutdown position and operate at least one residual heat removal loop in the suppression pool cooling mode.                              *
: 3. With the suppression pool average water temperature greater than 120*F, depressurize'the reactor pressure vessel to less
                    -          than 200 psig within 12 hours.
             #See Specification 3.5.3 for ECCS requirements.
GRAND GULF-UNIT 1                        3/4 6-20

       ..                      =   . _ .                  __ - -.. -_          - .- .           _ _ _
: 15. (GG N5-MO)
CONTAIMENT SYSTEMS
_ BASES t
3/4.6.2.6 DRYWELL AVERAGE AIR TEMPERATURE The limitation on drywell average air temperature ensures that peak drywell temperature does not exceed the design temperature of 330*F during LOCA conditions and is consistent with the safety analysis.
3/4.6.3 DEPRESSURI'ZATION SYSTEMS The specifications of this section ensure that the drywell and containment pressure will not exceed the design pressure of 30 psig and 15 psig, respectively, during primary system blowdown from full operating pressure.
l The suppression pool water volume must absorb the associated decay and q   structural sensible heat released during a reactor blowdown from 1089 psig.
op Using conservative parameter inputs the maximum calculated containment g pressure during and following              a design basis accident is below the containment m design pressure of 15 psig. Similarly the drywell pressure remains below the
:  N design pressure of 30 g ig. The maximum and minimum water volumes for the
     ~5uppression pool are 1.~,41 cubic feet and4%,10 cubic feet, respectively.
These values include the water volume of the gcontainment pool, horizontal l
vents and weir annulus. Testing in the MarkQLI Pressure Suppression Test l
Faci 1Ityandanal.tsishaveassuredthatthesuppressiomoooltemperaturewill not rise above 185'F for the full range of break sizes. ( g g y cy Should it be necessary to make the suppression pool inoperable, this shall only be done as specified in Specification 3.5.3.
Experimental data indicates that effective steam condensation without                           i excessive loads on the containment pool walls will occur with a quencher device and pool temperature below 200*F during relief valve operation.                         ,
Specifications have been placed on the envelope of reactor operating condi-tions to assure the bulk pool temperature does not rise above 185'F in compliance with the containment structural design criteria.
In addition to the limits on temperature of tha su)pression pool waterrelief operating procedures define the action to be taken in tie event a safetyl valve inadvertently opens or sticks open. As a minimum this action shal include: (1) use of all available means to close the valve, (2) initiate su)pressionpoolwatercooling,(3)initiatereactorshutdown,and(4)if otter safety-relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open safety relief valve to assure mixing and uniformity of energy insertion to the pool.
The containment spray system consists of two 100% capacity trains, .each with three spray rings located at different elevations about the inside RHR A pump sup)1ies one train and RHR pump circumference of the containment.
8 supplies the other.         RHR pump C cannot suppl tse containment spray system.
350 nozzles in each train,
!          Dispersion of the flow of water is effected b he containment volume and enhancing the condensation of water vapor inHeatrejectionisthroughtheRHRhe preventing overpressurization.The turbulence caused by the spray system aids in exchangers.
containment hir volume to maintain a homogeneous mixture for H                2 control.
The suppression pool cooling function is a mode of the               TheRHR   system purpose     of theand functions as part of the containment heat removal system.
system is to eni,ure containment integrity following a LOCA by preventing GRAND GULF-UNIT 1 8 3/4 6-4 l
:           '16. (GCNS - 876) l            

==SUBJECT:==
Technical Specification 3.3.2. Table 3.3.2-2, page 3/4 3-15.
DISCUSSION:      The trip setpoints for the:
: a. Containment and Drywell Ventilation Exhaust Radiation Monitor,
: b. Fuel Handling Area Ventilation Exhaust Radiation Monitor, and 1
: c. Fuel Handling Area Pool Sweep Exhaust Radiation Monitor are being revised. Technical Sper.fication Table 3.3.7.1-1 currently gives an upper limit foi the high alarm and the high-high trip setpoints for these monitors. While this table identifies these values as the setpoints, the body of the specification intends them to be maximum values for the                               ,
setpoints. The values in this table for the trip setpoint have been interpreted in Table 3.3.2-2 to be the allowable value.
l The actual high-high trip setpoint is incorrectly stated in                           -
Table 3.3.2-2. The current values happen to be the same as the maximum values for the alarm setpoints given in Table 3.3.7.1-1. However, the alarm and trip functions cre separate.
The trip setpoints in Table 3.3.2-2 should read:
: a. less.than or equal to 3.6 mR/hr
: b. less than or equal to 3.6 mR/hr
: c. less than or equal to 30 mR/hr , respectively.
JUSTIFICATION: The proposed change corrects the trip setpoints for these radiation monitors. The revised values are within the allowable technical specification Ifmits, but are now higher than the high radiation alarm allowable value. The intended separation between the alarm and trip function are thus provided by the change.
SIGNIFICANT HAZARDS CONSIDERATION:
The proposed changes are editorial or administrative in nature.
These changes do not cause a significant increase in the probability or consequences of an accident previously evaluated 1
nor do they create the possibility of a new or different kind of accident from that previously evaluated. These changes do not constitute a significant hazards consideration.
i G77sp3
    .-    .            -,-            . . . - - - . . -    . . - . . - = . .    . . - _ , - - _ . - - - . - - . _ .

i                                                                     *                  .
i
* T MLE 3.3.2-2                                                    ,
j ISOLAff0R ACTURTftNI IftSTRUMENTAff0N SET MfHf5   .
ALLOWWLE TRfP SETP0tWF                          VALUF _
Trip FWICT191 k
jg    1. PRIMMY CONTAlfGEltf ISOLATI0ll
: a. Reacter Vessel Water Level -
Q                                                               1 -41.8 fnehes
* 1 -43.8 inches w                 Low Low. Level 2 i 1.73 psfg                  $ 1~.93 psIg
: b. Dryuell Pressure - High l          c. Contaivupnt and Dryuell Venttistion
                                                                                                $ 4.0 mr/hr**
l                   Exhaust Radiation - High High                 $ J " r,L'*
g l
Itanuel Inttistion                                M 3.(o m R
!           d.
l      2. IWLIN STEM LINE ISOLATI0ll
: a. Ileacter Vessel Water Level -                                                  1 -152.5 inches Lew Low Low, Level 1                          1 -150.3 inches *
                                                                    < 3.9 x fu11 power           < 3.5 a full peuer g*             Ilmin Steam Line Radiation - Nigh                                              ~ background
: b.                                                          background Y                 s 1 837 pois E              IIsin Steam Line pressure - Low                  1 M9 pole
: c.                                                                                  5 176.5 psId Main Steen Line Flow - High                      3 169 psfd d.
19 inches Hg. Vacuum         1 8.7 inches Mr. Vocaus         '
: e. Condenser Vacuum - Low Main Steam Line Tunnel Teeperature - Nigh       $185'F**                     3 191*F**
l f.
: g. Main Steau Line Tunnel & Temp. - Nigh            $ 101*F**                   5 104*F**           A O
M                   A l                   Manuel Initiation IIR
:             h.
SECOWAltY tolffAffgENT 150LAffgIl g
: 3.                                                                                                             '
!              a. Reactor Vessel Water Level -
1 -41.6 inches
* 1 -43.8 inches       j l
Low Low, Level 2 Dryuell pressure - Nigh                          i 1.73 psig                i1.93 pots          O j              b.
Feel Handling Aree Ventilation                                                           ee                 i l              c.                                                     $ & + aR/hr**               $ 4.0 sR/hr Exheast Radition - High High                      3#
l
!              d. Feel Handling Aree Peel Suesp                                                1 35 mR/hr**
Exhaust Radiation - Mfgh High                 148 uR/hr**
l                                                                      M D                        M f,             e. Olennel InitIatien
: 17. (GGNS - X29)

==SUBJECT:==
Technical Specifications 6.9.1.5 and 6.9.1.12, pages 6-16 and 6-20.
DISCUSSION:    Item II.K.3.3 of NUREG-0660, the TMI Task Action Plan, required that safety / relief valve failures be reported promptly and challenges be reported annually. GGNS committed to meeting these requirements in FSAR Section 18.1.30.1.
The proposed changes incorporate these commitments into the technical specifications.
JUSTIFICATION: The proposed changes add reporting requirements to the technical specifications. These reporting requirements reflect MF&L commitments.
The format of the changes conforms to the guidance provided in Generic Letter 83-02 (MAEC-83/0034).
SIGNIFICANT HAZARDS CONSIDERATION:
The proposed changes are administrative in nature. They do not cause a significant increase in the probability or consequences of an accident previously evaluated nor do they create the possibility of a new or different accident from those previously evaluated.
G77sp2

     ' ~
  ..                                                                                     17. (GG NS- G.D
                                                                                                                                      ~
_ ADMINISTRATIVE CONTROLS STARTUP REPORTS (Continued) 6.9.1.2 The st'a'   r tup report shall address each of the tests identified in the FSAR'and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications.                                    Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report.
6.9.1.3 Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest.       If the Startup Report does not cover all three events, i.e. , initial criticality, completion of startup test program, and resumption or commencement of commercial operation, supplementary' reports shall be submitted at least every three months until all three events have been completed.
ANNUAL REPORTS1!
6.9.1.4 Annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year.
The initial report shall be submitted prior to March 1 of the year following initial criticality.
:0 6.9.1.5   Reports required on an knnual basis shall include 4a tabulation on an                                         l annual basis of the nuaber of station, utility, and other personnel, including contractors, receiving exposures greater than 100 ar manrem exposure according to work and job functions,g/yr                      e.g., reactorand               their associated operations and surveillance, inservice inspection, routine maintenance, special maintenance                                            '
(describe maintenance), waste processing, and refueling. The dose assignments to various duty functions say be estimated based on pocket dosimeter, TLD, or i           film badge measurements. Small exposures totalling less than 20 percent of
(           the individual total dose need not be accounted for. In the aggregate, at                                                   ;
least 80 percent of the total whole body dose received from external sources should be assigned to specific major work functions.
h Mor ume.f$cSem ek O Ckede                    %    Mb                               50k*-k f8I N V81V85-4 1
_1/A single submittal may be made for a multiple unit station. The submittal                                                 7 should combine those sections that are common to all units at the station.
This tabulation supplements the requirements of 520.407 of 10 CFR Part 20.
l GRAND GULF-UNIT 1                      6-16

17.(Ga ns-x2D                          _
ADMINISTRATIVE CONTROLS i
PROMPT NOTIFICATION WITH WRITTEN FOLLOWUP (Continued)                                                        r
: d.         Reactivity anomalies involving disagreement with the predicted value
                     .        'of reactivity balance under steady state conditions during power opera-
                             - tion greater than or equal to 1% delta k/k; a calculated reactivity balance indicating a SHUTD0WN MARGIN less conservative than specified                        !
in the technical specifications; short-term reactivity increases that correspond to a reactor period of less than 5 seconds or, if subcritical, an unplanned reactivity insertion of more than 0.5% delta k/k; or occurrence of any unplanned criticality.
: e.         Failure or malfunction of one or more components which prevents or could prevent, by itself, the fulfillment of the functional require-ments of system (s) used to cope with accidents analyzed in the SAR.
: f.         Personnel error or procedural inadequacy which prevents or could prevent, by itself, the fulfillment of the functional requirements of systems recuired to cope with accidents analyzed in the SAR.
i'
: g.         Conditions arising from natural or man-made events that, as a direct result of the event, require unit shutdown, operation of safety systems, or other protective measures required by technical specifications.
: h.        Errors discovered in the transient or accident analyses or in the methods used for such analyses as described in the safety analysis report or in the bases for the technical specificaticas that have or could have permitted reactor operation in a manner less conservative than assumed in the analyses.               .
: 1.        Performance of structures, systems, or components that requires remedial action or corrective measures to prevent operation in a manner less conservative than assumed in the accident analyses in the safety analysis report or technical specifications bases; or discovery during unit life of conditions not specifically considered in the safety analysis report or technical specifications that require remedial action or corrective measures to prevent the existence or development of an unsafe condition.
J.        Offsite releases of radioactive materials in liquid and gaseous effluents which exceed the limits of Specification 3.11.1.1 or 3.11.2.1.
: k.        Exceeding the limits in specification 3.11.1.4  Thefor the storage written              of follow-up radioactive materials in the listed tanks.
report shall include a schedule and a description of activities planned and/or taken to reduce the contents to within the specifled limits.
85-k,        M lure c.r enedhc.bi c               de     TC      'kNIM s-20               ***ada'at "a          7 sRANo sutF-UNIT 1

i
: 18. (GGNS - X40)

==SUBJECT:==
Technical Specifications 3.3.1 and 3.3.2, pages 3/4 3-1 and 3/4 3-9.
DISCUSSION:        Action "b" of Technical Specification 3.3.2 requires that with the number of OPERABLE channels less than required by the MINIMUM OPERABLE channels per Trip System requirement for one trip system, place that trip system in the tripped condition
* within one hour. The proposed change is to adopt the Standard Technical Specification wording that allows placing the inoperable channel (s) and/or that trip system in the tripped condition
* within one hour.
Present "*" notation at the bottom of pages 3/4 3-1 and 3/4 3-9 requires that with a design providing only one channel per trip system, an inoperable channel need not be placed in the tripped condition where this would cause the Trip Function to occur.
The proposed change is to change the present "*" notation to agree with the applicable Standard Technical Specification "*"
notation.
JUSTIFICATION: The Standard Technical Specification wording for action "b"                               of
;                               Technical Specification 3.3.2, its associated "*" notation, and the "*" notation of Technical Specification 3.3.1 does not require an inoperable channel to be placed in the tripped condition where this would cause the Trip Function to occur.
However, restoration of the channel is required within 2 hours or ACTIONS required by Table 3.3.2-1 for that Trip Function shall be taken. The Standard Technical Specification clarifies present action "b" of Technical Specification 3.3.2 and "*"
notation which addresses a design providing only one channel per trip system.
SIGNIFICANT HAZARDS CONSIDERATION:
The proposed change brings present action "b" of Technical Specification 3.3.2, its associated "*" notation, and the "*"
notation of Technical Specification 3.3.1 in conformance with the Standard Technical Specifications. This change does not involve the reduction of safety margins and no significant l
increase in the probability or consequences of an accident previously eva*oated is involved nor is the possibility of a new or different kind of accident from any accident previously evaluated created. Thus the proposed change to the Technical Specification does not involve any significant hazards consideration.
4 f
G90sp9
_.               _    -    ..       _          .          _ . _ _ _ _ _ _ . . . ~ . _ _ . . _ . __ ._                        __

f t. CMus - x @
3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor protection system instrumentation channels shown in Table 3.3.1-1 shall be OPERABLE with the REACTOR PROTECTION SYSTEM RESPONSE TIME as shown in Table 3.3.1 2.
APPLICABILITY:    As shown in Table 3.3.1-1.
ACTION:
: a. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for one trip system, place the inoperable channel and/or that trip system in the tripped condition
* within one hour.      The provisions of Specification 3.0.4 are not applicable.
: b. With the number of OPERABLE channels less than required by the Minirum OPERABLE Channels per Trip System requirement for both trip systems, place at least one trip system ** in the tripped condition within one hour and take the ACTION required by Table 3.3.1-1.
SURVEILLANCE REQUIREMENTS 4.3.1.1     Each reactor protection system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.1.1-1.
4.3.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.
4.3.1.3 The REACTOR PROTECTION SYSTEM RESPONSE TIME of each reactor trip functional unit shown in Table 3.3.1-2 shall be demonstrated to be within its limit at least once per 18 months. Each test shall include at least one chan-nel per trip system such that all channels are tested at least once every N times 18 months where N is the total number cf redundant channels in a specific reactor trip system.
            "itt : d::ign pr:;iding 0:1y ::: ch::::1rt-t-i;ryctr.$ninoperable                   l channel need not be placed in the tripped condition where this would cause the Trip Function to occur. In these cases, the inoperable channel shall be restored to OPERABLE status within 2 hcurs or the ACTION required by i
Table 3.3.1-1 for that Trip Function shall be taken.
mm If more channels are inoperable in one trip system than in the other, place the trip system with more inoperable channels in the tripped condition, except when this would cause the Trip Function to occur.
l        GRAND GULF-UNIT 1                      3/4 3-1

INSTRUMENTATION
[$, (68 A/S-X YD) 3/4.3.2 !$0LATION ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION e
d 3.3.2 The isolation actuation instrumentation channels shown in Table 3.3.2-1
  '                                                  shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.2-2 and with ISOLATION SYSTEM RESPONSE TIME as shown in Table 3.3.2-3.
APPLICA81LITY:               As shown in Table 3.3.2-1.                                                                                                                             ,
ACTION:
: a. With an isolation actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.2-2, dec1.are the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.
b.-  With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for one trip system, lac that trip system in the tripped condition
* within one hour.                                                                                            l 14, ;,,,,,.Ak he provisions of Specification 3.0.4 are not applicable.
                              .A. . ., ico
                            "*'E/*"                                  c. With the number of OPERABLE channels less than required by the Minimum
;                                                                         OPERABLE Channels per Trip System requirement for both trip systems, place at least one trip system ** in the tripped condition within one hour and take the ACTION required by Table 3.3.2-1.
SURVEILLANCE REQUIREMENTS 4.3.2.1 Each isolation actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.2.1-1.
4.3.2.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.
4.3.2.3 The ISOLATION SYSTEM RESPON.LE TIME of each isolation trip function I                                                             shown in Table 3.3.2-3                 Each      shall testbeshall demonstrated include at least                   to beone   within its limit channel              per    at trip  least once per 18 months.
system such that all channels are tested at least once every N times 18 months, l
where N is the total number of redundant channels in a specific isolation trip system.
l "S!! :     ::f;n ; :;idin; : Y ::: ch::::1 ;:r tri; :y t: , J n inoperable channel need not be placed in the tripped condition where this would cause the Trip Function to occur. In these cases, the inoperable channel shall be restored to OPERABLE status within 2 hours or the ACTION required by Table 3.3.2-1 for that Trip Function shall be taken.
                                                                **If more channels are inoperable in one trip system than in the other, place the trip system with more inoperable channels in the tripped condition, except when this would cause the Trip Function to occur.
GRAND GULF-UNIT 1                                          3/4 3-9
    - - - - - - - - - . - _ - . - - - - - - - - - - - - - - - _                       -__-,,--,-,-w---,-,--w,e                -n_a-.-,---n,._-,.,,.,,-,e-_                                m,,,w.--,,,-,,n,,,, , -, , , -,,   nn,g_m,..,-m
: 19. (GGNS - X36)

==SUBJECT:==
Technical Specification Table 4.3.7.12-1, pages 3/4 3-92 and 3/4 3-93.
DISCUSSION:     The flow rate monitors listed in Table 4.3.7.12-1 as items 1.d.
3.d 4.d and 5.d presently have Quarterly frequencies for Channel Functional Tests. The proposed change is to delete the Channel Functional Test requirements for these items.
JUSTIFICATION: This change to the Technical Specifications reflects the design and as-built condition of the flow rate monitors for the following systems:
: a. Radwaste Building Ventilation Monitoring
: b. Containment Ventilation Monitoring
: c. Turbine Building Ventilation Monitoring
: d. Fuel Handling Area Ventilation Monitoring These flow rate monitors measure the total flow being discharged at each release point and provide input for release rate calculations. The flow rate monitors (above) provide input to local flow recorders, but do not provide alarm or trip functions that would require verification by a Channel Functional Test. The present Channel Calibration on a refueling frequency and the Channel Check on a daily basis are not affected by this change.
Channel Functional Tests are used to check trip actuations, alarms and channel failure trips. The flow rate monitors listed as items 1.d, 3.d, 4.d. and 5.d in Table 4.3.7.12-1 do not have any trip or alarm functions and as such the Channel Functional Test requirement should be deleted.
SIGNIFICANT HAZARDS CONSIDERATION:
Since there are no alarm or trip functions from the flow rate monitors listed above, the deletion of the Quarterly Channel Functional Tests presents no significant hazards considerations. This change does not involve the reduction of safety margins and no significant increase in the probability or consequences of an accident previously evaluated is involved nor is the possibility of a new or different kind of accident from any accident previously evaluated created. Thus the proposed change to the Technical Specification does not involve any significant hazards consideration.
G90sp5

j                                                                                                                      TABLE 4.3.7.12-1                                                            .
RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS o
MDDES IN WNICH 8                                                                                                                           CHANNEL l                            G                                                                                CHANNEL           SOURCE    CHANNEL       FUNCTIONAL      SURVEILLANCE CHECK  CALIBRATION        TEST           REQUIRED E                               INSTRUMENT                                             CHECK
:                             E I                             "                               1. RADWASTE BUILDING VENTILATION                                                                     ,
i                             "                                        MONITORING SYSTEM
: a.      Noble Gas Activity Monitor -                                                                      *
.;                                                                            Providing Alarm                          D         M          R(3)           Q(2)
I                                                                   b. Iodine Sampler                              W         N.A.       N.A.           N.A.
t
* i                                                                   c. Particulate Sampler                         W       . N.A.        M.A.          N.A.
* l w                                   d. Flow Rate Monitor                           D         N.A.        R            t W.A w                                  e. Samp1'er Flow Rate Monitor                  D          N.A.       R             N.A.
a N
l                                                             2. MAIN CONDENSER OFFGAS TREATMENT i                                                                       SYSTEM EXPLDSIVE GAS MDNITORING                                                                                     g -
j                                                                       SYSTEM
: a. Nydrogen Monitor                             D         N.A.       Q(4)          M
 '                                                                                                                                                                                              2
: 3. CONTAINMENT VENTILATION                                                                                                     w MONITORING SYSTEM                                                                                                      X
-}                                                                                                                                                                                              w
: a. Noble Gas Activity Monitor                                                                                            w
!
* v 1                                                                             Providing Alarm                           0         M          R(3)           Q(2)
: b. Iodine Sampler                              W         N.A.       N.A.           N.A.                                      ,
: c. Particulate Sampler                         W       . N.A.       N.A.           N.A.
1 j                                                                   d. Effluent System Flow Rate                                                                                       l i                                                                            Monitor                                   D         N.A.        R            t N.A
* i I                                                                                                                                                                             *
,                                                                    e. Sampler Flow Rate Monitor                   D         N.A.        R             N.A.                                     r i                                                                                                                                                                                                    .
L                                                                                                  ,

V           -
TABLE 4.3.7.12-1 (Continued)
E g                    RADI0 ACTIVE GASEOUS EFFLilENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS e
G;                                                                                           CilANNEL    MODES IN WillCH a
2 CilANNEL     SOURCE       CllANNEL      FUllCTIONAL   SURVEILLANCE INSTRUMENT                                        CilECK       CllECK   CALIBRATION          TEST        REQUIRED
    ~  4. TURBINE BLDG. VENTILATION MONITORING SYSTEM
: a. Noble Gas Activity Monitor              D         M              R(3)           Q(2)
: b. Iodine Sampler                          W         N.A.           N.A.           N.A.           *
            'c . Particulate Sampler                     W         N.A.           N.A.           N.A.           *
: d. Flow Rate Monitor w
D         N.A.           R             t N.A-
* N
,   w        e. Sadpler Flow Rate Monitor               D         N.A.           R              N.A.           *
: 5. FUEL IIANDLING AREA VENTILATION MONT0 RING SYSTEM
: a. No*le a     Gas Activity Monitor            D         M              R(3)           Q(2)
{          b.      Iodine Sampler                          W         H.A.           N.A.           N.A.
* 4 j           c.      Particulate Sampler                     W.        N.A.          N.A.           N.A.           *
: d.      Flow Rate Monitor D         N.A.           R            tN. A.          *
                                                                                                                         )l
: e.      Sampler Flow Rate Monitor               D         N.A.          R              N.A.
* Q h
 ,.                                                                                                                       d, i
v
: 20. (GCNS - 783)

==SUBJECT:==
Technical Specification 6.5.1.2, page 6-7.
DISCUSSION:    The proposed ch'anges add two members to the PSRC in Specification 6.5.1.2.
The plant responsibilities of the proposed new members are described in FSAR Section 13.1.2.2.10. Both positions are shown in FSAR Figure 13.1-2 reporting to the Technical Support Superintendent.
JUSTIFICATION: The proposed change to Specification 6.5.1.2 not only expands the membership of the PSRC by the addition of two members whose job specification requires a broad-based knowledge of GGNS operation, it also enhances the capabilities of the PSRC by permitting the insights based on these new members' functional responsibilities to be utilized in the review function.
SIGNIFICANT HAZARDS CONSIDERATION:
The proposed changes are administrative in nature. They do not increase the probability or consequences of an accident previously evaluated nor do they create the possibility of a new or different accident from those previously evaluated.
These proposed changes do not constitute a significant hazards consideration.
l
(
f l
i G75sp1

20(GSMS- 793 )
1 ADMINISTRATIVE CONTROLS PLANT SAFETY REVIEW COMITTEE (PSRC) (Continued)                                                             ,
C05tPOSITION 6.5.1.2 The PSRC shall be composed of the:                                                                  ;
Chairman:             Assistant Plant Manager Vice Chaiman:         Nuclear Support Manager Member:               Operations Superintendent i                                             Member:               Technical Suppert Superintendent Member:               Quality Superintendent Member:               Chemistry and Radiation Protection Superintendent Member:               Maintenance Superintendent
'                                             m ~k a '              re< A ,'es t                             f. pre sa,.
                                              "- :                  a   .r.-       gn.,
: e. aerj
                                                                                           . .. ,g s.,e . .. ,
l ALTERNATES 6.5.1.3 All alternat.e members shall be appointed in writing by the Plant Manager to serve on a temporary basis; however, no more than two alternates shall participate as voting members in PSRC activities at any one time.                                           ,
MEETING FREQUENCY f                                    6.5.1.4 The PSRC shall meet at least once psr calendar month and as convened 4
by the PSRC Chairman or Vice Chairman.
I QUORUM j                                        6.5.1.5 The minimum quorum of the PSRC necessary for the performance of tbc PSRC responsibility and authority provisions of these Technical Specifications shall consist of the Chairman or Vice Chairman and four members including alternates.
RESPONSIBILITIES 6.5.1.6 The PSRC shall be responsible for review of:
: a. Station administrative procedures and changes thereto.
: b. The safety evaluations for 1) procedures, 2) changes to procedures, equipment or systems, and 3) tests or experiments completed under the provision of Section 50.59, 10 CFR, to verify that such actions did not constitute an ur.'eviewed safety question and all programs required by Specification 6.8 and changes thereto.
: c. Proposed procedures and changes to procedures, equipment or systems which may involve an unreviewed safety question as defined in Section 50.59, 10 CFR.
: d. Proposed tests or experiments which may involve an unreviewed safety question as defined in Section 50.59, 10 CFR.
: e. Proposed changes to Technical Specifications or the Operating License.
GRAND GULF-UNIT 1                           6-7
: 21. (GGNS - 770)

==SUBJECT:==
Technical Specification Table 3.3.3-2 and ?.3.5-2, pages 3/4 3-28 and 3/4.3-47.
j              DISCUSSION:        The subject tables currently identify the allowable value for the Supression Pool Level-High HPCS and RCIC isolation setpoints as 6.5".                                     This should be corrected to 7" in both tables.
JUSTIFICATION: This change is proposed to make the Technical Specifications consistent with current plant design. GGNS design information (level setting diagram) has been reviewed. An elevation of 111'-10" corresponds to a level of 7" in the transmitter range.
The design specification data sheets require the nominal trip setpoint to be 1.1" below the technical specification limit; therefore, the 5.9" figure in the tables is correct.
SIGNIFICANT HAZARDS CONSIDERATION:
i                                 The level criteria for HFCS and RCIC isolation has not changed;
}                                 only the reference elevation of the instrumentation was revised. The proposed change does not significantly increase the probability or consequences of an accident previously evaluated nor do.es it create the possibility of a new or different accident from those previously evaluated. It does not constitute a signific.'nt hazards consideration.
4 i
                                                     \
G81sp5

fi.(GGN5 77c) n                                        F
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Ci                                          i 3/4 3 28 SRAND SULF-UNIT 1

                                                                                                                                            ?
 . ss Ta8tE 3.3.5-2 E
     =
     @                                                       REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUNENTATION SET
     !i                                                                                                                ALLOWA8LE
 !                                                                                                  TRIP SETPOINT        VALUE_
      -4 FUNCTIONAL UNITS N                                        a. Reactor Vessel Water Level -                                      1 -43.8 inches Low Low, Level 2                              1 -41.6 inches
* Reactor Vessel Water Level - High, level 8       1 53.5 inches
* 1 55.7 inches b.
~
Condensate Storage Tank Level - Low              1 0 inches       1 -3 inches
!                                              c.
l                                                                                                    1 5.9 inches    1 4 6 inches
: d. Suppression Pool Water level - High                                   ~1. 0 l
NA i                                                                                                    NA w                                        e. Manual Initiation                                                                   n 4
1 w
1
 !     A                      "See Bases Figure 8 3/4 3-1.
w i
i
                                                                                                                                         .H
 !                                                                                                                                       o         -
:                                                                                                                                        O r   2 in l                                                                                                                                        -4
:                                                                                                                                        -a i
.i I
: 22. (GGNS - X15)

==SUBJECT:==
Technical Specification 3.6.2.5, page 3/4 6-18, and its Bases, page B3/4 6-3.
DISCUSSION:    This specification currently states that the differential pressure from the drywell to the containment shall be maintained between (-) 0.1 and (+) 2.0 psid. In the bases 3/4.6.2.5 the maximum external drywell dif ferential pressure is given as 1.0 psid. Both the low end of the range of differential pressures ((-) 0.1) and the maximum drywell external differential pressure in the Bases 3/4.6.2.5 (1,0) are incorrect; these values should be the same magnitude.
These values should be revised as follows:
: a. The range in Technical Specification 3.6.2.5 should read
(-) 0.26 to 2.0 psid; and
: b. the discussion in the Bases should utilize a value of
                              +0.26 psid.
The dif ference in the two signs (positive / negative) being used is required due to the way the value is used in each discussion.
JUSTIFICATION: The (-) 0.1 figure was derived from the Standard Technical Specification without verifying the plant specific information.
The number in the basis is a typographical error in copying the STS information.
In the original design the drywell vacuum breaker setpoints were established at (-) 0.5 psid. However, in resolving the Humphrey concern relative to overflow of the weir wall (ACTION PLAN 29) a revision of these setpoints to (-) 0.15 psid was made. Utilizing this setpoint and considering an instrumentation tolerance of 10.11 psid, the drywell pressure can be maintained at (-) 0.26 psid relative to the containment pressure.
The proposed change makes the technical specification representative of the plant design which incorporates changes to resolve one of the Humphrey concerns.
SIGNIFICANT HAZARDS CONSIDERATION:
The proposed change better represents the plant design. The particular aspect of the design which affects this change, the vacuum breaker setpoint, has been analyzed in studying the Humphrey concerns. The proposed change does not significantly increase the probability or consequences of an accident nor does it create the possibility of a new or different accident from those previously evaluated. It does not constitute a significant hazards consideration.
H2sp1

                                                                                                 /GGNS-XI5)
L CONTA!I00ENT SYSTEMS l
i DRYWELL INTERNAL PRESSURE LIMITING CONDITION FOR OPERATION                                                                                                  l 3.6.2.5 Drywell to containment differential pressure shall be maintained between N and + 2.0 psid.
                    -0.26 APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.
ACTION:
With the drywell to containment differential pressure outside of the specified limits, restore the differential pressure to within the limits within I hour or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
SURVEILLANCE REQUIREMENTS 4.6.2.5 The drywell to containment differential pressure shall be determined to be within the limits at least once per 12 hours.
i l
l l
I i
l l
l l
GRAND GULF-UNIT 1                                3/4 6-18

22, ( GGNS-%15)
CONTAINMENT SYSTEMS BASES CONTAllMENT PURGE SYSTEM (Continued)
Leakage integrity tests with a maximum allowable leakage rate for purge supply,and exhaust isolation valves will provide early indication of resilient material seal degradation and will allow the opportunity for repair before gross leakage fail'ures develop. The 0.60 L leaking limit shall not be exceededwhentheleakageratesdetermined$ytheleakageintegritytestsof these valves are added to the previously detemined total for all valves and penetrations subject to Type B and C tests.
3/4.6.2" DRYWELL 3/4.6.2.1 ORYWELL INTEGRITY Drywell integrity ensures that the steam released for the full spectrum of drywell p pe breaks is condensed inside the primary containment either by the suppress on pool or by containment spray. By utilizing the suppression pool as a heat sink, energy released to the containment is minimized and the severity of the transient is reduced.
3/4.6.2.2 DRYWELL BYPASS LEAKAGE The limitation on drywell bypass leakage rate ensures that the maximum leakage which could bypass the suppression pool during an accident would not result in the containment exceeding its design pressure of 15.0 psig. The integrated drywell leakage value is limited to 10% of the design drywell leakage rate.
The limiting case accident is a very small reactor coolant system break which will not automatically result in a reactor depressurization. The long term dif-i            ferential pressure created between the drywell and containment will result in a l
significant pressure buildup in the containment due to this bypass leakage.
3/4.6.2.3 DRYWELL AIR LOCKS The limitations on closure for the drywell air locks are required to meet the restrictions on DRYWELL INTEGRITY and the drywell leakage rate given in Specifications 3.6.2.1 and 3.6.2.2. The specification makes allowances for the fact that there may be long periods of time when the air locks will be in a closed and secured position during reactor operation. Only one closed door in each air lock is required to maintain the integrity of the drywell.
3/4.6.2.4 DRYWELL STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the drywell will be maintained comparable to the original design specification for the life of the unit. A visual inspection in conjunction with Type A leakage tests is sufficient to demonstrate this capability.
3/4.6.2.5 DRYWELL INTERNAL PRESSURE                              + O ~J 6 The limitations on drywell-to-containment differential pressure ensure                            ._.
that the drywell peak pressure of 22.0 psig doe at exceed the design pressure of 30.0 psig and that the containment peak pressur of 11.5 psig does not exceed the design pressure of 15.0 psig during LOCA                       ons. The maximum external drywell pressure differential is limited to                     psid,   well below the 2.3 psid at which suppression pool water will be forced over the wier wall and into the drywell. The limit of 2.0 psid for initial positive drywell to containment pressure will limit the drywell pressure to 22.0 psig which is less than the design pressure and is consistent with the safety analysis.
GRAND GULF-UNIT 1                               B 3/4 6-3
: 23. (GGNS - 802)

==SUBJECT:==
Technical Specification Table 3.7.6.5-1, page 3/4 7-36 and 37.
DISCUSSION:       The' subject table is being revised to reflect two new hose stations of which one (13G) was added.by design change and the other (55C) was inadvertently omitted from the. table.
                                    .The following changes are proposed co Table'3.7.6.5-1.
To page 3/4 7-36 add hose station'13G located at Q-5.8 on elevation 245'-0".-
To page 3/4 7-37 add hose station 55C located at K.2-18.8 on elevation 177'-0".
JUSTIFICATION: The proposed change makes the operability and surveillance requirements of the Technical Specification applicable to the.
new hose stations. . This change is in keeping-with the intent of the specification as stated in the Bases for 3/4 7.6. The bases states that the OPERABILITY of the fire suppression systems ensures' adequate capability is available to confine and extinguish fires in areas where safety related equipment is
                                     . located. The addition of these hose stations adds to the fire extinguishing capability at Grand Gulf.
SIGNIFICANT HAZARDS CONSIDERATION:
The proposed changes do not significantly increase the
                 ,                    probability or. consequences of an accident previously evaluated nor do they create the possibility of a new or different accident from those previously evaluated. They do not
                                     -constitute a significant hazards consideration.
G91sp2

23, (GGNS- Bo2.)
TABLE 3.7.6.5-1 FIRE H0SE STATIONS HOSE RACK
.          LOCATION                                          ELEVATION                      INDENTIFICATION I
AUXILIARY BUILDING Q.1-6.0                              103'-0"                               13A Q-5.7                                119'-0"                               138 l                        Q.1-6.1                             139'-0"                               13C Q-6.0                               166'-0"                               13D Q-5.9                               185'-0"                                13E Q-6.0       - Q - 5.8               208'-0" -- 24 5'- d'                   13F- 13G Q-11.3                                  93'-0"                             14A P.4-9.0                              119'-0"                               148 P.4-9.0                              139'-0"                               14C P.4-8.6                              166'-0"                               140 P.4-9.5                              185'-0"                               14E P-10                                208'-10"                              14F i                         D.4-12.5                            139'-0"                               15A P.4-12.5                            166'-0"                               158 P.4-13.1                            185'-0"                               15C R-13.7                              208'-10"                              150 M.2-15.1                            103'-0"                               16A M.7-15.1                          .119'-0"                                16B L.7-15.1                            139'-0"                               16C L.7-15.1                             166'-0"                              160 L.7-15.1                             185'-0"                              16E M.7-15.1                             208'-10"                             16F H.3-13.8                             103'-0"                              17A J.4-13.8                             119'-0"                              17B H-13.8                              139'-0"                               17C J-13.8                               166'-0"                               170 G.4-11                              103'-0"                               18A G.4-11.7                            119'-0"                               188 G . 4-12. 2                         139'-0"                               18C G.4-11.3                             166'-0"                              180 G.4-7.5                              103'-0"                              19A 119'-0"                              19B G.4-8.3 139'-0"                              19C G.4-7.5 166'-0"                              190 G.4-8.4 103'-0"                              20A G.6-6.4 119'-0"                              20B G.6-6.4                                                                   20C H-6.2                                 139"-0" 166'-0"                             200 H-6.2 103'-0"                             21A L-6.2                                                                    218 L-6. 2                               119'-0" 139"-0"                             21C L-6.2                                                                     210 L-6.2                                166'-0" GRAND GULF-UNIT 1                                        3/4 7-36
: 23. (GGNS- 802)
TABLE 3.7.6.5-1    (Continued) l FIRE HOSE STATIONS HOSE RACK LOCATION                         ELEVATION                     INDENTIFICATION CONTAINMENT M.7-7.8                    120'-10"                            22A H.8-8.1                    135'-4"                             23A
!             J.1-8.1                    161'-10"                            238 l             J.8-7.2                    184'-6"                             23C l             J.4-7.5                    208'-10"                            23D M.2-7.2                    135'-4"                             24A M.8-7.9                    161'-10"                            24B M.2-7.2                    184'-6"                             24C N-8.2                      208'-10"                            24D M.6-12.4                   135'-4"                             25A N.2-11.5                   161'-10"                            25B N.3-11.3                   208'-10"                            25C J.1-12.0                   135'-4"                             26A J-11.6                     161'-10"                            26B K.2-13.1                   184'-6"                             26C J-11.8                     208'-10"                            26D CONTROL BUILDING J.9-18.8                   133'-0"                             53A K.2-18.8                   111'-0"                             538 G.1-18.4                   111'-0"                             548 G.2-18.4                   133'-0"                             54C G.1-18.7                   148'-0"                             54D G.2-18.8                   166'-0"                             54E G.1-18.7                   189'-0"                             54F K.2-18.8                   148'-0"                             55A K.2-18.8        K.2- 6.g 166 ' -0" e g 77 '_ o"              SSB_ gg K.2-18.8                   189'-0"                              55D DIESEL GENERATOR BUILDING R-10.6                     133'-0"                             66A R-8.4                      133'-0"                             66B GRAND GULF-UNIT 1                     3/4 7-37
: 24. (GGNS - X32)

==SUBJECT:==
Technical Specification Table 3.3.3-2, page 3/4 3-28.
DISCUSSION:    A Technical Specification change was submitted to the subject table as Item 3 of AECM-83/0338 dated June 14, 1983. Part of the requested change modified the Allowable Values for item A.2.f and B.2.e from greater than or equal to 122 psig, increasing to 115-135 psig. Amendment 8 to the Grand Gulf Technical Specifications changed only Item A.2.f (LPCI Pump A Discharge Pressure - High) and did not make the same change in Item B.2.e (LPCI Pump B and C Discharge Pressure - High). The proposed change modifies the Allowable Valve for Item B.2.e on Table 3.3.3.-2 from the present greater than or equal to 122 psig, increasing, to 115-135 psig, increasing.
JUSTIFICATION: The proposed change reflects current General Electric design specification values and does not affect established transient
                         . analyses. The changes to the Allowable Value for the LPCI Pump B and C Discharge Pressure - High are proposed to be consistent with the present Allowable Value for the LPCI Pump A Discharge Pressure - High.
The lower boundary on allowable value ensures a LPCI pump is running prior to ADS initiation. The upper boundary on allowable value ensures that ADS is not prevented from initiating due to a setpoint higher than pump discharge pressure.
SIGNIFICANT HAZARDS CONSIDERATION:
The change to the LPCI Pump B and C Discharge Pressure - High Allowable Values is made to promote consistency with the LPCI Pump A Allowable Value and as such constitutes a purely administrative change to the Technical Specifications. This change does not involve the reduction of safety margins and no significant increase in the probability or consequences of an accident previously evaluated is involved nor is the possibility of a new or different kind of accident from any accident previously evaluated created. Thus the proposed change to-the Technical Specification does not involve any significant hazards consideration.
2 G90sp1

TABLE 3.3.3-2                                                           ,
ENERGENC ' CORE COOLING SYSTEM ACTUATION INSTIRBENTATION SETPOINTS ALLOlNISLE TRIP SETPOINT                     VALUE
: E   TRIP RNICTION l5     A. DIVISION 1 TRIF SYSTEM
: 1. RHR-A (LPCI MDDE) AIS LPCS SYSTEM 3                                                                     > -150.3 inches *       > -152.5 incives Z              a. Reactor Vessel Water Level - Low Low Low, level 1                         i 1. M psig
: b. Drywel' Pressure - High                           7 1.89 psig w                                                                     7 5 seconds             i 5.2$ seconds
: c. LPCI Pump A Start Time Delay Relay NA                      NA Manual Initiation i                 d.
l
: 2. AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "A"
                                                                         > -150.3 inches *       > -152.5 inches
: a. Reactor vessel Water Level - Low Low Low, Level 1                         i 1.94 psig i
: b. Drywell Pressure - High                           7 1.89 psig i 105 seconds           i 117 seconds
: c. ADS Timer                                                                 I 10.8 inches
)
: d. Reactor Vessel Water Level-Low, level 3           I 11.4 inches *
!                                                                        145 psig, increasing    125-165 psig, increasing i                 e. LPCS Pump Discharge Presgure-High                                          115-135 psig, increasing LPCI Pump A Discharge Pressure-High                125 psig, ir.::reasing
) g,               f.
: g. Manual Initiation                                  NA                      NA j
1  w    B. DIVISION 2. TRIP SYSTEM fg            1. RHR B AND C (LPCI ISDE)
Reactor Vessel Water Level - Low Low Low, level 1  > -150.3 inches *       > -152.5 inches
: a.                                                                             31.94psig l
: b. Drywell Pressure - High                            31.89psig
                                                                          < 5 seconds             < 5.25 seconds
;                  c. LPCI Pump 8 Start Time Delay Relay                                        NA NA
;                  d. Manual Initiation l             2. AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "B" > -150.3 inches *            > -152.5 inches
:                   a. Reactor Vessel Water Level - Low Low Low, Level 1  31.89psig               31.94psig j                  b. Drywell Pressure - High                           < 105 seconds           < 117 seconds i                   c. ADS Timer                                         I 11.4 inches
* I 10.8 inches
: d. Reactor Vessel Water Level-Low, Level 3 LPCI Pump 8 and C Discharge Pressure-High         125 psig, increasing -499 psig, Increasing         l l                   e.                                                    NA                      M4 %,,g _g 3 g           g
: f. Manual Initiation                                                                                  $
l
;        C. DIVISION 3 TRIP SYSTEM                                                                                      q
: 1. HPCS SYSTEM                                                                                              qs g-41.6 inches
* 1-43.8 inches i              a. Reactor Vessel Water Level - Low Low, level 2            5 1.89 psig             5 1.94 psig             qs l
: b. Drywell Pressure - High                                                          $ 55.7 inches           7
: c. Reactor Vessel Water Level - High, level 8               5 53.5 inches
* 3 inches 1 0 inches                                       m l
: d. Condensate Storage Tank Level - Low                      < 5.9 inches            1
                                                                                                   < 6.5 inches              3
;              e. Suppression Pool Water Level - High                      NA                      NA                       X l
: f. Manual Initiation                                                                                       w
:                                                                                                                           N
'                                                                                                                          V

      +
: 25. (GGNS - X33)

==SUBJECT:==
Technical Specification Table 3.8.4.2-1, page 3/4 8-44.
DISCUSSION:     The proposed change corrects the listing for valves Q1M71F593A and Q1P41F018 which should be Q1M71F593 and Q1P41F018A, respectively.
JUSTIFICATION: This change corrects two typos in Table 3.8.4.2-1 and is made to reflect the correct valve numbers.
SIGNIFICANT HAZARDS CONSIDERATION:
This change involves the correction of two typos and as such is purely administrative in nature. This change does not involve the reduction of safety margins and no significant increase in the probability or consequences of an accident previously evaluated is involved nor is the possibility of a new or different kind of accident from any accident previously evaluated created. Thus the proposed change to the Technical Specification does not involve any significant hazards consideration.
l l
G90sp2
                               ..       -                     ,.._.~           _ . . _ . _ . _ . , . _ . _ _

25 CG&ns -x 5 3 )
TABLE 3.8.4.2-1 (Continued)
'                           leTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION BYPASS DEVICE (CON-TINUOUS) (ACCIDENT      SYSTEM (S)
VALVE ORNSER                CONDITIONS) (NO)        AFFECTED Q1G41F028                       Continuous          Spent Fuel Pool Cooling and Cleanup System Q1G41F029                       Continuous          Spent Fuel Pool cooling and Cleanup System Q1G41F044                       Continuous          Spent Fuel Pool Cooling and Cleanup System Q1G41F021                           No              Spent Fuel Pool Cooling and
  -                                                                    Cleanup System Q1G41F043                           No              Spent Fuel Pool Cooling and Cleanup System
* Containment /Drywell I&C Q1M71F591A
;
* Containment /Drywell I&C       l Q1M71F593r
* Containment /Drywell I&C Q1M71F5928
* Containment /Drywell I&C Q1M71F595
* Containment /Drywell I&C Q1M71F5918
* Containment /Drywell I&C Q1M71F592A
* Containment /Drywell I&C Q1M71F594
      ~
Continuous          Makeup Water Treatment System Q1P21F017 Continuous          Makeup Water Treatment System Q1P21F018 Continuous          SSW Systan 01P41F237                                                                         I Continuous          SSW System Q1P41F018A Continuous          SSW System Q1P41F241 Continuous          SSW System Q1P41F238 Continuous          SSW System QSP41F081A Continuous          SSW System QSP41F064A Continuous          SSW System Q1P41F068A Continuous          SSW System Q1P41f014A                                          SSW System Q1P41F159A                      Continuous Continuous          SSW System 01P41F160A Continuous          SSW System Q1P41F113 Continuous           SSW System Q1P41F168A Continuous           SSW System Q1P41F001A                                           SSW System Q1P41F016A                      Continuous Continuous           SSW System Q1P41F015A                                           SSW System Q1P41F006A                       Continuous Continuous          SSW System Q1P41F005A                                           SSW System Q1P41F007A                       Continuous Continuous          SSW System QSP41F074A                                          SSW System QSP41F066A                       Continuous Continuous          SSW System QSP41F125                                           SSW System Q1P41F0188                      Continuous Continuous          SSW System Q1P41F1608                                          SSW System Q1P41F159B                      Continuous Continuous          SSW System Q1P41F1688                                          SSW System QSP41F154                  Accident Conditions GRAW GULF-UNIT 1                     3/4 8-44 C__-_.___.________
: 26. (GGNS - X34)

==SUBJECT:==
Technical Specification Surveillance Requirement 4.8.1.1.2.d.9, page 3/4 8-6.
DISCUSSION:      Amendment 8 to the Grand Gulf Technical Specifications changed the subject specification by correcting the referenced Surveillance Requirement paragraph from 4.8.1.1.2.d.4.a to 4.8.1.1.2.d.7.a. This change makes the same correction in notation "*" at the bottom of page 3/4 8-6.
JUSTIFICATION: The correction to the "*" note was inadvertently omitted when the change to 4.8.1.1.2.d.9 was previously submitted to the NRC. This change corrects a typo.
SIGNIFICANT HAZARDS CONSIDERATION:
This change involves the correction of a typo and as such is purely administrative in nature. This change does not involve the reduction of safety margins and no significant increase in the probability or consequences of an accident previously evaluated is involved nor is the possibility of a new or different kind of accident from any accident previously evaluated created. Thus the proposed change to the Technical Specification does not involve any significant hazards consideration.
l l
i i
i i
j        G90sp3 l
t

2 b, (G6NS -X3 9-)
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
: 9. Verifying the diesel generator operates for at least 24 hours.
During the first 2 hours of this test, the diesel generator shall be loaded to greater than or equal to 7700 W for diesel gen-erators 11 and 12 and 3630 W for diesel generator 13 and during the remaining 22 hours of this test, the diesel generator shall be loaded to 7000 W for diesel generators 11 and 12 and 3300 W for diesel generator 13. The generator voltage and frequency shall be 4160 1 416 volts and 6011.2 Hz within 10 seconds after the start signal; the stea% state generator voltage and frequency shall be maintained within these limits during this test. Within 5 minutes after completing this 24-hour test, perform Surveillance Requirement 4.8.1.1.2.d.7.a).2) and b).2)*.
: 10. Verifying that the auto-connected loads to each diesel generator do not exceed the continuous rating of 7000 W for diesel generators 11 and 12 and 3300 W for diesel generator 13.
: 11. Verifying the diesel generator's capability to:
a)   Synchronize with the offsite' power source while the generator is loaded with its emergency loads upon a simulated restoration of offsite power, b)   Transfer its loads to the offsite power source, and c)   8e restored to its staney status.
: 12. Verifying that with the diesel generator operating in a test mode and connected to its bus that a simulated ECCS actuation signal:
a)    For Divisions 1 and 2, overrides the test mode by return-ing the diesel' generator to standtsy operation.
b)    For Division 3, overrides the test mode by bypassing the diesel generator automatic trips per Surveillance Require-l                                 ment 4.8.1.1.2.d.8.b).
: 13. Verifying that with all diesel generator air start receivers          i I
pressurized to less than or equal to 256 psig and the compres-sors secured, the diesel generator starts at least 5 times from ambient conditions and accelerates to at least 441 rps for diesel generators 11 and 12 and 882 rps for diesel generator 13 in less than or equal to 10 seconds.
a                                              7                                     l I
If Surveillance Requirement 4.8.1.1.2.d.g a)2) or b)2) are not satisfactorily        .
completed, it is not necessary to repeat the preceding 24 hour test. Instead,        l the diesel generator any be operated at rated load for one hour or until operating temperatures have stabilized.
GRAND GULF-UNIT 1                      3/4 8-6                                         l

2
               -27. (GGNS - X38)

==SUBJECT:==
Technical Specification Table 3.3.1-1, page 3/4 3-4.
DISCUSSION:        Present ACTION 2 in Table 3.3.1-1 requires verification that all control rods are inserted in the core. The proposed change is to verify all insertable control rods are inserted in the core.
JUSTIFICATION: The Standard Technical Specification for ACTION 2 on Table 3.3.1-1 requires verification that all insertable control rods are inserted-in the core. This change takes into consideration any rod that is inoperable due to being immovable.
              - SIGNIFICANT HAZARDS CONSIDERATION:
This change is purely administrative in nature since it involves conforming present Technical Specifications with the Standard Technical Specifications. This change does not involve the reduction of safety margins and no significant increase in the probability or consequences of an accident previously evaluated is involved nor is the possibility of a new or different kind of accident from any accident previously evaluated created. Thus the proposed change to the Technical Specification does not involve any significant hazards consideration.
l l
l l
G90sp8
                                                        . _._ --     ~ . _ . _ . _ _ _ _ - - - _ _ . _ , - _ . _ . _ _ , _ _ _ _

27, cssus-x 3 o)
INSTRUMENTATION TABLE 3.3.1-1 (Continued)
REACTOR PROTECTION SYSTEN INSTRUMENTATION ACTION ACTION 1    -
Be in at least HOT SHUTDOWN within 12 hours.                          .
insuTakl*
Verify all 3control     rods to be inserted in the core and lock the ACTION 2    -
l reactor mode switch in the SHUTDOWN position within one hour.
ACTION 3    -
Suspend all operations involving CORE ALTERATIONS *, and insert all insertable control rods within one hour.
ACTION 4    -
Be in at least STARTUP within 6 hours.
ACTION 5    -
Be in STARTUP with the main steam line isolation valves closed within 6 hours or in at least HOT SHUTDOWN within 12 hours.
ACTION 6    -
Initiate a reduction in THERMAL POWER within 15 minutes and reduce turbine first stage pressure to less than the automatic bypass setpoint within 2 hours.
ACTION 7    -
Verify all insertable control rods to be inserted within one hour.
ACTION 8    -    Lock the reactor mode switch in the SHUTDOWN position within one hour.
l                 ACTION 9     -
Suspend all operations involving CORE ALTERATION 5*, and insert all insertable control rods and lock the reactor mode switch in the SHUTDOWN position within one hour.
                   *Except movement of IRM. SRM or special movable detectors, or replacement of IPRM strings provided SRM instrumentation is OPERABLE per Specification 3.9.2.
l                                                                                                            .
e l
GRAND GULF-UNIT 1                          3/4 3-4
: 28. (GGNS - 788)

==SUBJECT:==
Technical Specification Table 4.3.3.1-1, page 3/4 3-31.
DISCUSSION:        A reference to note (a) is to be added to the CHANNEL CALIBRATION entry for item A.2.c. LPCS Pump Discharge Pressure
                             - High. This reference was inadvertently omitted from the Technical Specification.
JUSTIFICATION: The proposed change adds a requirement to calibrate the above trip unit' monthly. This revision will make the calibration requirements from this trip unit consistent with those of other Division 1, 2, and 3 trip systems.
SIGNIFICANT HAZARDS CONSIDERATION:
The proposed change is an additional surveillance requirement.
It does not significantly increase the probability or consequences of an accident previously evaluated nor does it create the possibility of a new or different accident from any previously evaluated. It does not constitute a significant hazards consideration.
J C91 spi

TABLE 4.3.3.1-1
                                                                                                                                      ~
!                       EMERGENCY CORE CODLING SYSTEM ACTMTION INSTRMENTATION SURVEILLANCE REQUIRGENTS o                                                                         CHAMEL                         OPEMTIOML g                                            CHAMEL                     FUNCTIONAL   CHAMEL          CONDITIONS FOR WICM 4
q; TRIP FMCTIM                               _ CHECK                      TEST     CALIBRATION       SUL~dILLANCE REQUIRED h  A. DIVISION I TRIP SYSTEM j  U        1. RHR-A (LPCI MODE) Ale LPCS SYSTEM
!"              a. Reactor Vessel Water Level -
!                      Low Low Low, Level 1         S                        M                 I        1, 2, 3, 4*, 5*
;               b. Drywell Pressure - High        S                        M               R R II
                                                                                                 **I    1, 2, 3
.               c. LPCI Pump A Start Time
:                      Delay Relay                  M                                        I          1, 2, 3, 4*, 5*
: d. Manual Initiation              NA                       M(bIIcI R               Q IdI      1, 2, 3, 4*, 5*
l           2. AUTOMTIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "A"#
l 1 **            a. Reactor Vessel Water Level -
lD                     Low Low Low,. Level 1        5                        M               RI *I      1,2,3 jy              b. Drywell Pressure-High          5                        M               RI ")      1, 2, 3
;g              c. ADS Timer                      M                        M               Q          1,2,3
!               d. Reactor vessel Water Level -
I                      Low, level 3                 S                        M               RI *)      1, 2, 3 l                e. LPCS Pump Discharge l                      Pressure-High                S                        M               Pq         1,2,3
: f. LPCI Pump A Discharge i                      Pressure-High                S                                        Rg,)       1,2,3 i                g. Manual Initiation              MA                       M(b)
R               NA         1,2,3 l     8. DIVISION 2 TRIP SYSTEM                                                                                                g I           1. RHR 8 AND C (LPCI MODE)                                                                                        y l                a. Reactor Vessel Water Level -
Rg,)
m
!                      Low Low Low, Level 1         5                        M                          1,2,3,4*,5*             O i                b. Drywell Pressure - High        5                        M               R(,)       1, 2, 3                 p l                c. LPCI Pump B Start Time                                                                                     g
!                      Delay Relay                  M N(d) 1, 2, 3, 4*, 5*
M(b)(*)
l
: d. Manual Initiation              MA                       R               Q          1, 2, 3, 4*, 5*         [
                                                                                                                                -4 00 b

l AfD       Ousnges      On    TMa      Pamps - IdemWon 0dy TABLE 4.3.3.1-1 (Continued)          2F, @N s - 7FT) !
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS                           !
NOTATION i
f           Not required to be OPERABLE when reactor steam done pressure is less than or equal to 135 psig.
Applicable when the system is required to be OPERABLE per Specification 3.5.2 or 3.5.3.
Required when ESF equipment is required to be OPERABLE.
(a) Calibrate trip unit at least once per 31 days.
(b) Manual initiation switches shall be tested at least once per 18 months during shutdown. All other circuitry associated with manual initiation shall receive a CHANNEL FUNCTIONAL TEST at least once per 31 days as a part of circuitry required to be tested for automatic system actuation.
(c) Manual initiation test shall include verification of the CPERABILITY of the LPCS and LPCI injection valve interlocks. (See Note 1)
(d) This calibration shall consist of the CHANNEL CALIBRATION of the LPCS and.
LPCI injection valve interlocks with the interlock setpoint verified to be
                      < 150 psig.     (See Note 1)
(e) Tunctional Testing of Time Del y Not Required Note 1:         Until restart after the first refueling outage, the requirements of (c) and (d) above do not apply.
GRAND GULF-UNIT 1                             3/4 3-33
: 29. (GGNS - X43)

==SUBJECT:==
Technical Specification 5.1.3 and Fi.gures 5.1.1-1 and 5.1.3-1, pages 5-1, 5-2 and 5-4.
DISCUSSION:    This change involves correction of terminology and a typo.
Under Note 3 for Figure 5.1.1-1 the reference to Figure 5.1.4-1 should be changed to 5.1.3-1. The words " Unrestricted Area Boundary" should be changed to " Effluent Release Boundary" in 4 places.
JUSTIFICATION: The terminology " Unrestricted Area Boundary" is not appropriate for gaseous and liquid effluents and is inconsistent with the terminology used in the Emergency Flan. The appropriate terminology is " Effluent Release Boundary". The Effluent Release Boundary is shown in Figure 5.1.3-1 instead of the non-existent Figure 5.1.4-1.
SIGNIFICANT HAZARDS CONSIDERATION:
This change is purely administrative in nature and involves only the correction of terminology and a typo. This change does not involve the reduction of safety margins and no significant increase in the probability or consequences of an accident previously evaluated is involved nor is the possibility of a new or different kind of accident from any accident previously evaluated created. Thus the proposed change to the Technical Specification does not involve any significant hazards consideration.
l l
l G96sp2
                                 .-               .-              - - - . .       .~   . , - ,
: n. (G6ns-x n) 5.0 DESIGN FEATURES 5.1 SITE EXCLUSION AREA                                                                                                  5 5.1.1 The exclusion area shall be as shown in Figure 5.1.1-1.
LOW POPULATION ZONE 5.1.2 The low population zone shall be as shown in Figure 5.1.2-1.
EFFLvENT ARELEME
                 -ifMRE9YeteffiHtRE* BOUNDARY FOR GASEOUS EFFLUENTS AND FOR LIQUID EFFLUENTS 5.1.3 The d                        % nt nluez ur.r::tricted    : :: boundary for gaseous effluents and for liquid effluents shall be as shown in Figure 5.1.3-1.                                              The gaseous effluent release points are shown in Figure 5.1.1-1.
5.2 CONTAINMENT C5hFIGURATION 5.2.1 The containment is a steel lined, reinforced concrete structure composed of a vertical right cylinder and a hemispherical dome. Inside and at the bottom of the containment is a reinforced concrete drywell composed of a vertical right
,                   cylinder and a steel head which contains an approximately eighteen to nineteen foot deep water filled suppression pool connected to the drywell through a series of horizontal vents. The containment has a minimum net free air volume of 1,400,000 cubic feet. The drywell has a minimum net free air volume of 270,000 cubic feet.
DESIGN TEMPERATURE AND PRESSURE
                  ~
<                    5.2.2 The containment and drywell are designed and shall be maintained for:
: a. Maximum internal pressure:
: 1. Drywell 30 psig.
: 2. Containment 15 psig.
: b. Maximum internal temperature:
: 1. Drywell 330'F.
: 2. Suppression pool 185'F.
l
: c. Maximum external-to-internal differential pressure:
: 1. Drywell 21 psid.
: 2. Containment 3 psid.
!                     SECONDARY CONTAINMENT 5.2.3 The secondary containment consists of the Reactor Building and the Enclosure Building, and has a minimum free volume of 3,640,000 cubic feet.
GRAND GULF-UNIT 1                                                        5-1
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                                                                                                                                                        $u % :                                                                   y GRAND GULF-UNIT 1                                                                                      5-2

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GRAND GULF-UNIT 1                                                                                               5-4
                                                .,...-.......---,_,,_.---_-e,.,,._m_r- . . _ , .                                   -

O 130. (GGNS - 832)

==SUBJECT:==
Technical Specification 6.10.2, page 6-23.
DISCUSSION:     Technical Specification 6.10.2.1 currently refers to lists of snubbers in Tables 3.7.5-1 and 3.7.5-2. These table numbers are -incorrect and the reference is being revised to Tables 3.7.4-1 and 3.7.4-2 which provide the information.
JUSTIFICATION: The proposed change corrects a typographical error.
SIGNIFICANT HAZARDS CONSIDERATION:
The proposed change is editorial in nature. It does not significantly increase the probability or consequences of an accident previously evaluated nor does it create the possibility of a new or different accident from those previously evaluated. Therefore, this change does not constitute a significant hazards consideration.
C91sp3

N (66NS-832).
ADMINISTRATIVE CONTROLS
          , RECORD RETENTION (Continued) d.'
Records of gaseous    and liquid radioactive material released to the
                           , environs.
: e. Records of transient or operational cycles for those unit components identified in Table 5.7.1-1.
: f. Records of reactor tests and experiments.
: g. Records of training and qualification for current members of the unit staff.
: h. Records of in-service inspections performed pursuant to these Technical Specifications.
: i. Records of Quality Assurance activities required by the Operational Quality Assurance Manual.
: j. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.
: k. Records of meetings of the PSRC and the SRC.
: 1. Records of the service lives of all hydraulic and mechanical snubbers listed on Tables 2.7.5-1 and 3.'.5 ' including the date at which the servicelifecommencesandassociated\installationandmaintenance records.
pA _1               3. l.9 -2
: m. Records of analyses required by the radiological environmental monitoring program.
6.11 RADIATION PROTECTION PROGRAM 6.11.1 Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure.
6.12 HIGH RADIATION AREA 6.12.1 In lieu of the " control device" or " alarm signal" required by paragraph 20.203(c)(2) of 10 CFR 20, each high radiation area in which the intensity of radiation is greater than 100 mrem /hr but less than 1000 mrem /hr shall be barri-caded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP).*
Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:
: a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.
            " Health Physics personnel or personnel escorted by Health Physics personnel shall be exempt from the RWP issuance requirement during the performance of their assigned radiation protection duties, provided they are otherwise following plant radiation protection procedures for entry into high radiation areas.
GRAND GULF-UNIT 1                            6-23

4
: 31. (GGNS - 521)

==SUBJECT:==
Technical Specification Table 3.7.4-2, pages 3/4 7-16 through 3/4 7-25.
DISCUSSION:    Technical Specification Table 3.7.4-2 contains a listing of Safety Related Mechanical Snubbers. The purpose of this change is to make necessary corrections to the table that include the-following:
: 1. Additions of snubbers that were inadvertently left off the table.
: 2. Deletion of snubbers that have been voided, superceded, or incorrectly placed in the table.
: 3. Correct typos 4      Add Non-Q Mechanical Snubbers which were included in the stress analysis of Q-piping.
As part of this change, snubbers with the same number and location are grouped together to prevent duplicate listings.
The snubbers that are grouped together are indicated by number in parentheses.
The changes to Table 3.7.4-2 are discussed in detail in the justification section.
JUSTIFICATION: The changes to Table 3.7.4-2 are administrative in nature and do not involve changes to the plant. These changes are Table corrections due to typo's, inadvertent omissions from the table, and deletions due to superceded, voided or incorrect listings. The inclusion of Non-Q Snubbers is provided since they are on piping which is included in stress analysis calculations that affect Q piping. The changes to Table 3.7.4-2 are justified below:
l                      1. Additions to the table due to inadvertent omissions l                            include:
l                           a. QlB33G122R01, Area 11, Elevation 108 i
: b. Q1B33G355R01, Area 11, Elevation 102 - One of these i                                 snubbers was submitted in AECM-83/0314. There should be two at elevation 102.
: c. Q1B21G023R020, Area 11. Elevation 120
: d. Q1B21G163R01, Area 11, Elevation 113
: e. Q1B21G163R02, Area 11, Elevation 113
: f. Q1B21G189R01, Area ll, Elevation 153
      'H7sp1

               .- .       .-                   _-     -           .                       ~.                                          .--
t
: g. Q1B21G369R01 (2), Area 11. Elevation 148
: h. Q1B21G382R02 (2), Area 11. Elevation 155 - One j                                            additional snubber is at this location.
: i. Q1B21G384R01, Area 11. Elevation 152
: j. Q1E12G025C01 (2), Area 8 Elevation 95 - One additional snubber is at this location.
: k. Q1E31G122R01 (2), Area 11 Elevation 149 - One additional snubber is at this location.
: 1. Q1E51G003R12 (2), Area 8. Elevation 106.
: m. Elevations are added as follows:
Q1G33G002R24. Flevation 102 Q1P41G007R19, Elevation 144 Q1P41G007R20. Elevation 144 Q1P41G007R23 (2), Elevation 138 Q1P41G007R24 (2), Elevation 137
: 2. Deletions to the table are the following:
: a. Q1B21G023R08, Area 11, Elevation 126 - Only one Q1B21G023R08 snubber exists. One remains in the table,
: b. Q1B21G032R04, Area 11, Elevation 127 - Only one Q1B21G032R04 snubber exists. One remains in the table.
;                                        c. Q1B21G218R02 and Q1B21G218R03, Area 11, Elevation 161. These two were not installed in the plant.
: d. Q1B21G382R01, Area 11, Elevation 155. This snubber was voided and not installed.
: e. Q1P41G002CO3, Area 8, Elevation 95. This snubber was superceded by Q1E12G025C01 (2), Area 8 Elevation 95.
: 3. Typo corrections include the following.
: a. Present Q1B21G022R10 should be Q1B21G023R10.
: b. Presently there are two Q1E21G002R01 snubbers. One of these should be corrected to Q1E21G002R02.
: c. Presently there are two Q1E31G178R08 anubbers. One of these should be corrected to Q1E31G178R09.
;                                        d. Present Elevation 100 for Q1B33G355R01 (2) and Q1B33G318R01 should be Elevation 102.
H7sp2
: e. Elevation for Q1B21G226R01 should be 173' instead of present 172'.
: f. Both of Q1B21G195R02 snubbers are at elevation 160' instead of present 160' for one and 161' for the other.
: 4. Non-Q Mechanical Snubbers are added to the Table as Section 2. Section 1 of the Table is Safety Related Mechanical Snubbers and Table 3.7.4-2 title is changed to
                            " Mechanical Snubbers" to allow the two separate sections.
The Non-Q Mechanical Snubbers are added because even though they are on Non-Q piping this was included in Stress analysis calculations for adjacent Q piping.
: 5. Duplicate listing of snubbers is avoided by listing only one and using parenthesis to indicate the number of snubbers with identical listings.
SIGNIFICANT HAZARDS CONSIDERATION:
The changes proposed to Table 3.7.4-2 are administrative in nature and do not involve changes to the plant. This change does not involve the reduction of safety margins and no significant increase in the probability or consequences of an accident previously evaluated is involved nor is the possibility of a new or different kind of accident from any accident previously evaluated created. Thus the proposed change tc- the Technical Specification does not involve any significant hazards consideration.
I l
l l
l i
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                                                                                                              ,             ss s
                      ,                          1                                                                                                  .
s l                                                                                                                      .
TABLE 3.7.4-2
!                   n                            gepggy WEtM99 MECNAMICAL' $NUBBER$*e * +
!  8                               n    MCAL wgasy1
!  t;;   l. [A M REM W b l  j         SNUBBER                                            1                           SNUBBER
!  -         N3.                            AREA          ELEVATION                        NO.                           AREA        ELEVATION
;  a
   "         a. RECTRCULATION SYSTEM                                                      RECIRCULATION SYSTEN (Continued)-
Q1833G023R01(2)           11             117                              Q1833G112R02                 11          101 l
015? --- : 1              11             117                              Q1833G124R01                 11          122 i
j                 Q1833G024R01              11             102                              Q1833G128C01 (1)             11          121 Q1833G024R02 (2)          11             102                               Q10%12;G1                   11          121 010Z:-' Z                 11             1;^                               Q1833G129C01 11-     x 121 Q1833G024R05              11             101                               Q1833G262902                11          103 Q1833G105C01              11      .      101                               Q1833G265C01                11          102 11             101                               Q1833G265R04                11          107 i
'   g             Q1833G105R01 11             101                               Q1833G265R05                11          112
    .             Q1833G105R02 (1)
       ,          010%;%:--                 11             101                                Q1833G322R01 (C            11          112 11             101                               Q1CLi;R1                    11          112
!   g             Q1833G100C01 11             101 Q1833G331R02              11          111
;                 Q1833G108R01 O)
Qi;^3G;;; ;l              11             11                                 Q1833G337R02               11          109 l                  Q10%;,2;;                11             101                                Q1833G339R01               11          111' 11             101                                 Q1833G346R01              11          105
!                  Q1833G108R02 (4 j                  q;C%1;; ,-                11            101                                 Q1833G355R01(2)           11          400 1o2 i                                                                                   fQ1833G318R01                        11      s   M i01
[qtB33&E11Ra'                         to          10 8
!             a j               Snihbers may be added to safety related systems withaut prior License Amendeont to Table 3.7.4-2 i               provided that a revision to Table 3.7.4-2 is included with the next License Amendment request.
l            M4 The nus,.La v e* *. per<s3ktus                 }c Tie non Lav of sw Wes assocish) wirl.                                          %
                         . . ~ , . . . - ..,,e,r. u
                  ,u                                                      -. u -
                                                                 - . :.r.) -.s y. ,~.,, s. , ,,,
a..      ,s       ,, .n. ~ ~ -                                                                   c.                           y i                                                                                                                                                z i
I
!'                                                                                                                                               h a
                                                                                                                                                 ~
l,
* TABLE 3.7.4-2 (Continued)                               .
E                        -
N MECHANICAL SNUBBERS */ 48                               '
h,   l, SAFtr1 GCLMED Marengnyy- y         g, SNUBBER j       SNUBBER ELEVATION              NO.                    AREA   ELEVATION NO.                      AREA
  -4
* b. MAIN STEAM SYSTEM                                   MAIN STEAM SYSTEM (Continued) l 11      141                    Q1821G023R18 Cs)       11     119 Q1821G021C04 11      135                    ^1001002%10             11    11" Q1B21G022R01 (2)                                                                  131 11      13;               I Q1821G024C01              11 Q10:1000001                                                                11     137 Q1821G022R03 09     11      133                    Q1821G024R04 11      1r                     Q1B21G024R05 (1)       11     132 Q1""10002^^0                                                                     22 11      124                    Q1"^1002005             11 Q1821G022R% (1)                                                                  125 11      120                    Q1821G024R06            11 Q1"210022^00                                                                     119 11      132                    Q1821G024R07 (2)        11 R            Q1821G022R12 CO                                                                  in Q1=1002=12          11      Ir                     Q1 10 0=7               11 11    138 T            Q1821G022R13 (1)    11      131                    Q1821G024R11 11    127 0            Q10210022"13        11      101                    Q1821G024R12 (1)
Q1%;^000G2             11    107 Q1821G022R14        11      126 11      125                    Q1B21G024R13            11    123 Q1821G022R15                                                                     128 11      121                     Q1B21G024R17           11 Q1821G022R16                                                                     128 11      137                     Q1821G025R02           11 Q1821G023R03                                                                     125 11      133                     Q1821G025R03           11 Q1B21G023R05 11      133                     Q1821G025R04 (2)       11    124 Q1821G023R06 (1)                                                           11    120 Q10210020ZO         11      103                     Q1001002000 11      126                     Q1821G025R05           11    120 Q1821G023R08                                                                     143 11      120                     Q1821G026C01 (1)       11 Q1 21000"00 Q1821G023R09        11      122                     Q1^210000001           11    100 143 M
11      122                     Q1821G026CO2 (1)       11 3 ^Q1821G02%10                                                                 11    102   -
Q1B21G023R11 (2)    11      120                     Q1 210020002 Q1"21000"11      --
11 11 120-141 Q1821G026R01 Q1821G026R02 (2) 11 11 143 153 N{
Q1821G023R14 Q1821G023R15 (t)    11      141                     Q:""10000"^^           11 11 12 149 kg Q102 0020"10         11     101                     Q1821G026R03 133                                            11    153         %
Q1821G023R16         11                             Q1821G026R04 (2) 11     121                     v ouoveun-,            it    isa          N Q1821G023R17 fqasus013RIO             it    slo         O

 !       @                                                         TABLE 3.7.4-2 (Continued)
N o                                                    $mPffMttt*Yt9-MECHANICAL SNUBBERS 8!r *
* c)
F  g, ShFCVf MELATED accHAndCCAL CNBBEws m                                                                                SNUBBER E      SNUBBER ELEVATION               NO.                    AREA   ELEVATION 5      NO.                                      AREA
         -4
         "                                                                                MAIN STEAM SYSTEM (Continued)
MAIN STEAM SYSTEM (Continued) 11    170 Q1821G026R05                         11        143                    Q1Z101"Z01 Q1821G149R01 (2)        11    172 Q1821G026R06 D)                      11        143 103                    Q1021010%01             11    17" Q1%10000"^0                         11 11    174 Q1821G026R07                       11        143                    Q1B21G153C01 11    122 Q1821G026R08                       11        149                    Q1821G153C02 Q1821G153CO3 (z)        11    171 Q1B21G030R03 (2)                  8         143 Q121012CO               11    171 Q1"2100%"03                       0         l'2 11    181 11       129                    Q1B21G153R01 R
Q1821G032R03 11       127                     Q1821G153R02 6)        11     175 i                      Q1821G032R04                                                                              11     175 11       127                     Q1021010%00 l        Y             Q1S210032"^0 Q1821G153R03 (2)       11     172 g              Q1B21G032R05 (t)                 11       120 Q1Z1015:00              11    172 Q102100 2^05                     11        120 165                    Q1921G153R05 (2)        11    170 Q1B21G123R01                    11 l
11    170 Q1821G126R01                    11        159                    Q121015:05 Q1821G162R01            11    113 100 Q1"210127"^1                   11
(                                                                                         Q1B21G171R01            11    165 i
Q1821G127R01 (2)               11        193 Q1B21G174C01 G)         11    196 Q1B2.G127R041                 11        186 Q1%10170001             11    1^0 Q1821G139R01                  11        150 11    197 11        150                    Q1821G174P01 Q1821G139R02                                                                           11    1%
173                    Q1821G174ROZ i
Q1821G141R01                  11 Q1821G175R01 (*)        11    153        y Q1821G142R01 (1)             11       173 i
Q1%1017Z01              11     153-       ~
Q1^,210142Z1                 11       173                                                             '
l Q1821G175R02 (2)        11     158 11       173 j
Q1821G144R01 Q1821G146C03 (2)            11       169                    Q1%1017Z00              11 11 1%
152 h'
t;s Q1 210100000               11        10"                   Q1B21G180R01 g
Q1821G180R02 (2)        11     158 11        169 Q1821G146C04      ,
11        173                    Q1%10100^02             11    1%         9
 !                              Q1821G146R03                                                                              161 Q1821G147CO2               11 11 167 173 Q1821G180R03 Q1821G181C01 11 11    158        y' Q1821G148C01 CO l
(QtslIGI63Rol              is     31 3     &
i (qsstic10not               Ie     in3

         -                                                  n                                                       ,
TABLE 3.7.4-2 (Continued)                            ,
1AFEHHIEtMfD NECHANICAL SNUBBERS *e *
* 8 q  g, SAFETY ADATnD McHA+Irena           SMdesEas j     SNUBBER                                                       SNUBBER
     -     NO.                         AREA      ELEVATION               NO.                    AREA      ELEVATION H
* MAIN STEAM SYSTEM (Continued)
MAIN STEAM SYSTEM (Continued) 1
 ,             Q1821G1B3R01 C O        11        152                     Q1821G213R02 CO         11       152
!              Q1021010301             11        102                     Q1B21G217R02            11       159 l              Q1821G183R02            11        151                     Q10210210;0^            11        101
;              Q1821G194R01            11        161                     01%;%;%%                11        101
!              Q1821G194R02 (O         11        159                     Q1821G219R01 (2)        11       157 Q100101^^^0^            11        IZ                      Q1001021%01             11        107 l
i Q1821G195R01            11        161                     Q1821G222R01            11        160            l
. R         Q1821G195R02 C O        11        160                     01B21G224R01            11        152
* 11        101                                             11       147 Q102101^%02                                               Q1821G225R01 T         Q1B21G196R01 (1)        11        151                     Q1821G226C03            11        158 j    U         Q100101^0001            11        101                     Q1821G226R01 (C         11        17M j              Q1821G197R01 (2)        11        157                     Q10210220""1            11        172 Q102101^7^01            11        107                     Q1821G304R01            11        156 l                                               158                                             11        151 j              Q1821G201R01            11                                Q1821G306R01
 !             Q1821G201R02 (1)        11        157                     Q1B21G311R01 (2)        11        152 i              Q10010001"C             11       107                     Q1%10111"^1             11        1%
Q1B21G204R01            11       152                     Q1821G355R01            11        147 Q1821G204R02 ( C        11       160                     Q1821G357C03            11        148 Q102102^^^02            11       100                     Q1821G359CO3            11        148 Q1821G205R01            11       159                     Q1821G361003            11        147 g}g1G372R01 (2)         11                   y j
g182}G2{gR02 (1)
                                        }}
                                                  }gg
__                     s....._.                ..
                                                                                                           }48 Q1821G208R01            11        157                    Q1821G382R02 (*)        11        155 Q1821G208R03            11        160                    ^1%1%; ; ,1             11 11 100 147 g
g
;               Q1B21G210R01 (t)        11        157                    Q1821G423R01 j              Q1Z10010701             11        107                    Q1821G424R01            11        147        2 i               Q1821G213s01            11        151                  A Q1821G490R03            11        152        w
                                                                       %su sssvnot                is       ar2          e l
M^"''"''d'"''            ^^
gqigt Guinos (t)          sg        syy        g L(qsa1\Gs91Roi                 st      IS3                                                                 2 i
v   ,

I i
                                                    @                                                                                     TABLE 3.7.4-2 (Cont.inued)
E                                                                                                                                                         ,
4AFMW-4E4AMG MECHANICAL SNL98ERS*e *
* E q;          j, SAFKtY RELATEb M MM ANNAL. CNug 8 EJt .5
: d.             SNUBILER-                                                                                     SNUBBER 5*
NO.                                                    AREA           ELEVATION               NO.                    AREA      ELEVATION 1
j
* c. SLC SYSTEM                                                                                RESIDUAL HEAT REMOVAL SYSTEM (Continued)
Q1C41G113C02                                         11           195                     Q1E12G010R15           8         103 i                                                                      Q1C41G113C03                                         11           181                     Q1E12G010R16           8         104
!                                                                      Q1C41G113R02                                         11           181                     Q1E12G010R17 (2.)      8         104
:                                                                       Q1C41G113R03                                         11          181                     Q1:1M"10"17            0         Zt Q1C41G117C02                                        11           145                     Q1E12G010R18(2.)       8         %
Q1C41G117ROI                                         11          151                     QZi^0^1^^10            0         ^0 Q1C41G119R01 (23                                      11          129                     Q1E12G011R02 (J)       8         99
                                                                                                                                                                                                  ^^
t'                 Q10010110;;l                                          11         10^                     QZ150112               0
* 11         114                     Q1:120011^,02          0
                                                                                                                                                                                                  ^^
Q1C41G119R03 Y                 Q1C41G119R04                                          11         112                     Q1E12G012R02 (t)       7         114
                                                      @                 Q1C41G119R05                                          11         112                     Q1:120012;;Z           7         110 Q1C41G120C05                                          11         155                     Q1E12G012R04           7         142 i
Q1C41G124R01                                          11         159                     Q1E12G012R05           7         142
;                                                                       Q1C41G124R03                                          11         162                     Q1E12G012R08           8         104 Q1E12G012R09           8         102
;                                                                   d. RESIDUAL HEAT REMOVAL SYSTEM                                                             Q1E12G012R13           7         119 Q1E12G012R15           7         133 Q1E12G009R03                                          7          134                     Q1E12G012R16           7         99 Q1E12G009R04                                          7          134                     Q1E12G012R18           11        133 Q1E12G009R05                                          8          134                     Q1E12G012R19           11        133 Q1E12G009R06                                          8          134                     Q1E12G013C01           7         110
{                                                                       Q1E12G010R02                                          8          105                     Q1E12G013C02           7         130            U Q1E12G010R04                                           8         103                     Q1E12G013R02 C D       7         115            -
Q1E12G010R05                                           8         125                     QZ12001;;;-            7         110 Q1E12G010R07                                           8         133                     Q1E12G013R03           7         110 Q1E12G010R10                                .          8         142                     Q1E12G013R04           7   ,
119 Q1E12G010R11                                            8        142                     Q1E12G013R05 (t)       7         100           2 Q1E12G010R13 (2)                                        8        113                     QZi;;;;;;;,            7         l'0           W
  ;                                                                     s......,                                                .        . . . ,
Q1E12G013R06 (3)       7         320 f
V         ,
:                                                                                                                                                                                                             v          ,
l

TAklE3.7.4-2(Continued)                                .
l z                                                                                                                   /
                                     -6AFEPHIEMTEG MECHANICAL SNUBBERS */ * *                    ,
8 G; f, SAfkTY AMATED nrcHAntess. guagggyg                        SNUBBER SNUB 8ER h
-     NO.
AREA       ELEVATION            NO.                     AREA    ELEVATION
-e
* RESIDUAL HEAT REMOVAL SYSTEM (Continued)
RESIDUAL NEAT REMOVAL SYSTEM (Continued) 11      214 Q1:12001C;;           7          120                  Q1E12G015R19 11      144 Q1 12001000           7          IC                    Q1E12G015R20 140 121                   Q1E12G015R21 (s)       11 Q1E12G013R07         7 105                   Q1:12001221            11       l'0 Q1E12G013R08         7 11       192 7          97                    Q1E12G015R28 O)
Q1E12G013R11 110                   ^ CC012C               11       1C Q1E12G014C01         8                                                                1^^-
8          106                   Q1C"^012C              C Q1E12G014C03                                                                 11       205 T                               8          130                   Q1E12G015R33 (1)
* Q1E12G014C04 129                   ^1CC^12C               C        ZZ Q1E12G014R01 (t)      8                                                               157 12^                   Q1E12G015R38           11 T          Q1000^1 C 1           0 11       143 0          Q1E12G014R03 (1)      8         98                    Q1E12G016C01 11       146
           ^1:1^O01 C O          O         ^0                    Q1E12G016R01 11       143 Q1E12G014R04 (3)      8         122                   Q1E12G016R02 11       143
           ^ C 2001 C ^          0         IC                    Q1E12G016R03 143 IC                    Q1E12G016R05 Cs)       11 Q1C 001 C "           0 105                   ^1:1^0010:::           11       l'2 Q1E12G014R05          8                                                               139 106                   Q1E12G019R05 (C        8 Q1E12G014R07          8                                                               IC 109                   01CC01^^C              0 Q1E12G014R10 (*)      8                                                               149 8
Q1:12001^.10
                       ^
0          10-                  Q1E12G019R07 149 Q1E12G019R08            7 Q1E12G014R11 (9       8          110 7       143          y
           ^1:120^1""C           0          C0                   Q1E12G019R09 (3.)
7       l'2            -
11         156                  01:1^C00^^^                                  '
Q1E12G015R02 Q1E12G020R01 (t)        8       148 -
11         143 Q1E12G015R04 CS) 1"2                  Q1:1^00C21              0        00 01:100""*"            11 11        143                  Q1E12G020R02 (2)        7       148          S[
Q1E12G015R06                                         Q C;CGC-                7        1C          2 11        214                                                               %
Q1E12G015R07                                                                 8        148 11        210                   Q1E12G020R03                                  I Q1E12G015R08                                                                 8        148 11        143                   Q1E12G020R04 (t)
Q1E12G015R11                                          ^ 1C ^00 -^ = "        0        l'O 11        210 Q1E12G015R17

f c3 h
* 8 g                                                                                         TABLE 3.7.4-2 (Continued) 4AFEfV-#Et*Tf9-MECHANICAL SNUB 8ERS*e W #
l, (AF[rY MELATEb MEC HA N TCAL Mn!EEEAC j                                                -                                 SNUB 8ER                                                                           SNUB 8ER NO.                                          AREA      ELEVATION                   NO.                     AREA   ELEVATION l
:                                                                                  RESIDUAL HEAT REMOVAL SYSTEM (Continued)                                        f. HPCS SYSTEM Q1E12G020R05                            7         147                         Q1E22G001R10 (2)        8      96 l                                                                                       Q1E12G020R07 (t)                        7         147                         ^1;ZOC l 1"             0      %
;                                                                                       Q1:100^0^Z7                             7         1"7                         Q1E22G002R02 (2)        8      96 Q1E12G020R09                            7         147                         Q1:22000ZZ              0      ^0 l
i                                                                                       Q1E12G021R01                            8         147                         Q1E22G002R03            8      %
:                                                                                       Q1E12G021R03 (2)                        8         146                         Q1E22G003R01            11     153 I                                                                                                                               0         115                         Q1E22G003R02            11     153 M                                                                             8         110                                                 11     149 l
)
                                                    *                              [Q1:120001"^1 Q1E12GC25R01 Q1E12G119R02                            7         152 Q1E22G003R03 Q1E22G00?.9%            11     150
                                                    ?                                   Q1E12G159R01                            7         126                         Q1E22G003R05            11     151
,                                                  M                                    Q1E12G159R03                            7         126 Q1E12G159R04                            7         131                                                                    l
[Q1E12G025C01 ( O                        8         95                       g. RCS LEAK DETECTION SYSTEM
: e. LPCS SYSTEM                                                                  Q1E31G116R01            11     169 Q1E31G122R01 (2-)       11     149
:                                                                                       Q1E21G001R05                            9         96                          Q1E31G124R01 (1)        11     151 i
Q1E21G001R07 (0                         9         96                          Q1C1012^Z1              11     151
                                                                                                                                          ^'                                                  11     149 Q1E210^^1"^7                            ?                                    Q1E31G126C01 1
11       150                                                 11     159           w Q1E21G002R01                                                                 Q1E31G140R01 i
Q1E21G002RQT1                           11       150                         Q1E31G140R02 (O         11     159           ?
l                                                                                        Q1E21G002R03                            11       151                         ;10101"^^^^             11     15"           A Q1E21G002R04                            11       153                         Q1E31G148R01 (2)        11     151 Q1E21G002R05                            11       153                         ^1CZi"^Z1               11     Z1            z m
l                                                                                        Q1E21G002R06                            11       153                         Q1E31G149R01 G)         11     151 Q1E21G002R07                            11       150 I
Q1CZi"^Z1               11 Z1            kN i
* v i.
M

        .-                                                     m                                                  ,
1 l
i i
y                                                TABLE 3.7.4-2 (Continued)                          ,
z i                         .                  5AFEfW-REtAfE9 NECHANICAL SNUB 8ERS*/ * *         ,
l  E
;  q   p. Werv netArta s,ecnnnrou                       cwe een s SNUBBER j       SNUB 8ER NO.                     AREA  ELEVATION
 ! -       NO.                          AREA      ELEVATION g  -n
   "                                                                      RCIC SYSTEM (Continued)
}          RCS LEAK DETECTION SYSTEM (Continued) 11        158                     Q1E51G001R10 ( O        11     134 Q1E31G168R01 11         151                    010 Z C1"10             11     1%
Q1E31G174R01 (4                                                                          178 11         Z1                     Q1E51G001R15            11
,               Q112^17C01                                                                               190 11         147                    Q1E51G001R17 (C         11
:               Q1E31G176C01 11          179                   Q1010CC17               11     1^"
l               Q1E31G178R08 11          179                   Q1E51G001R18            11     194 l               Q1E31G178R079 11          156                   Q1E51G001R19 (2)        11     194
;               Q1E31G181R01 Q10100101^             11     1^"
1 R*
Q1E31G243R01            11          144 7      126 Q1E31G243R02            11          140                    Q1E51G003R03 7      117
:   T           Q1E31G246R01 (23        11          144                    Q1E51G003R04 7      127
!   E3          Q10100'0""1              11         l"                     Q1E51G003R05 (2) l                                                                          alC=CCC                7      27 Q1E51G003R07           8      112
!           h. MSIV LEAKAGE CONTROL SYSTEM Q1E51G003R08 (0         8     112 l                                                                                                        110 8          122                    Q1CIC C "^^            0 l               Q1E32G103C01 (2)                                                                         109 0          20                     Q1E51G003R09 (4         8 Q1C"010001                                                                        ^
1^^
!                Q1E32G106C01            8          121                    Q1CZ^^ZC 4
8          122                    Q1E51G003R10            8     105 Q1E32G109C01                                                                            100 8          148                    Q1E51G003R11 (C         8 Q1E32G119C01                                                                             1^^
Q10100 C11              0 i
8      106       W i           1. FEEDWATER LEAKAGE CONTROL SYSTEM                          Q1E51G003R12 (0 Q1E51G004CO2 (0         8      97        7 Q1E38G102R01            8           145                   Q15 Z C"-^^             0      07 .      m Q1E51G004R01 D )        8      98        h
                                                                                                          ^^
D
: j. RCIC SYSTEN                                               Q15 Z^^""^1              0 8     106       %
Q1E51G004R05 00
* 8          104                   Q1CICC""^5               0     20 Q1E51G001R05                                                                             96 8          109                   Q1E51G004R06 D-)         8 Q1E51G001R06                                                                             30 11         133                   Q1CIC^""^0               0 l
Q1E51G001R09                                                                                       N i                       -
Q i
i

1 4
!                                                                                                                                            TABLE 3.7.4-2 (Continued)
'                       l z
                                                                                                                                    --6AFEfV-ret *TE& MECHANICAL SNUB 8ERS*/ *
* o c                                                                                                                                                                                                              .
1                       G; e
I C4FEry ACLATED W CHANCO6L SMU8ggkg                                                                                                                                                         l
,                       c      SNUB 8ER                                                                                                                            SNUB 8ER                                             l l                            NO.                                                -
AREA       ELEVATION                NO.                               AREA   ELEVATION 1                       w                                                                                                                                                                                                              t RCIC SYSTEM (Continued)                                                                                                             RWCU SYSTEM (Continued) i                                                                      Q1E51G004R07                                       (2)  8          97                       Q1G33G002R10 (1)                  11     102 yam.....,                                              _          ..                       , ,. ,.._
!                                                                       Q1E51G004R08 (1)                                       11         164                      Q1G33G002R11                      11     102 l                                                                      QZ51000""00                                            11         10'                      Q1G33G002R12                      11     102 Q1E51G004R11                                           8          97                       Q1G33G002R13 (D                   11     102 Q1E51G004R13 (2.)                                      11         167                      ;10 %^^^^10                       11     1^2
                                                                  - Q101000^Z                                                   11        Z7                       Q1G33G002R14 (2)                  11     102 f                       M
* Q1E51G004R14 (O                                        11         152                      ';Z2'^00^^1'                      11     20                         '
!                                                                        Q15 Z00'"I'                                            n         50                       Q1G33G002R16                      11     112 l                       T                                                Q1E51G158R03 (1)                                       11        143                      Q1G33G002R17(O                    8      125 i                       5-                                               Q13101'^CO                                             11        l'O                      01C2^^^C17 105 Q1G33G002R18                      8      116
: k.                                      CopBUSTIBLE GAS CONTROL SYSTEM                                                           Q1G33G002R19                      8      116 Q1G33G002R21 (C                   11     102
,                                                                         Q1E61G001R07                                          11        189                      Q10%^0GC                          11     100 Q1G33G002R22                      11     102
: 1.                                     RWCU SYSTEM                                                                              Q1G33G002R24                      11     801 Q1G33G011R01                      11     140 l                                                                         Q1G33G002CO3                                    (2.1   11       113                      Q1G33G011R03 (E)                  11     145
                                                                           -.-, ,-- --,                                          ,,       ,,,                                                        ,,     ,4.
?
y au J Jiavva. mav e                                  aa        ..w
                                                                                                                                                                   -.,...,,-,,-.y yausesuvasvtv                     aa     aTw I                                                                          Q1033G002R03 (4                                      8         136                      Q1G33G012R01 (C                   11     142
                                                                        -410:00002:00                                           0          100                     Q10:20010001                      11     l'O Q1G33G002R05 (O                                       11       140                      Q1G33G012R02                      11     152 103 d
j                                                                     '"1C'002"^5                                                11        10"                     Q1G33G105R01(N                    11                        -
Q1G33G002R08 (2.)                                    11        102                     Q1CZ10ZO;                          11     1C               rs
;                                                                           q = ==0 ==                                           11        =0        -
q1====n                           n   -
1=               tb h
:                                                                           Q1G33G002R09 (3)                                     11        102 Q10r0^^==                                            n         =0                                                                                  2
{                                                                           ';r 20^^r^^                                          n         1^2                                                                                  ?
;                                                                                                                                                                                                                               w
:                                                                                                                                                                                                                               N I                                                                                                                                                                                                                          Cs i
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m E
     =
     =
                              .                      TABLE 3.7.4-2 (Continued)                  ,
c h                                        -6AFEPHMMATE& MECHANICAL SNUB 8ERS*/ W*
I    C
'    5
     -e f.$AFzr1 RtlATED MECHAMECAL SNOB 6 Ens _
_,    SNUB 8ER                                                        SNUBBER NO.                           AREA      ELEVATION               NO.                    AREA E_LEVATION l          m. FPCC SYSTEM                                                SSW SYSTEM (Continued) 9         114                       17^10CC10            0    1^0 Q1G41G006R01 7         99                      Q1P41G002R12 (23       8    106
;               Q1G41G006R07 (3) i               Q10'10000^^7             7
                                                   ^^
                                                                           ^17010GC 2             0    1^0
                                         '         ^^
Q1P41G006C01           8    99 Q10'10^0'"07 Q1G41G015R09             11        204                     Q1P41G006C17           8    91 11        163                     Q1P41GG37R19           025A   /84Y i               Q1G41G016C08 j               Q1G41G016R04             11        166                     Q1P41G007R20           025A   / 84 9
!     w                                  11        163                     Q1P41G007R23 & >       025A    s/IJF 1         Q1G41G016R24 Q1G41G016R27        (2)  11        203                     Q17'1^^^7"C            ^ ^C 11        000                     Q1P41G007R24 (2)       025A   / /J7 y         Q10010010"07
                                                                                                  ^ ^C y         Q1G41G016R28 (17          11       206                      ^1""10^^7"^^
i s..........,              ..        b..
Q1G41G016R32              11        197                c. CCW SYSTEM l'
Q1G41G018R06              9         197 Q1P42G002R06(1)       9    193
                                                                                                  ^
I
: n. SSW SYSTEM                                                 Q1""C^0C00                 IC l                                                                           Q1P42G002R07 (I-)     9    186
                                                                                                  ^
7         98                      Q1"'C^^^^07                100 Q1P41G001R14 (C
 ;               ^1^01000=10              7          ^0                     Q1P42G002R11(1)       9    186
                                                                                                  ^
                 ^17 ^ ^ 0^.,0CC          0          %                      Q1.^0^000011               100          y i
j                                        8          106                    Q1P42G002R13(C         9    186         s Q1P41G002R10 U.)                                           .... .....             .    .--         .
eF   4%#%#
33 7 kwww b 5%.e#
1                                                                                                                   S w
i                                                                                                                  D 1                                                                                                                  N b                                                                                                                  Q

TABLE 3.7.4-2 (Continu:d) h4M-f2f)      .
MECHANICAL SNUBBERS '
: 2.        NON-Q MECHANICAL SNUBBERS SNUBBER NUMBER                       AREA       ELEVATION                      SNdBBER NLHBER       AREA      ELEVATION
: a. MAIN STEAM SYSTEM N1B33G108C02         11        101 N1B21G118R01                         11         148                            N1B33G108R03 (2)     11        101 N1B21G118RG2                         11         147                            N1B33G108R05         11        101 N1B21G191C02                         11         137                            N1B33G108R06 (2)     11        101 N1B21G192CO3                         11         136                            N1B33G108R07         11        101 N1B21G193R01 (2)                     11         138                            N1B33G119R04         11        112 N1B21G193R04                         11         136                            N1B33G120R03          11       101 N1B21G231R01 (2)                     11         163                            N1B33G123C01         11        102 N1B33G362R03          11       102
: h. RECIRCULATION SYSTEM                                                    c. RESIDUAL HEAT REMOVAL SYSTEM NIE12G172R02          11       129 N1B33G104R02                         11         102                           NIE12G212R01          11       136 N1B33G105C01                         11         101                           NIE12G212R03          11       133 N1B33G105C03                         11         101 N1B33G105C04                         11         101 N1B33G105C05                         11         101 N1B33G10%01                          11         101 N1B33G106R01                         11         102 NIB 33G107R01                        11         102 N1B33G107R02                         11         102 H7sp4

                                            -                                                                    e TABLE 3.7.4-2 (Continu:d)
MECHANICAL SNUBBERS '
: 2. NON-Q MECHANICAL SNUBBERS AREA      ELEVATION                      SNUBBER NUMBER       AREA     ELEVATION SNUBBER NUMBER
: d. STANDBY DIESEL GEN. SYSTEM                                    REACTOR WATER CLEANUP SYSTEM NIG33G002R05 (2)     11       147 12        165                            NIG33G002R08 (2)     11       164 N1P75G004R01 12        152                            NIG33G002R10 (2)     11       147 N1P75G004R02 (2) 12        152                            NIG33G002R11 (3)     11       180 N1P75G004R03 (2) 12       165                            NIG33G002R12 (3)     11       180 N1P75G005R01 12       152                            NIG33G002R13         11       178 N1P75G005R02 (2) 12       152                            NIG33G002R14           8      120 N1P75G005R03 (2)
NIG33G002R21           8      120
: e. REACTOR CORE ISOLATING COOLING SYSTEM NIE51G120R01         11        127
: f. REACTOR WATER CLEANUP SYSTEM NIG33G002R01           7       120 N1G33G002R02           8       118 NIG33G002R03           8       123 NIG33G002R04           8       123 H7sp5

I
  +
: 32. (GGNS - X41)

==SUBJECT:==
Technical Specification 3.1.3.2 and 3.1.3.3, page 3/4 1-7 and 3/4 1-8.
DISCUSSION:               The su*sject technical specifications address the maximum scram insertion times and the scram accumulator operability requirements for OPERATING COND1TIONS 1 and 2          The proposed change exempts these specifications from the applicability of 3.0.4. The change consists of adding a note d to the ACTION statement of 3.1.3.2 and a note c to the ACTION statement of 3.1.3.3 stating: The provisions of 3.C.4 do not apply.
JUSTIFICATION: As noted in the Bases for 3.0.4, exceptions to 3.0.4 can be provided when startup with inoperable equipment would not affect plant safety. The Bases further provide that such exemptions will be noted in the ACTION stetements of the appropriate specifications. Startup can i.e safely accomplished without the full complement of accumulator; and with rod insertion times greater than the maximum limits of these specifications.
This exemption has been added to these specifications in the Standard Technical Specifications, Revision 3 (draft).
SIGNIFICANT HAZARDS CONSIDERATION:
The proposed changes do not significantly increase the probability or consequences of an accident nor do they create the possibility of a new or different accident from those previously evaluated. They do not constitute a significant hazards consideration.
G91sp4

y2., ( Gs N S - X4 0 REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
ACTION:           (Co,ntinued) l     b.           With a " slow" control rod (s) not satisfying ACTION a.1, abo've:
: 1. Declare the " slow" control rod (s) inoperable, and
: 2. Pdrform the Surveillance Requirements of Specification 4.1.3.2.c at least once per 60 days when operation is continued with three or more " slow" control rods declared inoperable.
Otherwise, be in at least HOT SHUTDOWN within 12 hours.
: c.           With the maximum scram insertion time of one or more control rods exceeding the maximum scram insertion time limits of Specification 3.1.3.2 as deter-mined by Specification 4.1.3.2.c, operation may continue provided that:
: 1.    " Slow" control rods, i.e., those which exceed the limits of Specification 3.1.3.2, do not make up more than 20% of the 10% sample of control rods tested.
: 2. Each of these " slow" control rods satisfies the limits of ACTION a.1.
: 3. The eight adjacent control rods surrounding each " slow" control rod are:
a)     Demonstrated through measurement within 12 hours to satisfy the maximum scram insertion time limits of Specification 3.1.3.2, and b)     OPERABLE.
: 4. The total number of " slow" control rods, as determined by                 '
Specification 3.1.3.2.c, when added to the sum of ACTION a.3, as determined by Specification 4.1.3.2.a and b, does not exceed 7.
Otherwise, be in at least HOT SHUTDOWN within 12 hours.
SUR$EILLANCEREQUIREMENTS                   ^ d.        %* F*M **S *S S? **Ib' N N, 0 cab Y 4.1.3.2 The maximum insertion time of the control rods shall be demonstrated through measurement with reactor coolant pressure greater than or equal to 950 psig and, during single control rod scram time tests, the control rod drive pumps isolated from the accumulators:
: a. For all control rods prior to THERMAL POWER exceeding 40% of RATED     %
THERMAL POWER following CORE ALTERATIONS
* or after a reactor shutdown o that is greater than 120 days,
: b. For specifically affected individual control rods following            ~
maintenance en or modification to the control rod or control rod       C drive system which could affect the scram insertion time of those
!                         specific control rods, and c.. For at least 10% of the control rods, on a rotating basis, at least once per 120 days of POWER OPERATION.
       "Except movement of SRM, IRM, or special removable detectors or normal control            y rod movement.                                                                           o u                        r-e The prov.m.ns of Spch h 4.oA o re rM o.yp cable                       .{.re.nkryin4.
OPERATtoNAL couwTteN 2 provida! Ais surve'illue.a is carwilded 96.c                 =
do OY ink OPL9tMloggt CoMbtTiou 1.                                                    I GRAND GULF-UNIT 1                                3/4 1-7                                  p

D,(GGNs M1)
REACTIVITY CCNTROL SYSTEMS CONTROL ROD SCRAM ACCUMULATORS trMITING CONDITION FOR OPERATION 3.1.3.3 All control rod scram accumulators shall be OPERABLE:
APPLICABILITY:        OPERATIONAL CONDITIONS 1, 2 and 5*.
ACTION:
: a.       In OPERATIONAL CONDITIONS 1 and 2:
: 1. With one control rod scram accumulator inoperable, within 8 hours:
a)   Restore the inoperable accumulator to OPERABLE status, or b)   Declare the control rod associated with the inoperable accumulator inoperable.
Otherwise, be in at least HOT SHUTDOWN within the next 12 hours.
: 2. With more than one control rod scram accumulator inoperable',
declare the associated control rods inoperable and:
a)   If the control rod associated with any inoperable scram accumulator is withdrawn, immediately verify that at least one control rod drive pump is operating by inserting at least one withdrawn control rod at least one notch or place the reactor mode switch in the Shutdown position.
b)   Insert the inoperable control rods and disarm the associated directional control valves either:
: 1)    Electrically, or
: 2)    Hydraulically by closing the drive water and exhaust water isolation valves.
Otherwise, be in at least HOT SHUTDOWN within 12 hours,
: b.      In OPERATIONAL CONDITION 5*:
: 1. With one withdrawn control rod with its associated scram accumulator inoperable, insert the affected control rod and l                             disarm the associated directional control valves within one
!                              hour, either:
a)   Electrically, or b)    Hydraulically by closing the drive water and exhaust water isolation valves.
: 2. With more than one withdrawn control rod with the assoc'iated l                             scram accumulator inoperable or with no control rod drive pump l                             operating, immediately place the reactor mode switch in the Shutdown position.
;              C. Tke "At least        the prcNeens accumulatorc6   associated     Spechhn  3.0.4 with eachare. ret a control rod. p withdrawn     Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
GRAND GULF-UNIT 1                                 3/4 1-8

                                                                                   =-- .-             _   _ - _             . - .           . _ - -
Y   h 33.-(CGNS - 826, 176)                                                                                                                       ,

==SUBJECT:==
Technical Specification 3.7.8, Table 3.7.8-1, page 3/4 7-43.
I DISCUSSION:                     The subject Technical Specification provides the limiting conditions for operation relative to area temperatures.                                  It's APPLICABILITY is stated to be whenever the equipment in an affected area is required to be operable. In view of the intended concern with equipment qualification the temperature limit for the control room was set too low. The 77'F Ifmit should be changed to 90*F.
The' proposed change also deletes the " Equipment Not Operating" column and the heading " Equipment Operating".
l JUSTIFICATION: The 77'F limit is derived from human factors considerations.
NUREG-0700, Guidelines for Control Room Design Reviews, requires that the control room HVAC system bet capable of maintaining the dry bulk temperature between 73*F and 77'F.
This is a system performance standard for maintaining the comfort zone for personnel control room occupancy. Minor fluctuations in temperature above 77'F can occur periodically due to changes in the climate control of the control room.
!                                           The proposed limit of 90*F is based upon a review of the control room equipment qualification data sheets. The lowest environmental qualification temperature of any equipment in the Control Room was found to be 90*F.                 Control Room temperatures exceeding this limit for more than eight hours would require an
                                           . evaluation of the impact on the qualified life of the affected equipment as required by the technical specification ACTION statement. The use of the lowest qualification temperature for                                            i this area is consistent with the limits established for other i                                            areas.                Since the existing control room technical specification i                                            limit of 77*F is based on a human factors performance standard i                                            and not an equipment performance standard, as is the intent of Table 3.7.8-1, a change to 90'F is fully justified.
The Standard Technical Specification addresses only the temperature Ifmit for each area and does not address separate temperature limits for equipment not operating and for equipment operating. The proposed change deletes the present 4
area temperature limits when equipment is not operating since this may not be the limiting temperature for an area. The change to a single temperature limit will eliminate confusing requirements.
SIGNIFICANT HAZARD CONSIDERATIONS:
The proposed temperature limit is established consistent with other limits for other areas. And, while the proposed limit the lowest qualification temperature for equipment in the area, the temperature is not expected to exceed this limit and equipment operability is not impacted. The deletion of the 1
G81spl.
_-_   a.-.__ _ _ . . . . _ _ _ _ _ . . . . _ . _ _ _ _ _ _ - . _ . . __._ _                              .. ._ _ ... _- . . _ _ . . . _ -

d'
                             " Equipment Not Operating" column and the heading " Equipment Operating" is purely administrative in nature since it involves conforming present Technical Specifications with the Standard Technical Specifications. The change does not significantly increase the probability or consequences of an accident previously evaluated nor does it create the possibility of a new or different accident from those previously evaluated. It does not constitute a significant hazards consideration.
i 4
i i
I i
G81spl.1

J), (GGNS -82b> I?S TABLE 3.7.8-1 AREA TEMPERATURE MONITORING AREA                                  TEMPERATURE LIMIT (*F)
                                                  ;Q;.;;7;-;;m                         ;;;,;7;;;=
NTOPERATINh                           WeRW?tte
: a. Containment Inside Drywell                                   135                                    150 CRD Cavity                                       135                                    185 Outside Drywell                                   80                                    105 Steam Tunnel                                     12                                     125
: b. Auxiliary Building General                                             4                                   104 ECCS Rooms                                          5                                   150 ESF Electrical Rooms                             10                                     104
: c. Control Buildina ESF Switchgear and Battery Rooms                 104                                    104 Control Room                                     77                                      % 9
: d. Diesel Generator Rooms                           125                                    125
: e. SSW Pumphouse                                    10 ''
* 104*
    *For this area, the limit shall be the greater of 104*F or outside ambient temperature plus 20*F, not to exceed 122*F for greater than one hour.
l GRAND GULF-UNIT 1                   3/4 7-43
                                                   - -- -    -m.  - ,,,, ,, --    n-.-   --,,.-n      - , - - - - - - - , , , . . . ,

e
: 34. Deleted.
I i
l l
l I
l r
l l
l i
G79sp4

se,
                       ~
                   '35. (GCNS - X45).

==SUBJECT:==
-      Technical Specification Figure 6.2.2-1, page 6-4.
DISCUSSION:   . The major changes to the Unit Organization Chart, Figure
                                  - 6.2.2-1, are summarized as follows:
1,   The Site Manager and Assistant Site Manager have been deleted.
: 2. Three Assistant Plant Managers replace the present Assistant Plant Manager Nuclear and the Nuclear Support Manager.
: 3. The new Assistant Plant Manager Support will have reporting.to him; Compliance, Office Services, Security.
Training, and Accounting.
: 4. Reporting to the new Assistant Plant Manager Operations is Operations, Technical Support, Chemistry / Radiological Controls, and the Independent Safety Coordinator.
: 5. The new Assistant Plant Manager Maintenance will have reporting to him; Outage' Planning and Scheduling, Maintenance, and Contracts.
: 6. The Maintenance Engineering Supervisor reports to the maintenance Superintendent instead of the present
              '!_                       ' Technical Support Superintendent.
: 7. A Computer Supervisor has been added to report to the Technical Support Superintendent.
The Licensing Engineer now reports to the Compliance
                                                            ~
8.;
Superintendent instead of the present Technical Support Superintendent.
: 9. The present-Startup Manager has been deleted.
: 10. The three Assistant Plant Managers are required to have a
                 '                        SRO or to have been certified SRO for a plant of this -
type.
: 11. 'A new note has been added to the bottom of Figure 6.2.2-1 to state that changes to Unit Organization can be made without prior Technical Specification approval.
JUSTIFICATION: .The changes to the Unit Organization chart are proposed to more effectively utilize personnel; resources in a growing staff organization. Qualified individuals are being utilized to provide necessary management direction and accept input from H10 spi

O their staff. Positions requiring a license or certification have not been reduced by this change. Deletions are made where positions are no longer utilized, needed or reported to the plant staff. The new note on Figure 6.2.2-1 allows changes to Unit Organization without prior Technical Specification approval. This change allows flexibility in management organization at Grand Gulf.
SIGNIFICANT HAZARDS CONSIDERATION:
The proposed changes to the Unit Organization chart are made to adapt to a growing plant staff. Job qualification requirements are not relaxed by this proposed change. This change does not involve the reduction of safety margins and no significant increase in the probability or consequences of an accident previously evaluated is involved nor is the possibility of a new or different kind of accident from any accident previously evaluated created. Thus the proposed change to the Technical Specification does not involve any significant hazards consideration.
H10sp2

r.
o Unit Organization
             .                                                                               Sin OA           ~ ~ ~ ~ ~ ~
NUCLEAR g                                                                     REPRESEsTArrvt3                                       PLANT MANAGER E
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V FIGURE 6.2.2-s HANGCs ~To U NsT OG A NIZATscN can se espe                                                                                    sus 7HouT PhnoA 8
TccHNsc.AL. SPECIFicNisotu APPMcVAL.,}}