ML20116D147

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Proposed Tech Specs Consisting of Proposed Change,Revising TS Per Generic Ltr 88-16, Removal of Cycle Specific Parameter Limits from TS, That Will Change Each Fuel Cycle
ML20116D147
Person / Time
Site: Grand Gulf Entergy icon.png
Issue date: 10/29/1992
From:
ENTERGY OPERATIONS, INC.
To:
Shared Package
ML20116D128 List:
References
GL-88-16, NUDOCS 9211050207
Download: ML20116D147 (63)
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Text

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SUBJECT:

Generic Letter 8816 - Removal of Cycle Specific Parameters from Technical Spacifications , Technical Specifications: index,1.7a (new), 2.1.2, 3/4.2.1, 3/4.2.3, 3/4.2.4, 5.3.1, 6.9.1 and bases. Affected Pages: Index pages i,iv and xx. TS pages 12,21,3/421 through 3/4 2-7c, 5 5, 619, B 3/4 21, B 3/4 2-7. B. DISCUSSION: Cycle specific parameters limits are evaluated every reload fuel cycle, which can result in - the need for licenso amendments to the Grand Gulf Unit 1 Technical Specifications (TS). 'When the methodology for determining these limits is documented in an NRC-approved Topical Report or in a plant specific submittal, the NRC review of the proposed TS changes is primarily limited to confirmation that the updated limits are calculated using an NRC-approved methodology and is consistent with all applicable limits of the safety analysis. For these situations, the NRC determined that this process is an unnecessary burden on -licensee and NRC resources. To Jimplify this process, Generic Letter (GL) 8816 (Reference 1) provides specific _ guldence to remove the cycle specific parameter limits from the TS. This proposed _ amendment is consistent with GL 8816 and revises the necessary specifications to remove the cycle specific information such as numerical values (or representation by curves) for certain core and_ fuel bundle type dependent power distribution limits in place of the actual numerical value or curve, each specification references a formal report called the Core Operating Limits Report (COLR). The COLR contains the cycle-specific limits . developed using methods previously' reviewed and approved by the NRC. The COLR will be provided to the NRC for each reload cycle in accordance~with the reporting requirements proposed in'this request. This proposed amendment to the Grand Gulf Nuclear Station (GGNS) Technical Specifications (TS) requests changes to the TS index; Specifications 1.0, Definitions: - 2.1, Safety Limits: 3/4.2, Power Distribution Limits and bases: 6.3, Design Features; and 6.9, Reporting Requirements. 'The proposed changes are described as follows:

1) A new specification is added after 1.7 as 1.7a. Specification 1.7a provides the definition of the CORE OPERATING LIMITS REPORT (COLR).

2)l The word "both" is deleted from specification 2.1.2. This is an editorial change.

3) Limiting Condition for Operation (LCO) 3.2.1 and its Action statement are reworded to

- reference the Core Operating Limits Report for the specific Average Planar Linear Heat Generation Rate (APLHGR) limits. Figure 3.2.1-1 is deleted. .t h m

b ,,? Grand Gulf Nuc! car Station Unit 1 to PCOL 92/07 GNRO 92/00093 Page 2 of 6

4) Limiting Condition for Operation (LCO) 3.2.3 and its Action statement are reworded to reference the Core Operating Limits Report for the specific Minimum Critical Power Ratio (MCPR) limits. Figures 3.2.31, 3.2.3-2 and 3.2.3 3 are deleted.
5) ' Limiting Condition for Operation (LCO) 3.2.4 is reworded to reference the Core Operating Limita Report for the specific Linear Heat Generation Rate (LHGR) limits.

The 3.2.4 Action statement is revised to delete the source of the limits. Figures 3.2.4-1, 3.2.4 2 and 3.2.4-3 are deleted.

6). Section 3/4.2 page numbers and the index are revised to be consistent with the number of pages needed to contain specifications 3/4.2.1, 3/4.2.2, 3/4.2.3 and 3/4.2.4 as follows. This is an editorial change.

3/4.2.1 Page 3/4 21 3/4.2.2 Page 3/4 2 2 3/4.2.3 Page 3/4 2 3 3/4.2 4 Page 3/4 2-4

7) The bases of Specification 3/4.2.1 and 3/4.2.4 are revised to be rnore general and to reflect the application of Generic Letter 88-16.

. 8) The fuel assemblies design features in TS 5.3.1 is revised consistent with the use of a Core Operatint Limits Report.

9) A new specification (6.9.1.11) is added to section 6.9.1, Routine Reports, that describes the administrative requirements for the Core Operating Limits Report.
10) The index is revised to add the CORE OPERATING LIMITS REPORT subsection and page numbers are revised /added to reflect changes described in items 1 and 8 above.

This is an editorial change. C. JUSTIFICATION: -The proposed amendment removes the values of the cycle-specific parameter limits from the Grand Gulf Technical Specifications and makes related administrative and format changes based on guidance provided in Generic Letter (GL) 88-16. As stated in GL 8816, three separate actions are needed to modify the Technical Specifications (TS):

  • -the addition of the definition of a named formal report that includes the values of cycle specific parameter limits established using an NRC-approved methodology and consistent with all applicable limits of the safety analysis,
  • the addition of an administrative reporting requirement to submit the formal report on cycle-specific parameter limits to the Commission for information and,
  • the modification of the individual TS to note that cycle-specific parameters shall be maintained within the limits provided in the defined formal report.

7 7 Grand Gulf Nuclear Station Unit 1 to ~* PCOL 92/07 GNRO 92/00093 Page 3 of 6 The cycle specific information from the applicable TS is assigned to separate entries within the COLRi Controlled copies of the COLR will be maintained at GGNS and revised in -accordance with the requirements for future GGNS fuel cycles. Furthermore, the submittal of the COLR to the Commission will allow the Staff to continue to trend and review the values of these limits as stated in GL 8816. 'The cycle specific parameters to be located in the Core Operating Limits Report are as follows: Avernoe Planar Lingar Heat Generation Rate (APLHGR) Limits The Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) power distri-c L bution limit curves are provided to address all fuel types in the core. These limits are determined by considering both thermal-hydraulic and emergency core cooling system performance. New limit values may have to be added to TS as now fuel designs are utilized.- The Singla Loop Operation (SLO) multiplier can be revised based on analyses performed each cycle. To eliminate licanse amendment requests to update MAPLHGR limits, the MAPLHGR curves and SLO multiplier should be relocated to the COLR. L Minimum Critical Power Ratio (MCPR) Limits - L MCPR limits are required to prevent exceeding fuel design and safety criteria in the event cf Anticipated Operational Occurrences (AOO). The MCPR limits are determined for various power / flow / exposure conditions throughout a given cycle. The ' applicability of these MCPR limits for specific fuel designs and core configurations is evaluated for each fuel cycle [ References 3 and 41. Because of the potential for cycle to cycle variations in these limits, it is appropriate to remove them from the TS and . relocate them to the COLR.- Linear Heat Generation Rate (LHGR) The LHGR limits are fuel design and core configuration dependent. These limits are required to prevent exceeding fuel design and safety criteria. The LHGR limits are determined-for various power / flow / exposure conditions to protect the fuel during off- . rated condition transient events [ References 3 and 4]. The applicability of these LHGR limits for specific fuel designs and core configurations is evaluated for each fuel cycle. TS license. amendment requests may be required to modify the LHGR power distribution limits or the off-rated factors for each. cycle. Therefore, consistent with .the relocaticn of _the APLHGR and MCPR limits, the LHGR limits should also be relocated to the COLR. t As stated in GL 88-16, the proposed amendment will result in a resource savings for the licensee and the NRC by eliminating the. majority of license amendment requests for changes in values of cycle-specific parameter limits in the TS. Indirectly, this is a safety irnprovement because these saved resources can be utilind on more important tasks. 1 .,C

. ~. Grand Gulf Nuclear Statiso Unit 1 Attachmem 2 to b, PCOL 92/07 GNRO 92/00093 Page 4 of G The description of fuel assembly design features is revised to delete the reference to the initial core design and to require that designs be developed and analyzed using NRC-approved codes and methods. Additionally, the design description requires fuel assembly types to be identified in the COLR This is consistent with the Technical Specification improvement Program and Reference 6. The editorial change to delete the word "both" from specification 2.1.2 clarifies this specification consistent with Reference 5. This word was inadvertently retained in Amendment 99 to the TS issued May 28,1992. The remaining editorial changes are justified by the need to maintain the TS as an organized document. D. NO SIGNIFICANT HAZARDS CONSIDERATIONS: The Commission has provided standards for determining whether a no significant hazards consideration exists as stated in 10CFR50.92(c). A proposed amendment to an operating license involves a no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the poss niity of a new or different kind of accident from any accident previously evaluated; or (3) its rolve a significant reduction in a margin of safety. Entergy Operations Inc. has evaluated the no significant hazards considerations in its request for a license amendment. In accordance with 10CFR50.91(a), Entergy Operations Inc. is providing the analysis of the proposed amendment against the three standards in 10CFR50.92(c). A description of the no significant hazards considerations determination follows:

1. - The proposed changes do not significantly increase the probability or consecuences of an accident previously evaluated.

The proposed amendment is in accordance with the guidance provided in Generic Letter (GL) 8816 for removing the cycle-specific parameter limit values from Technical Specifications (TS). There will be no changes in the operation of the facility as a result of these changes. No safety-related equipment or function Will be altered. The proposed amendment merely relocates cycle-specific parameter limits from the TS to the Core Operating Limits Report (COLR). The TS is revised to reference their inclusion in the COLR. NRC approved analytical methodologies will continue to be used as the bases for the limits contained in the COLR. The editorial changes only clarify a specification or provide organizational notation in the TS. Removal of the cycle specific parameter limits from the Grand Gulf Nuclear Station (GGNS) Technical Specifications has no influence or impact on the probability of any accident or malfunction evaluated in the GGNS Updated Final Safety Analysis Report (Reference 2). The cycle-specific operating limits, although not in the TS, will still be followed in operating GGNS. The proposed amendment requires exactly the same

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PCOL 92/07 GNRO 92/00093 Page 5 of 6 actions to be taken if limits are exceeded as are required by the current M. Therefore, the consequences of any accident previously evaluated have t..st increased. Based on the above, these proposed changes cannot increase the probability or consequences of any accident previously evaluated. 2. These changes would not create the possibility of a new or different kind of accident from any previously anal; ad. As stated above, no safety-related equipment, safety functions, or operating practices are altered as a result of these changes. No new accident modes are created. The proposed amendment is in accordance with the guidance provided in Generic Letter (GL) 8816 for removing the cycle specific parameter limit values from TS. The establishment of these limits in accordance with NRC approved methodologies and . incorporating these limits in the Core Operating Limits Report will ensure that a new and different kind of accident is not created. The removal of the cycle specific limits has no influence on, nor does it contribute in - any way, to the possibility of a new or different kind of accident or malfunction from those previously analyzed. Cycle specific limits are calculated using NRC approved methods. Technical Specifications will continue to require operation within the required core operating limits and appropriate actions will be taken when or if limits are exceeded.' Therefore, the proposed changes do not create the possibility of a new or different type of accident from any accident previously analyzed. 3. The proposed changes do not involve a significant reduction in a margin of safety. The proposed changes do not alter the requirement that the plant operate within the limits specified for a given operating cyclo, nor does it alter the required actions if these limits are not met. The margin of safety is not affected by relocating these limits from the Technical Specifications to the COLR. The margin of safety provided ' by the current technical specifications is unchanged. The proposed amendment still f-requires operation-within the core limits determined using NRC approved reload design methodologies and appropriate actions when or if limits are violated. Therefore, the proposed changes do not result in a significant reduction in a margin of safety, i . Based on the above evaluation, operation in accordance with the proposed amendment involves no significant hazards considerations. N._

.ac 3 g.. ,_.y i 16 g ~ b 3. C--- Grand Gulf Nuclear Station Unit 1

Attachment 2 to; W~

'PCOL 92/07 - GNRO 92/00093 Page 6 of G l J 1E.

REFERENCES:

t - 1. ' USNRC Generic Letter 8816, " Removal of Cycle Specific Pararnater Limits from the. Technical Specifications," October 4,1988 (MAEC 88/0313). - 2. Gran_d Gulf Nuclear Station Final Safety Analysis Report, Updated through Revision 6, Chapter 15. ', ~ - 3; EMF 91-169, Revision 1, " Grand Gull Unit 1 Cycie 6 Reload Analysis." Siemens. / Nucle'ar Power Corporation, July 1992. 4 -4. EMF-91 168, Revision 1, " Grand Gulf Unit 1-Cycle 6 Plant Transient Analysis," Siemens. Nuclear Power Corporation, July 1992. - 5.. - Safety Evaluation by the Office of NRR related to Amendment No. 99 to Facility Operating License No. NPF 29, May 28,1992. ? l, 6. NUREG.1434, Standard Technical Specifications, General Electric BWR/6 Plants, l- . Revision 0, September-28,1992. s i r i 1-y + N i- ~ ~ ^

l to GNRO 92/00093 GGNS PCOL 92/07 - MARKED UP TECHNICAL SPECIFICATION PAGES r Do

Pche / of M l INDEX MFINITIONS SECTION 1.0 DEFINITIONS PAGE 1.1 ACTION.................. 11

1. 2 AVERAGE PLANAR EXPOSURE......................................,,,

1-1

1. 3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE.............

1-1 1.4 CHANNEL CALIBRATION..............................,,.........,,,,,,,,, 11

1. 5 CHANNEL CHECK.........

1-1

1. 6 CHANNEL FUNCTIONAL TEST..............................................

1-1

1. 7

->1. 8 CORE ALTERATION...................................................... 1-2 CRITICAL POWER RATI0................................................. 1-2 1.9-DOSE EQUIVALENT I-131..............................................., 1-2 1.10 DRYWELL INTEGRITY.................................................... 1-2(p 1.11 E-AVERAGE DISINTEGRATION ENERGY............................... 1-3 1.12 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE T3.E................... 1-3 1.13 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME........ 1-3 1.14 FRACTION OF LIMITING POWER DENSITY............. 1-3 1.15 FRACTION OF RATED THERMAL P0WER............................... 1-3 1.16 FREQUENCY N0TATION................................, 1-3 1.17 GASEOUS RADWASTE TREATMENT (OFFGAS) SYSTEM........................... 1-3 1.18 IDENTIFIED LEAKAGE................................................... 1-4 1.19 ISOLATION SYSTEM RESPONSE TIME....................................... 1-4 1.20 LIMITING CONTROL R00 PATTERN......................................... 1-4 1.21 LINEAR HEAT GENERATION RATE.......................................... 1-4 1.22 LOGIC SYSTEM FUNCTIONAL TEST......................................... 1-4 1.23 MAXIMUM FRACTION OF LIMITING POWER 0ENSITY........................... 1-4 1.24 MEMBER (S) 0F THE PUBLIC.......................................... 1-4 1.25 MINIMUM CRITICAL POWER RATI0..................................... 1-5 1.26 0FFSITE DOSE CALCULATION MANUAL (00CM).......................... 1-5 1.27 OPERABLE --0PERABILITY............................................... 1-5 -v 'W l 7e Coed A 26W6 $'U SE# # l~E f ~ - GRAND GULF-UNIT 1 1-i Q[ed AN --

  1. WEY la c

!Ti .1 INDEX LIMITINGCONDITIONS-FOROPERATIONANDSURVEiLLANCEREQUI SECTION- _PAGE 3/4.0-APPLICA81LITY............................................... 3/4 0 1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1-SHUTDOWN MARGIN.............................................. 3/4 1-1 3/4.1.2 REACTIVITY AN0MALIES...:..................................... 3/4 1-2 3/4.1.3 CONTROL RODS n Co n t ro l Rod Ope rab i l i ty...................................... 3/4 1-3 Control Rod Maximum Scram InsertionTiaes.................... 3/4 1-6 Control Rod Scram Accumulators............................... 3/4 1-8 Control Rod Drive Coupling................................... t 3/4 1-10 Control Rod Position-Indication.............................. 3/4 1-12 Control Rod Drive Housing Support............................ 3/4 1-14 3/4.1.4 CONTROL' ROD PROGRAM CONTROLS Control Rod Withdrawal....................................... i 3/4 1-15 Rod Pattern Control Systen................................... 3/4 1-16 3/4.3 5 STANDBY LIQUID CONTROLSYSTEM................................ 3/4 1-1B 3/4.2 POWER DISTRIBUTION LIMITS-3/4.2.1 _ AVERAGE PL/MR LINEAR HEAT GENERATION RATE.......... 3/4 2-1 3/u. ot t t1 tD................................................ 3/42@ 3/4.2.3 MININJM CRITICAL POWtR RATI0........................... 3/42-M '3/4.2.4 LINEAR HEAT GENERATIONRATE.................................. 3/42-h s GRAND GULF-UNIT 1-iv Amendment No.16, -{

p ^ f &d.3 A WO 12/ boo 93 l };',~ l0 age 30/29' -INDEX ~ ! ADMINjsTRATIVE' CONTROLS - SEC"0N - PUE -kb ' SAFETY REVIEW COMMITTEE (SRC) (Continued) Review....................................................... 6-10 Audits....................................................... 6.. . Authority.................................................. 6-12 Records........................... E-12 6.5.3 TECHNICAL REVIEW AND CONTROL + . Activities................................................... s 6-12

6. 6 _ EPORTABLE= EVENT ACTI0N...........................................

R 6 6.7 SAFETY LIMIT VIOLATION............................................ ( 6-13' ~ a 6.8 PROCEDURES AND'PR0 GRAMS...........................................- 6-14 6.9 : REPORTING REQUIREMENTS Routine Reports..............................................- 6-15 .Startup Reports.............................................. 6-15 An n ua l : Rep o rt s............................................... 6-16 1 Annual Radiological Environmental Operating Report 6-17 Semiannual Radioactive Ef fluent Release Report............... I = .6 i, Monthi Atrjtting Reports._ m m............... 6-J9 w o & Q in ts peci a f Reports,....g<(coa). L n ...v. . ~ -. ~ __6 i .....................".. w..;n...T., ~ 6-19 D 6210L RECORD RETENTION.................................................. 6-19h g ~ -6.11^ RADIATION PROTECTION PROGRAM.'........ c.......................... 6-21 J p 6.'12 HIGH RADIATION AREA....'.......................................... 6-21 E '6.13 PROCESS PROGRAM CONTROL ~(PCP)............................ 6-22 6.14 OFFSITE DOSE CALCULATION MANUAL(00CM);.......................... 6., [GRANDGULF_-UNITL1 xx Amendment No. 87, - 6 1 2x a

- & c h W $ ~l* AWO VE/ YO' V 0: ; DEFINIT!0NSO frf 2 >y i -CORE ALTERATION' b ?1. 7 CORE ALTERATION shall be the addition, removal,-relocation or movement of fue),? sources,<incore instruments or: reactivity controls within the reactor s -pressure vessel lwith.the vessel head removed and fuel in the vessel. Normal- _ movement-ofJthe-SRMs,-IRMs,-LPRMs; TIPS, or special movable detectors is not considered to-be CORE: ALTERATION. Suspension of CORE ALTERATIONS shall not lprecluda: completion of-the _ movement of-a component to a safe conservative position. 3 " & c. CRITICAL-POWER RATIO l1.8 =The CRITICAL' POWER RATIO (CPR)'shall be the ratio of that power in the assembly n ich is calculated _by application of the ANF8' correlation to cause some point =in the-assembly to ex actualz assembly operating power.perience boiling transition, divided by the DOSE-EQUIVALENT I-131-1.9 DOSE _ EQUIVALENT;I-131 shall-be: that concentration of I-131, microcuries per gram, which:alone,would produce the _.same thyroid dose as the-quantity and isotopic mixture ofc I-131,1-132,- I-133,J I-134, and I-135 actually present. lThe thyroid dose conversion factors used for this calculation shall be those listed in Table III.of TIO-14844, " Calculation'of! Distance Factors for Power and Test Reactor Sites."-

.0RYWELL INTEGRITY

.1.10 DRYWELL INTEGRITY shall exist"when: ai Allldrywell-penetrations' required to be closed during accident 1 conditions are either: 1.'

Capable of being closed by 'an OPERA 8LE drywell-automatic isolation system, or
2. -

. Closed by at least one: manual valve, blind flange,:or deactivated automatic valve secured it, its closed position, except as provided in Table _3.6.4-1 of Specification 3.6.4. ?b. .The.drywell: equipment' hatch is, closed and: sealed. The:drywell-airlock is in' compliance with the requirements of c.- Spacification 3.6.2.3. d.- The drywell; leakageL rates are within the~ limite of Specification - 3.6.2.2. The suppression pool is'in compliance with the requirements of ' e. LSpecification~3.6;3.1. f. The sealing mechanism. associated with-each drywell penetration; 3 e g!, welds,Jbellows or 0 rings, is OPERABLE. f a LGRAND GULF-UNIT--1 1-2 Amendment No. 7 3, -

3:g. 4Wach n W Jr-f.o fMeo 92/9co93 j

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A k.g e -- s - J a f 2 9 Insert for Page 1-2 ~ CORE OPERATING LIMITS REPORT-(COLB1-2 '1.7a-The COLR is the Grand Gulf Nuclear. Station specific document that provides I

core operating lim 1ts for the current reload cycle. These cycle specific operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.11.

Plant operation within these operating limits is addressed in individual Specificatlons. I 4 e 4 1 e e y -ge. ,,y.g.w. .eg..y9

)?%cAe-ed[lo]Aio $Z/Ao1s a $ii &g e : 6 of 2 V 2.0~ SAFETY LIMITS AND LIMITING SAFETY-SYSTEM SETTINGS

. x 2.1' SAFETY LIMITS m

THERMAL POWER. Low Pressure or Low F1 w 2.1.1 THERMAL POWER shall not ex:eed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow. APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2. ACTION: With THERMAL POWER exceeding 25% of RNiED THERMAL POWtR and the reactor vessel steam dose pressure less than 785 psig or core flow less than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours and comply with the requirements of Specification 6.7.1.

a THERMAL-POWER. Hich Pressure and Hiah Flow

} 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall net be less than 1.06 i during 4et& two loop operation and 1.07 during single loop operation'with the reactor vessel steam done pressure greater than 785 psig and core flow greater than 10% of rated flow. APPLICABILITY,: -OPERATIONAL CONDITIONS 1 and 2. -ACTION: . With MCPR less;than the above limits and the reactor vessel steam dose pressure ' greater than 7A5 psig and core flow greater than 10% of rated flow, be in at least HOT SHUTOOWN within 2 hours and comply with the requirements of Specif t-cation 6.7.1. -REACTOR COOLANT SYSTEM PRESSURE-The reactor coolant rystes pressure, as measured in the reactor vessel 2,1.3 steam done, shall not exceed 1325 psig. -APPLICABILITY: OPERATIONAL CONDITIONS 1, 2' 3 and 4. ACTION: 'With the reactor coolant systes pressure, as measured in the reactor vessel steam; dose, above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant .y system' pressure less than or equal to 1325 psig within 2 hours and comply with 7 the requirements of Specification 6.7.1. I 99,__ ,i f Amtsndment No. J3. 5 I 2-1 GRANO GULF-UNIT 1 h

AdaeLed J -/o (wto 9tho 9,3 t Atge 7 of J V 3/4.2 PCVEK O!$TRIBUTION LIMIT 5 1/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIHlY!NG CON 0! TION FOR OPERATION 1::;-1;;r tten,Ae 411 AVERAGE PLANAR LINEAR HEAT GENERAT 3,2.1 -Ge43-tu: RATES ( APLHGRs) 4+r :::h ty;;4f-he4-eo-e-fvneWa-of-AVERA0E PLANA;; tuowRE--. shall not exceed the limits r.h;;r in Figure 3.2.1-li spedOed 'r & cckE o$rati M*l,V i if R on ( of AVfRAGE PLANAR EXPOSURE shall not exceed the limit shown in figure 3.2 " ~ - - Q Q by 0.8 APPLICABILI1~Y: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER. ACTION: During two loop cperation or single loop oseration, with an APLHGR exceeding the limits,i'iste corrective action within 15 minutes and restore APLNGR toc' figure 3 y-the ;;;repriett-muttiplicitia facter, in. within the required limits within 2 hours or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours. SURVEILLANCE *r 11REMENTS 4.2.1 All APL}. s shall be verified to be equal to or less than the required limits: a. At least once per 24 Murs, b. Within 12 hours after completion of a THERMAL POWER incraase of at least 15% of RATED THERMAL POWER, and Initially and at least once per 12 hours when the reactor is c. operating with a LIMITING CONTROL R00 PATTERN for A?LHGR, d. The provisions of Specification 4.0.4 are not applicable. 4 'l, -- Amendment No. 40rW'l GPAND GULF-UNIT 1 3/4 24

Aage #o/2v 8 1 1 5 I [ [,I w / R \\ d n s I I et 3 E I i l 9 i N [ I w I E I i .i a !1 i i i t 'O WN90 WDKWWI GitAND GULF-UNIT 1 3/4 2 2 Amendment No.11 39

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  • UNIT 11 3/4 3./ 2.

Amendment No. 16>._ t i ' f I^ I.,1 Li

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A$ckned J /c 6Affo f]/c,co93 POWER Dl5TR'BUTION t!MITS OS[rC /0 of M 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.3 The MINIMUM CRITICAL P0'aER RATIO (MCPR) shall be equal to or greater thantheMCPR/MCPPy era 7:4A-limits et-4*dieeted core f'cw, TliERMAL POE R, f and c qocure, et ihm" ' f qvre:-3.2.3-1, 3.0.3-2, end-h 2.3-3. ( APPLICABILITY: OPERATIO b l CONDITION 1, when THERMAL POWER is greater than or equalto2WofRATEDTHERMyA06,.. _ m-m - ACTION: speahed i. He CCR2 WEhM U'M AMT-With MCPR less than the applicable MCPR limits determined frca figures 3.2.0-L, -3.2.3-2, oi,d 3.2.0-3, initiate corrective action within 15 minutes and restore MCPR to within the required limits within 2 hours or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours. SURVFILLANCE REQUlREMENTS 4.2.3 MCPR shall be determined to be equal to or greater than the applicable MCPR limits, determined from Tigures 3.2. F 1, 0.2.0 2, eiid 3.2.3-3) At least oh,ce per 24 hours, a. b. Within 12 hours af ter completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and Initially and at least once per 12 hours when the reactor is c. operating with a LIMITING CONTROL R00 PATTERN for MCPR. d. The provisions of Specification 4.0.4 are not applicable. GRAND GULF-UNIT 1 3/4 2-g Amendment No. 73,- l l l

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  1. WdOM GRAND GULF UNIT 1 3/4 2-5 Amendment No, /j 95

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7 f., '/ l y .... l j. _.. m g g m e j /h g I clh G V w J-g i 1 i l .y g l l I m-a w q . v. M N N d id M CRAN 0 GULF-UNIT l' 3/4 2 6 Amendment i4o.1199

.--_---a--,---

/l$&d J /c 6AWd f// coo 9.1 A2f t' /3 c A 2 4 I 4 + t I ~ W n i 6 M g g h g@v/ = s Pt l w 5 n,- n r-e e e 'WdOW GRAND GULF-UNIT 1 3/4 2 $a Amndant go, ff F

1 PCbE4 Oll'#!8L*.*0N Lltil*$ s' 3/4 2.a LINEAE wtAT GENERAf!ON RATE fcAlt* /Y db EY v LIMITING CONDITION FOR OPERATION 3. Th t.1NEAR HEAT GEN RAT N RATE LHJGRA all not exce o de limit bMLT6 in Figure .2.4 Tas mu tip too oy tre smaller o L.HGR factor (LHGRFAC ) of Figure 3.2.4-2 or the power dependent trGR < acte ettner ne low eecence t f (LHGRFAC ) of Figure 3.6 p c Do$$f2[,i/AGrr espeer, ADPt,!C ABIL]TY: equal to 2ST of RAT [D THERMAL POWER. OPERATIONAL CON 0! ACTICN: with the LHGR of any fuel rod exceeding the limit -et-54p- ' ' ' ' ,,4way t he -- :Sps.cor. tate - e ! t !M4c4444a-14c44*. Initiate eorreetiye aetion within 15 minutes and restore the LNGR to within the limit reouce THERMAL POWER to ?ess than 25% of RATED T 4 hours, x SURVEILLANCE REQUIREMENTS 4.2.4 LHGR's shall be determined to be equal to or less than their allowa limits: At least once per 24 hours, a. b. Within 12 hours af ter completion of a THERMAL POWER increase least 15% of. RATED THERMAL POWER, Initially and at least once per 12 hours when the reactor is c. operating on a LIMITING CONTROL R00 PATTERN for LHGR, and d. The prothions of Specification 4.0.4 are not app 1tcable. ' GRAND GULF-UNIT 1 3/4 2-4 Amendment No. 73 r ;

haCIMN 3 /0 6M'20 f?l00093 OgC /5 of2'f i I' i i ,o 5 /l i ...............,_._._,,,,,,,,,,i i i i i l b I i 3 .._ 4. 1 I g w L b i-i W a D (,. \\ t n 1 .e w g n g. g S 3 3 3 g e e

  • bllM$ M

. GRAND rutF. UNIT l' 3/4 2*74 Amendment No. 77 99 l

y-n 854clws/ J fo (wgo 9g,(yeo93 . ?! Aty H, o / 2Y l i B t t .._......_p'.._..........._............. g i I j b;0' 9b i, i. g i i the W e 1 I 5 ri k / / l 2 g L e 0 I I I I I 8cymom GRAND GULF Uhli 1 3/4 2 7D Amendment No.13 95

A$bdd b 6AQ 9E/$0093 e c. k e /7 ol d Y g H ) ........;_....._....... l....... ..L j a i 3m t { i / i / 6 )-+- 8$ f a 10@&( W, E I I. 2 7 IA e; I I .L.. 6 g g /' / l 0 -f 2 i e I-I I I I E d DVMOH1 GilAND GULF-UNIT 1-3/4 2 7c Amendment No. 73 ed

/YGCNed 3 b bA@ 9Ll0053 3/4.2 POWER 0!$TRIBUTION LIMIT 5 h /8 of 2V l BASES The specifications of this section assure that the peak cladding temper-ature following the postulated design basis loss of coolant accident will not exceed the 2200'F Ilmit specified in 10 CFR 50.46, 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE This specification assures that the peak cladding temperature following the postulated design basis loss of coolant accident will not excced the limit rpecified in 10 CFR 50.46. The peak cladding temperature (PCT) following a postulated loss of coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is dependent only secondarily on the rod to rod power distribution within an assembly. The Maximus Average Planar Linear Heat Generation Rate (MAPLHGR) lialts -eMiture44rl*1-ere- -eppHeeble-te-two-1;;p ;;;retient ( For single loop operation, a MAPLHGR limit corresponding to the product of the&MAPLHGR/11gurs-3rt&l7and t96 dan be conservatively used to ensure that the PCT for single loop cperation is bounded by the RT for two loop operation. The daily reoutrement for calculating APLHGR when THERML POWER is greater than or equal to 25% of RATED THERMAL POWCR is sufficient since power distribu-tion shifts are very slow when there have not been significant power or control s reducI% Sc he yc3ied k 'de 60LR A el fE - O PE 9-hTINC, LIM I TS R L P o l2. T (COLR.f. Y>- 8 3/4 2-1 Amendment No. M GRAND GULF-UNIT 1

hack +l3 /c b/VE0 ldOO 93 }hge /9vf2V POWER DISTRIBUTION LlHITS 8A$ES HINIMUM CRITICAL POWER RATIO (Continued) During initial start up testing of the plant, a HCPR evaluation will be made at 25% of RATED THERMAL POWER level with minimum recirculation pump speed. The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level will be shown to be unnecessary. The daily requirement for calculating MCPR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The requirement to calculate MCPR within 12 hours after the completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER ensures thermal limits are met af ter power distribution shif ts while still a11otting time for the power distribution to stabilize. The requirement for calculating MCPR af ter initially determining a LIMITING CONTROL ROD PATTERN exists ensures that MCPR will be known following a change in THERMAL POWER or power shape, that could place operation exceeding a thermal limit. 3/4.2.4 LINEAR HEAT GENERATION RATE This specification assures that the Linear Heat Generation Rate (LHGR) in any rod is less than the desi0n linear heat generation even if fuel pellet densification is postulated. The LHGR limits ef Ti m 3 1. M multiplied by the smaller of either the flow dependent LHGR factor (LHGRFAC ) or the power dependent LHGR factor f (LHGRFAC ) corresponding to the existing core flow and power state to ensure p adherence to the fuel mechanical design bases during the limiting transient. LHGRFAC 's are generated to protect the core from slow flow runout transients. f A curve is provided based on the maximum credible flow runout transient for Loop Manual operation. The resuit of a single failure or single operator error during operation in loop Manual is the runout of only one loop because both recirculation loops are under independent control. LHGRFAC 's are p generated to protect the core from plant transients other than core flow increases. The daily requirement for calculating LHGR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distri-Lbution shifts are very slow when there have not been significant power or control rod changes. The requirement to calculate LHGR within 12 hours after the completion of a THERMAL F0WER increase of at least 15% of RATED THERMAL POWER ensures thermal limite are met after power distribution shifts while still allotting time for the power distribution to stabilize. The requirement for calculating LHGR after inittaity determining a LIMITING CONTROL ROD PATTERN exists ensures that LHGR will be-known following a change in THERMAL POWER or power shape that could place operation exceeding a ther.nal limit. 9 %, -- GRAND GULF-UNIT 1 8 3/4 2-7 Amendment No. U, #

& J Jr'c N'*0Pf,dec93 r e /S.y 20 ofaV' DESIGN FEATURES

5. 3 REACTOR CORE F,UEL ASSEMBLIES 5.3.1 The reactor core shall contain 800 fuel assemblies.

Each fuel assembly shall contain fuel rods and water rods clad with Zircaloy cladding. Each fuel rod sh&1l have a design nominal active fuel length of 150 inches. -Th; i,itial= --cere loading.4ha4haus-a 4ssignaeminabenrtchment-of-4r708-wo4 ht-pecc+n&- - 9 -4M3th-Reload fuel shall have mechanical, thermal-hydraulic and neutronic characteristics compatible with the initial core loading. CONTROL ROD ASSEMBLIES 5.3.2 The reactor core shall contain 193 control rod assemblies, each consisting of a cruciform array of stainless steel tubes containing a design nominal 143.7 inches of boron carbide,.B.C powder surrounded by a cruciform shaped ;tainless steel sheath. 5.4 REACTOR C0OLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4i1 The reactor coolant system is designed and shall be raintained: a. In accordance with the code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuar.t to the applicable Surveillance Requirements, b. For a pressure of: 1. 1250 psig on the suction side of the recirculation pump. 2. 1650 psig from the recirculation punp discharge to the outlet side of the discharge shutoff valve. 3. 1550 psig from the discharge shutoff valve to the jet pumps, c. For a temperature of 575'F. E VOLUME 5.4.2 The total water and steam volume of the reactor vessel and recirculation system is approximately 22,000 cubic feet at a nominal T,y, of 533'F. I GRAND GULF-UNIT 1 5-5 Amendment No. 57, -

1 b 5Ac l M d J /c (* fC 9Z & 93 .(_ e ?agt 2 / of dY b '. insert for Page 5 5, (end of 5.3.1) e Fuel assemblies shall be limited to thoss fuel designs that have been analyzed with applic"ble NRC Staff approved codes and methods, shown to comply with all safety design bases, and are identified in the Core Operating Limits Report, t I.. i + i A;f; L. 4-I_" j

$Ned J Wo 6/.Td 9//00093 q ADMINISTRATIVE CONTROLS i-SEMIANNOALRADIDACTIVEEFFLUENTRLEASEREPORT(Continued) -Principal radionuclide (specify whether determined by measurement or c. estimate),

d. -

Type of waste (e.g., spent resin, compact dry waste, esaporator bottoms), e. Type of container (e.g., LSA, Type A, Type 8, large Quantity), and

f. - Solidification agent (e.g., cement, urea fonsaldehyde).

The radioactive effluent' release reports shall include unplanned releases from -the sitelto the UNRESTRICTED AREA of radioactive materials in gaseous and liquid offluents on a quarterly _ basis.

The radioactive effluent release reports shall include any chan es to the

' PROCESS CONTROL PROPRAM (PCP) 0FFSITE DOSE CALCULATION MANUAL 00CM) or radio-

activewastesystemsmadedurIngthereportingperiod.

MONTHLY OPERATING REPORTS 6.9.1.10 - Routine reports of operating statistics and shutdown experience yo.A

includin

-valves.gdocumentationofallchallengestomainsteamsystemsafety/rellef shall be submitted on a monthly basis no later than the 15th of each y month following the calendar month covered by the report. SPECIAL REPORTS Me" +' ^ed P^bt ~ 6 'In 6.9.2-'Special reports shall be submitted to the Nuclear Regulatory Commission

pursuant to Section 50.4 'of 10 CFR Part 50 within the time period specified
for each' report; 6.10 RECORD RETENTION In! addition to-the applicable record v tention requirements of Title 10 Code e

of Federal: Regulations the following records shall be retained for at least 'the minimum period-ind cated.- ? .6.10;1 The following records shall be retained for at least five years: Records-and logs of unit operation covering time interval at each a. . power _ _ level-. b.- Records-and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to - nuclear safety, c. All-REPORTA8LE EVENTS. I -~ GRAND GULF-UNIT.-11 _6-19 Anendeent No.80, - t 4 --_____mm.m...-------

e 1 /lkM l 3 l0 6 A'Z O ff /0 0 0 'U g /kge 2 3 o b d 'l L Insert for Page 619 CORE OPERATING LIMITS REPORT (COLR1 6.9.1.11 Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the CORE OPERATING LIMITS REPORT (COLR) for the following: i a. The Average Planar Linear Heat Generation Rate (APLHGR) for Technical Specification 3.2,1. b. The Minimum Critical Power Ratio (MCPR) for Technical Specification 3.2.3. c. The Lineur Heat Generation Rate (LHGR) for Technical Specification 3.2.4. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in the following_ documents. The appropriate revision / supplement number for each document shall be identifled in the Core Operating Limits Report.

1) XN NF 79 71(P), Exxon Nuc!carflant Transient Methodology.for Bollina Water Reactors, Exxon Nuclear Company, Inc., Richland, WA.

2)- XN NF 8019(P)(A), Volume 1, Fxxon Nuclear Methodology for Bollina WatgI Reactors - Neutmulc. Methods for Deslon and Analvsis, Exxon Nuclear Company, Inc., Richland, WA.

3) XN NF 8019(P)(A), Volume 1, Advanced Nuctsar Fuels Methodoloov for Bollina Water Reactors: Benchmark Results for the CASMO 3G/MICROBURN B Calculation Methodology, Advanced Nuclear Fuels Corporation, Richland, WA.
4) XN NF 8019(P)(A), Volume 3, Exxon Nuclear Methodology for Bollino Watnf Henctors THERMEX: Thermal Limits Methodology Summarv Descrlotion,"

l Exxon Nuclear Company, Inc., Richland, WA.

5) ;ANF-913 (P)(A) Volume -1,.COTRANSA2: A Comouter Proaram for Bollina Water Reactor Transient Analvsis, Advanced Nuclear Fuels Corporation, Richland, WA.
6) ANF 1125(P)(A), ANFB Critical Power Correlat.on, Advanced Nuclear Fuels Corporation, Richland, WA.

l

7) XN NF 84-105(P)(A), Volume 1, XCOBRA-T:

A Comouter Code for BWR 't ransient Thermal Hvdraulic Core Analvsis, Exxon Nuclear Company, Inc., Richland, WA.

8) XN NF 573(P), RAMPEX Pellet-Clad Interaction Evalnation Code for Power l

Ramos, Exxon Nuclear Company, Inc., Rl::hland. WA.

fl/h C l & J /o ShW 9/k00U frge n'of zv'

9) XN-NF 8158(P)(A), RODEX2:

Fuel Rod Thermal Mechanical ResDRD12 Evaluation Model, Exxon Nuclear Company, Inc., Richland< "'A.

10) XN NF 85 74(P)(A),RODEX2A(BWRh Fue!RodThermal MechanicalResnonse Evaluation Model, Exxon Nuclear Compariy, Inc., Richland, WA.

l

11) XN CC 33(P)(A), HUXY: A Generallred Multirod Heatuo Code _with 10CFR50 Accendix K Heatuo Ootion, Exxon Nuclear Company, Inc., Richland, WA.
12) XN NF 825(P)(A), BWR/6 Generic Rod Withdrawal Error Analysis, MCPR.igi p

Plant Qooration Within the Extended Oooratina Domain," Exxon Nuclear Company, Inc., Richland, WA.

13) XN NF 8151(P)(A), LOCA Seismic Structural Resnonse of an Exxon Nucle.at Comoany BWR Jet Pomo Fuel Assembiv, Exxon Nuclear Company, Inc.,

Richland, WA.

14) XN NF 84 97(P)(A), LOCA Ssismic Struntural Resoonse of an ENC 9x9 BWR Jet Pomo Fuel AssRIDbly, Advanced Nuclear Fuels Corporation, Richland, WA.
15) XN NF 86 37(P), Generic LOCA Break Soontrum Analysis for BWR/6 Plants, Exxon Nuclear Company, Inc., Richland, WA.
76) XN NF 82 07(P)(A), Exxon Nuclear Company ECCS Claddino Swelling._and Bitoture__Modql, Exxon Nuclear Company, Inc., Richland, WA.
17) XN NF 8019(A), Volumes 2, 2A, 28, & 2C, Exxon Nuclear Mettigdoloov for Bollino Water Reactors EXEM BWR ECCS Evaluation Model, Exxon Nuclear Company, Inc., Richland, WA.
18) XN NF 79 59(P)(A), Methodoloav for Calculation of Pressure Droo in BWR Fuel Assemblies, Exxon Nuclear Company, Inc., Richland, WA.

The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanicallimits, thermal hydraulic limits, Emergency Core Cooling System (ECCS) limits, Nuclear limits such as shutdown margin, transient analysis limits, and accident analysis limits) of the safety analysis are met. The COLR, including any mid cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC Document Control Desk with copies to the Regional Administrator and Posident inspector. 1 --~

-., ie. to GNRO 92/00093 GGNS PCOL 92/07 - PROPOSED TECHNICAL SPECIFICATIONS (Information Only) k

? = b ]

f> + to GNRO 9200093 i l' Page 1 cf 16 INDEX 'OEFINITIONS SECTION 1.0 DEFINITIONS EAq[ 1.1 ACT10N............................................................. 1-1 1.2 AVERAGE PLANAR EXP0SURE............................................ 11 1.3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE......................... 11 1.4 CHANNEL Call 8 RAT10N................................................ 11 r 1.5 CHANNEL CHECK...................................................... 1-1 !L 1.6 CHANNEL FUNCTIONAL = TEST............................................ 11 1.7 CORE AL1 ERAT10N.................................................... 12 L 1.7a CORE OPERATING-LIMITS REP 0RT....................................... 12 1.8 CRITICAL POWER RAT10............................................... 1-2 1.9 DOSE EQUIVALENT l-131.............................................. 12 .l.10 DRYWELL INTEGRITY.................................................. 1-2a l .1.11 E AVERAGE DISINTEGRATION ~ ENERGY.................................... 1-3 1.12 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME................. 13 1.13 ENOJ0F-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME.......... 13 1.14 FRACTION OF LIMITING POWER 0ENSITY.................. 1-3 1.15-FRAtt10N OF RATED THERMAL P0WER.................................... 1-3 1.16-FREQUENCY NOTAT10N................................................. 1-3 .l.17:GASE005 RADWASTE TREATMENT (OFFGAS)-SYSTEM......................... 1-3 1.18 IDENTIFIED LEAKAGE................................................. 14 l'.19 ISOLATION SYSTEM. RESPONSE TIME..................................... 1-4 1.201 LIMITING CONTROLL 00 PATTERN....................................... 1-4 R '1.21 LINEAR HEAT GENERATION RATE..............................-.......... 1-4 t 1 ; 2 2 : LOG I C SYST EM FUNCT IONAL T EST......................................-. 14 o 1.23 MAXIMUM FRACTION OF LIMITING POWER DENSITY......................... 14 1.24 MEMBER (5) 0F THE.PUBLIC................................. 14 -l',25 MINIMUM CRITICAL POWER RAT10....................................... 1-5 ~1.26 0FFSITE DOSE CALCULATION MANUAL.(00CM)............................. 1-5 -1.27 OPERABLE OPERABILITY............................................. 1-5 l GRAND GULF-UNIT 1; i-Amendment No. ???, e ...,i.i,..,i.,,,...-.... - I'- - - - - - - " - - - - - - - - - - - ^ - " - - - - - " - - - " - - - " " - ' - " " ' ^

i q Attacivnent 4 O GNRO 92/00093 l Page 2 cf 16 JNDEX LIMITING' CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS i l 1[CTION pjg 3/4.0 APPLICABILITY,............................................... 3/4 0-1 '3/4.1 'REACTIVill_ CONTROL SYSTEMS 3/4.1.1 ' SHUTDOWN MARGIN........................................... 3/4 1-1 3/4.1.2. REACTIVITY AN0MAllES...................................... 3/4 1-2 I -3/4.1.3 CONTROL RODS c Control Rod Operab(11ty................................... 3/4 1 3 l-Control Rod Maximum Scram insertion Times................. 3/4 1 6 o I Control Rod: Scram Accumulators............................ 3/4 1-8 Control Rod Drive Coupling................................ 3/4 1 10 Control Rod Position Indication........................... 3/4 1-12 Control Rod Drive Housing Support......................... 3/4 1 14 3/4.1.4: CONTROL R00'PROGRAMLCONTROLS Cont rol Rod Wi thdrawal..................................... - 3/4 1 15 p Rod Pattern Control-System................................ 3/4 1 " 23/4.1.5 STANDBY LIQUID CONTROL SYSTEM............................. 3/4 1 18 F 3/4^.2 POWER DISTRIBUTION LIMITS L L3/442.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE................ 3/4 2-1 c3/4.2.2 DELETED........................................ 3/4-2 2 3/4.2.3' MININUM CRITICAL POWER RAT 10............................... 3/4 2-3 3/4.2.4 LINEAR HEAT GENERATION RATE............................... 3/4 2., - r.. a vt' GRAND GULF-UNIT iv Amendment No. 16, ??? as .= -A

cf 3-to GNRO 92A)0093 L Pege 3 cf 16 INDEX ? ADMINISTRATIVE CONTROLS SECTION

ggf, SAFETY REVIEW ComITTEE (SRC)

(Continued) Review....................................................... 6 10 l Audits....................................................... 6 11 L Authority.................................................... 6 12 L Records...................................................... 6 12 ? 125.3 ' TECHNICAL REVIEW AND-CONTROL e Activities................. 6 12 s i L6.6-REPORTABLE EVENT ACT10N........................................... 6 13 g 6.7 SAFETY-LIMIT VIOLATION............................................ 6-13 6.8 HPROCEDURES AND-PROGRAMS........................................... 6 14 '65 9 REPORTING' REQUIREMENTS _ Routine Reports.............................................. 6 15 ^ Startup Reports.............................................. 6 15 Annual Reports............................................... 6 16 Annual Radiological ' Environmental Operating Report...........- 6 17-Semiann'ual Radioactive Ef fluent: Release Report............... 6 17 " Monthly Operating Reports................. 6 19 Core Operating Limits. Report................................. 6 19 f Speci al L Report s '............-................................. 6 19b i .6.10 RECORD RETENTION...'.............................................. 6 19b '6'11' RADIATION: PROTECTION PR0 GRAM..........................._.......... 6 21 16.12 =HIGH RADIATION AREA.............................................. 6 21 ~ 6.13' PROCESS: PROGRAM CONTROL (PCP)................................... 6-22 6.14D OFFSITE 00SE. CALCULATION MANUAL (0DCM)..i......................... 6 22 + TGRAND-GULF; UNIT 1: xx Ameridment No. 87, ??? T. m 9 y. _ y7 -%. g g a, 9 yn,,3 9 y 9 4, y-s. yy,y ,p-v.v--p .Wy.y3my 3W~-

F q 1 Attactwnent 4 t3 GNRO 9200093 DEFINITIONS see 4 f16 -m=, CORE ALTERATION 1.7' CORE ALTERATION shall be the addition, removal, relocation or movement of fuel, sources, incore instruments or reactivity controls within the reactor pressure vessel witF the vessel head removed and fuel in the vessel. Normal movement of the SRMs IRMs, LPRMs, TIPS, or special movable detectors is not considered to be CORE ALTERATION. Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe conservative position. { CORE OPERATING LIMITS REPORT (COLR) 1.7a The COLR is the Grand Gulf Nuclear Station specific document that provides core operating limits for the current reload cycle. These cycle-specific operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.11. Plant operation within these operating limits is addressed in individual Specifications. CB111 GAL POWER RATIO 1.8 The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in the assembly which is calculated by application of the ANFB correlation to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power. DOSE EOUly8 LENT l-131 1.9 DOSE EQUIVALENT I 131 shall be that concentration of I-131, microcuries per gram, which alone would produce the same thyroid dose as the quantity and isoto)ic mixture of 1 131, 1-132, I-133, 1 134, and I 135 actually present. The t1yroid dose conversion factors used for this calculation shall be those listed in Table III of TID 14844, " Calculation of Distance factors for Power and Test Reactor Sites." GRAND GULF-UNIT 1 1-2 Amendment No. 73, ??? 1 I

Attadwnent 4 to GNRO 92KX1093 d= s Page 5 cf 16 L DEFINITIONS S = DRYWELL INTEGR11Y l.10 DRYWELL INTEGRITY'shall exist when* $f All 'drywell penetrations required to be closed during accident a. conditions'are either: 1. Capable of being closed by an OPERABLE drywell automatic isolation system, or. .. t

2. -

Closed by at least one manual valve, blind flange, or s

p deactivated automatic valve secured in its closed position, except as provided in Table 3.6.41 of Specification 3.6.4.

.) n b. The drywell equipment hatch is~ closed and sealed.

The drywell airlock is.in compliance with the requirements of c.

Specification 3;6.2.3. [d. - .The dryweil'leakageirates are within the limits of. Specification 3.6.2.2. LThe. suppression pool is in. compliance with the_ requirements of e. Specification 3.6;3.1. e i fe Theisaaling mechanism associated with each drywell penetration; e.g., welds, bellows or 0 rings, is OPERABLE.< -w im h I i h -' ~ t. p s- [1 s t GRAND GULF-UNIT =le 1-2a Amendment No. ??? .'r-L' .~s a ~- -

41 to GNRO 9200093 1 P o] 6 cf 16 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS i 2.1 SAFETY LIMITS i _ THERMAL POWER Low Pressure or low Flow 2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow. 1 APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2. ACTION: With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours and comply with the requirements of Specification 6.7.1. IHERMAL POWER. Hioh Pressure and Hioh Flow 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.06 l during two loop operation and 1.07 during single loop operation with the reactor vessel steam dome pressure greater than 785 psig and core flov greater than 10% of rated flow. APPLICABILIT1L OPERATIONAL CONDITIONS 1 and 2. ACTION: With MCPR less than the abov6 limits and the reactor vessel steam dome pressure greater than 78s psig and core flow greater than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours and comply with the requirements of Specification 6.7.1. REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The reactor coolant system pressure, as measured 'n the reactor vessel i steam dome, shall not exceed 1325 psig. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and 4. ACTION: With the reactor. coolant system pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant system pressure less than or equal to 1325 psig within 2 hours and comply with the requirements of Specification 6.7.1. k GRAND ~ GULF-UNIT 1 2-1 Amendment No. 99, ??? \\!-

_ to GNRO 9200093 Page 7 cf 16 3/4.2 POWER DISTRIBUTION LIMITS ~' 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) shall not e exceed the limits specified in the CORE OPERATING LIMITS REPORT. t APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER. ACTION: e During two loop operation or single loop operation, with an APLHGR exceeding the limits, initiate corrective action within 15 minutes and restore APLHGR to within the required limits within 2 hours or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours. SURVEILLANCE REQUIREMENTS 4.2.1 All APLHGRs shall be verified to be equal to or less than the required limits: a. At least once per 24 hours, b. Within 12 hours after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and c. Initially and at least once per 12 hours when the reactor is operating with a LIMITING CONTROL R00 PATTERN for APLHGR. d..- - The provisions of Specification 4.0.4 are not applicable. l 1 GRAND GULF-UNIT 1 3/4 2-1 Amendment No. 99, ???

[6 8 p, f_QER DISTRIBUTION LIMITS 3/4.2.2 f0ELETEDJ, .e _ GRAND GULF-UNIT 1. 3/4 2-2 Amendment No. 16, ???

1 <*J Y Attachmet 4 to GNRO 92MXB3 Pcge 9 sf 16 ' POWER DISTRIBUTION LIMITS- ~ ' 3/4.2.3' MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION ~ 3.0. 3 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater than the MCPR limits :pecified in the CORE OPERATING LIMITS REPORT. APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or . equal to 25% of RATED THERMAL POWER. AGlLQKi 'With-MCPR less than the applicable MCPR limits, initiate corrective action .'within 15 minutes and restore MCPR to within the required limits within 2 -hours or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within 'the next 4 hours.- L SURVEILLANCE REQUIREMENTS .m k 4.2.3 MCPR shall be determined to be equal to or greater than the applicable --MCPR' limits; i p 'a. Atileast once per 24 hours, b. Within 12 hmirs after ~ completion-of a THERHAL POWER increase of at least 15% ef-RATED THERMAL POWER, and-c. Initially an'd at.least once per 12 hours when the reactor is . operating with a LIMITING CONTROL ROD PATTERN for MCPR. d. The provisions of Specificat:on_4,0.4 are not applicable. p I h F a t -GRAND GULF UNITel '3/4 2-3 Amendment No. 73, ??? yl.. ,9 -w

Wu + k,C A-ni.is 4 to GNRO 9200093 ' ' Mr M; J-Paos 10 cf 10 _. POWER' DISTRIBUTION LIMITS! q.[* 3/412.4 LINEAR HEAT GENERATION RATE. LIMITING CONDITION FOR.0PERATION. g m U< P '3;2;4 The LINEAR HEAT GENERATION RATE (LHGR) shall not exceed the: limits-yy .specified in the CORE OPERATING-LIMITS REPORT. E ec - APPLICdBILITY:.LOPERATIONALCONDITION:1,whenTHERMALPOWERisgre:terthanor ~ = equal.-to 25% of. RATED THERMAL-POWER.

eCTION

s 34 w With the.LHGR of any fuel er e aag_the limit, initiate corrective action

  • $.P?

.within-15 minutes and restore the LHGR to within the-limit within 2 hours or reduce: THERMAL = POWER to less than-25% of RATED THERMAL POWER within the next 4 Qo ' hours.-- g 5 ~ ! SURVEILLANCE REQUIREMEN1= + ?4.2.4-LHGR's sh'all' be determined to be equal to or less than their allowable limits: -- a.- 'At least once per-24.h N rs, p 1'

b.

' Within'12thours after completion of-a THERMAL POWER increase of at least115%iof-RATED THERMAL POWER, ' c.: . Initially and atileast once per 12 hours when the-reactor is - operating on a -LIMITING CONTROL :J0 PATTERN for LHGR, and' id.

T* 3 provisions Lof LSpecification '4.0.4 are not applicable.

c. R_ (r.) 1 1, g.-- ,s f < GRAND GULF 2VNIT:11 3/4 2-4 -Amendment No. 73, ??? ( 'h,1 g t 7 er w w m-rw-1 + ~ -m -p

e a Page 11 of 16 3/4.2 POWER DISTRIBUTION LIMITS BASES a. The specifications of this section assure tSat the peak cladding temper-ature following the postulated design basis loss of-coolant accident will not exceed the i,00*F limit specified in 10 CFR 50.46. 3/4.2.1 AVERAGE PJ.ANAR LINEAR HEAT GENERATION RATE This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in 10 CFR 50.46. The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axici location and'is dependent only secondarily on the rod to rod power distribution within an assembly. The Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limits for two loop operation are specified in the CORE OPERATING LIMITS REPORT (COLR). For single-loop operation, a MAPLHGR limit corresponding to the product of the two-loop MAPLHGR and a reduction factor specified-in the COLR can be conservatively used to ensure that the PCT for single loop operation is bounded V by the PCT for two loop operation. The daily requirement for calculating APLHGR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distribu-tion shifts are very slow when there have not been significant power or control -GRAND GULF-UNIT 1 8 3/4 2-1 Amendment No. 99, ??? l - - ~ - ~ ~ ~ - ~ ~ ~ - ~ ~

3 y n. -v Pa f POWER DISTRIBUTION LIMITS BASES + MINIMUM CRITICAL POWER RATIO (Continued) During initial start-up testing of the plant, a MCPR evaluation will be made at 25% of RATED-THERMAL POWER level with minimum recirculation pump speed. The MCPR margin will thus be demonstrated such that future MCPR evaluation below this puer level will be shown to be unnecessary. The daily requirement for calculating MCPR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The requirement to calculate MCPR within 12 hours after the completion of a THERMAL POWER increase of at least 15% of RATf4 THERMAL POWER ensures thermal lim:ts ar.e met after power distribution shifts while still allotting time for the 'ribution to stabilize. The requirement for calculating MCPR

afte, determining a LIMITING CONTROL ROD PATTERN exists ensures that E 'R wil' in following a change in THERMAL POWER or power shape, that e

.ld r* ation exceeding a thermal limit. s 1 EAT GENERATION RATE

ffication assures that the Linear Heat Generation Rate (LHGR) in any _

. ss than the design linear heat generation even if fuel pellet densification is postulated. The LHGR limits are multiplied by the smaller of either the flow dependent LHGR factor (LHGRFAC ) or the power dependent LHGR factor (LHGRFAC ) g p corresponding to the existing core flow and power state to ensure adherence to the fuel mechanical design bases during the limiting transient. LHGRFAC 's are g generated to protect the core from slow flow runout transients. A curve is provided based on the maximum credible flow runout transient for Loop Manual operation. The result of a single fatiure or single operator error during operation in loop Manual is the runout of only one loop because both # recirculation loops are under independent control. LHGRFAC 's are generated to p protect the core from plant transients other than core flow increases. The. daily requirements for calculating LHGR when THERMAL POWER is greater thantor equal to 25% of RATED THERMAL POWER is sufficient since power distri-bution shifts are very slow when there have not been significant pcwer or control rod ct.nges. The requirement to calculate LHGR within 12 hours after the complethn of a THERMAL POWER increase of at-least 15% of RATED THERMAL POWER ensu es thermal limits are met after power distribution shifts while stili allotting time for the power distribution to stabilize. The requirement for calculating LHGR after initially determining a LIMITING CONTROL R^O PATTERN exists ensures that LHGR will be known following a change in THERMAL POWER or power shape that could place operation exceeding a thermal limit. 1 GRAND GULF-UNIT 1 B 3/4 2-7 Amendment No. 93, ??? a

- to GNRO 92/00093 DESIGN FEATURES .a 5.3 REACTOR CORE-EVII ASSEMBLIES 5.3.1 The reactor core shall contain 800 fuel assemblies. Each fuel assembly shall contain fuel-rods and water rods clad with Zircaloy cladding. Each fuel rod shall have e design nominal active fuel length of 150 inches. Reload fuel shall have machanical, thermal-hydraulic and neutronic characteristics compatible wt the-initial core loading. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC Staff-appreved codes and methods, shown to comply with all safety design bases, and are identified in the Core Operating Limits Report. CONTROL R0D ASSEMBLIES 5.3.2 The reactor core shall contain 193 control rod assemblies, each J consisting of a-cruciform array of stainless steel tubes containing a design nominal 143.7 inches of boron carbide, 8 C, powder surrounded by a cruciform 4 shaped stainless steel sheath. 1.4 REACTOR COOLANT SYSTEM. DESIGN PRESSURE AND' TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained: a. In accordance with the code requirement: specified in Section 5.2 of the FSAR, with allowance-for normal degradation pursuant to the applicable Surveillance Requirements, = b. .For a pressure of: 1. 1250 psig on the suction side of the recirculation pump. 2. 1650 psig from tht recirculation pump discharge to the outlet side of the discharge shutoff valve. 3. 1550 psig from the discharge shutoff valve to the jet pumps. c. For a temperature of 575'F. o VOLUME 5.4.2 The total water and steam volume of the reactor vessel and recirculation system is approximately 22,000 cubic feet at a nominal T of 533*F. ave GRAND-GULF-UNIT 1 5-5 Amendment No. 57, ??? l l

p f . to GNRO 92K;0093 t ADMINISTRATIVE CONTROLS w SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT (Continued) n c. Principal radionuclide (specify whether determined by measurement or estimate), d. - Type of waste (e.g., spent resin, compact dry waste, evaporator bottoms), e. Type of container (e'.g., LSA, Type A, Type B, large Quantity), and f. Solidification agent (e.g., cement, urea formaldehyde). The radioactive effluent release reports shall include unplanned releases from the site to the UNRESTRICTED AREA of radioactive materials in gaseous and i liquid effluents on a quarterly basis. The radioactive effluent release reports shall include any changes to the PROCESS CONTROL PROGRAM (PCP), 0FFSITE DOSE CALCULATION MANUAL (0DCM) or radioactive waste systems made during the reporting period. MONTHLY OPERATING REPORTS '6.9.1.10 Routine reports of operating statittics and shutdown experience, including documentation of all challenges to main steam system safety / relief valves,.shall be submitted on a monthly basis no later than the 15th of each month following.the calendar month covered by the report. CORE OPERATING-LIMITS REPORT (COLR) 6.9.1.11 -Core operating limits shall be established prior to each reload -cycle,,or prior to any remaining portion of a reload cycle, and shall be documented in the CORE-OPERATING LIMITS P.EPORT (COLR) for the following: a. The Average Planar Linear Heat' Generation Rate (APLHGR) for Technical Specification 3.2.1. 1 b. The Minimum Critical Power Ratio (MCPR) for Technical k, Specification 3.2.3. .c.: - The Linear Heat Generation Rate (LHGR) for Technical Specification -3.2.4. The analytical methods' used to determine the core operating limits shall be .those_ previously reviewed and approved by the NRC in the following documents. -The appropriate revision / supplement number for each document shall be identified in the COLR. -1) XN-NF-79-71(P), Exxon Nuclear Plant Transient Methodoloav for Boilina ' Water Reactors,' Exxon Nuclear Company, Inc., Richland, WA. 2)- -XN-NF-80-19(P)(A), Volume 1, Exxon Nuclear Methodoloav for Boilina Water, Reactors - Neutronic Methods for Desian and Analysis, Exxon Nuclear Company, Inc., Richland, WA. GRAND GULF-UNIT 1 6-19 Amendrrent No. 80, ??? i

l L

  • to GNRO 92/00093 Pag) 15 of 16 ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (COLR) (continued)
3) XN NF-80-19(P)(A), Volume 1, Advanced Nuclear Fuels Methodoloav for Boilina Water Reactors: Benchmark Results for the CASMO-3G/MICR0 BURN-B Calculation Methodoloav, Advanced Nuclear Fuels Corporation, Richland, WA.
4) XN-NF-80-19(P)(A), Volume 3, Eng_n Nuclear Methodoloav for Boilina Water Reactors THERMEX: Thermal limits Methodoloav Summary Description," Exxon Nuclear Company, Inc., Richland, WA.
5) ANF-913 (P)(A) Volume i, COTRANSA2: A Computer Procram for Boilina Watar Reactor Transient Analysis, Advanced Nuclear Fuels Corporation, Richland, WA.
6) ANF-ll25(P)(A), ANFB Critical Power Correlation, Advanced Nuclear Fuels Corporation, Fsichland, WA.
7) XN NF-84-105(P)(A), Volume 1, XCOBRA-T:

A Computer Code for BW8 Transient Thermal Hydraulic Core Analysis, Exxon Nuclear Company, Inc., Richland, WA.

8) XN NF 573(P), RAMPEX Pellet-Clad Interaction Eva'uation Ccde for Power Ritmpl, Exxon Nuclear Company, Inc., Richland, WA.
9) XN-NF-81-58(P)(A),R0DEX2:

Fuel Rod Thermal-Mechanical Response Evaluation Model, Exxon Nuclear Company, Inc., Richland, WA.

10) XN-NF-85-74(P)(A), R0DEX2A (BWR):

Fuel Rod Thermal-Mechanical Response Evaluation Model, Exxon Nuclear Company, Inc., Richland, WA.

11) XN-CC-33(P)(A),HUXY:

A Generalized Multirod Heate9 Code with 10CFR50 Anoendix K Heatuo 00 tion, Exxon Nuclear Company, Inc., Richland, WA.

12) XN-NF-825(P)(A), BWR/6 Generic Rod Withdrawal Error Analysis. MCPR for p

Plant Operation Within the Extended Operatina Domain," Exxon Nuclear Company, Inc., Richland, WA.

13) XN-NF-81-51(P)(A), LOCA-Seismic Structural Response of an Exxon Nuclear Comoany BWR Jet Pumo Fuel Assembly, Exxon Nuclear Company, Inc.,

Richland, WA. 14) XN-NF-84-97(P)(A), LOCA-Seismic Structural Response of an ENC 9x9 BWR Jet Pumo Fuel Assembly, Advanced Nuclear. Fuels Corporation, Richland, WA. 15) XN-NF-86-37(P), Generic LOCA Break Spectrum Analysis for BWR/6 Plants, Exxon Nuclear Company, Inc., Richland, WA. 16) XN-NF-82-07(P)(A), Exxon Nuclear Company ECCS Claddina Swellina and Ruoture Model, Exxon Nuclear Company, Inc., Richland, WA. GRAND GblF-UNIT 1 6-19a Ame..Jment No. ???

. Attachment 4 to GNRO 9200093'- j s.- Paga 16 of 16 ADMINISTRATIVE: CONTROLS

CORE OPERATING ~ LIMITS REPORT (COLR) (continued) 17)

XN-NF-80-19(A), Volumes 2, 2A, 28, & 2C, Exxon Nuclear Methodoloav fcr Boilina Water Reactors EXEM BWR ECCS Evaluation Model, Exxon Nuclear Company,_-Inc., Richland, WA. 18) XN-NF-79-59(P)(A), Methodoloav for Calculation of Pressure Drop in BWR Fuel Assemblies,. Exxon _ Nuclear Company,- Inc., Richla , WA. The core operating limits shall be determined such that a applicable limits (e.g., fuel thermal-mechanical limits, thermal-hydraulic simits, Emergency Core Cooling System (ECCS) limits,-Nuclear limits such as shutdown margin, -t'ransient analysis limits,_ and accident analysis limits) of the safety

analysis.are met.

3 The COLR, including any'mid cycle revisions or supplements, shall-be provided upon; issuance for each reload cycle to the NPC' Document Control Desk with copies to the Regional Administrator and Resident inspector. ~SPECIAL REPORTS 16;9.2 Special' reports shall be submitted to the Nuclear Regulatory Commission -pursuant to Section 50.4 of 10 CFR 50-within the time period specified for each report. 6.10-- RECORD RETENTION D In addition to the applicable record retention requirements:of Title 10, Code _ of Federal. Regulations,- the following-records shall be-retained for at least the minimum period-indicated. !6,10,1 The following records:shall-be retained for at least five years: n a. . Records and logs of unit operation covering time interval at each i Lpower level. b, ' Records and logs i principal. maintenance activities, inspections, u repair and replacement of principal items of equipment related to __ nuclear safety;- i c. All REPORTABLE EVENTS. 3 GRAND' GULF-UNIT 1. 6-19b Amendment No. ??? -. 4,.. L

d , I to GNRO-92/00093 h;l 1 -I - CORE OPERATING LIMITS REPORT (COLR) - Example i (Information Only) 1 J 6

4* i 4 Att:chment 5 to GNRO 92/00093 Pago 1 cf 13 p V CORE OPERATING LIMITS REPORT (Example) GRAND GULF NUCLEAR STATION - CYCLE 6 INTRODUCTION: This Core Operating Limits Report for Grand Gulf Nuclear Station is prepared in

accordance.with Technical Specification (TS) 6.9.1.11. The core operating limits in this report were developed using NRC approved methods in accordance with TS 6.9.1.11.

The cycle-specific core operating limits for the following Grand Gulf Nuclear Station, 1 Unit 1~ Technical Specifications are included in this report: a. The' Average Planar Linear Heat Generation Rate (APLHGR) limits for each fuel type for both two-loop and single-loop operation. (Technical Specification 3.2.1) f' b. The Minimum Critical Power Ratio (MCPR) operating limit including the ) power, flow and exposure dependent curves. (Technical Specification 3.2.3) c. - The Linear Heat Generation Rate (LHGR) limit for each fuel type including the power and flow dependent parametric adjustment factor curves, LHGRFAC and LHGRFAC, respectively. (Technical Specification 3.2.4) g p GENERAL REFERENCES-1)' MAEC 88/0313, Generic Letter 8816, " Removal of Cycle Specific Parameter Limits from Technical Specifications"-

2). EMF-91 169, Revision 1, Grand Gulf Unit 1 Cvele 6 Reload Analvsis," Siemens Nuclear Power Corporation,' July 1992.

- 3)L ' EMF 91-168, Revision 2, Grand Gulf Unit 1 Cvele 6 Plant Transient Analvsis." - Siemens Nuclear Power Corporation, September 1992. METHODOLOGY REFERENCES (per TS 6.9.11.1) r -1) XN-NF 79 71(P), Revision 2, Exxon Nuclear Plant Transient Methodoloav for Boilina Water Reactors, Exxon Nuclear Company, Inc., Richland, WA, November -1981. - 2) - XN.NF-80-19(P)(A), Volume'1 and Supplements 1 and 2, Exxon Nucleat Methodoloov for Boilina Water Reactors - Neutronic Methods for Desian and Analysis; Exxon Nuclear Company, Inc., Richland, WA, March 1983. 2

1 Att:chrntnt 5 to GNRO 92/00093 Papa 2 of 13 CORE OPERATING LIMITS REPORT (Example) - GGNS CYCLE 6 Page 2

3) XN-NF 80-19(P)(A), Volume 1 Supplements 3 and 4,. Advanced Nuclear Fuels Methodoloav for Boilina Water Reactors: Benchmark Results for the CASMO 3G/MICROBURN-B Calculation Methodoloav. Advanced Nuclear Fuels Corporation, Richland, WA, November 1990.
4) XN-NF-80-19(P)(A), Volume 3, Revision 2, Exxon Nuclear Methodoloov for Boilina Water Reactors THERMEX: Thermal Limits Methodoloav Summary Descriotion," Exxon Nuclear Company, Inc., Richland, WA, January 1987.
5) ANF 913 (P)(A) Volume 1, Revision 1 and Volume 1 Supplements 2,3 and 4, COTRANSA2: A Comouter Procram for Boilinn Water Reactor Transient Analysis, Advanced Nuclear Fuels Corporation, Richland, WA, August 1990.
6) ANF-1125(P)(A) and Supplements 1 and 2, ANFB Critical Power Correlation.

Advanced Nuclear Fuels Corporation, Richland, WA, April 1990.

7) XN NF 84-105(P)(A), Volume 1 and Supplements 1 and 2, XCOBRA-T: A Comouter Code for BWR Transient Thermal Hydraulic Core Analysis, Exxon Nuclear Company, Inc., Richland, WA, February 1987.

P) XN-NF-573(P), RAMPEX Pe.!!st Clad interaction Evaluation Code for Power Ramos, Exxon Nuclear Company, Inc., Richland, WA, May 1982.

9) XN-NF 81-58(P)(A) and Supplements 1 and 2, Revision 2, RODEX2: Fuel Rod Thermal-Mechanical Resoonse Evaluation Model. Exxon Nuclear Company, Inc.,

Richland, WA, March 1984,

10) XN-NF-85 74(P)(A), RODEX2A (BWR): Fuel Rod Thermal-Mechanical Resoonse Evaluation Model. Exxon Nuclear Company, Inc., Richland, WA, August 1986.
11) XN CC 33(P)(A), Revision 1. HUXY: A Generalized Multirod Heatuo Code with 10CFR50 Anoendix K Heatuo Ootion, Exxon Nuclear Company, Inc., Richland, WA, November 1975.
12) XN-NF 825(P)(A) Supplement 2, BWR/6 Generic Rod Withdrawal Error Analvsis.

MCPR for Plant Ooeration Within the Extended Ooeratina Domain " Exxon p Nuclear Company, Inc., Richland, WA, October 1986.

13) XN NF-8151(P)(A), LOCA-Seismic Structural Resoonse of an Exxon Nuclear Comoanv BWR Jet Pumo Fuel Assembiv, Exxon Nuclear Company, Inc.,

Richland, WA, May 1986.

14) XN-NF-84-97(P)(A), LOCA Seismic Structural Resoonse of an ENC 9x9 BWR Jet Pumo Fuel Assembh, Advanced Nuclear Fuels Corporation, Richland, WA, August 1986.

p y{4 c. Attachmont 5 to GNRO 92/00093 ~ Page 3 of 13 w g,g , CORE. OPERATING LIMITS REPORT (Example)'- GGNS CYCLE 6 Page 3 guv l 'k '

15) XN NF 86 37(P), Generic' LOCA Break Soectrum Analysis for BWR/6 Plants,

' ExxonL Nuclear Company, Inc., Richland, WA, April 1986. .16) XN NF 82 07(P)(A), Revision 1, Exxon Nuclear Comoany ECCS Claddina ' Swellina and Ruoture Model, Exxon Nuclear Company, Inc., Richland, WA, L November.1982.

17) XN NF-8019(A), Volumes 2, 2A,.2B, & 2C, Exxon Nuclear Methodoloav for.
Boilina Water Reactors EXEM BWR ECCS Evaluation Model. Exxon Nuclear Company, Inc., Richland, WA, September 1982.
18) XN NF-79 59(P)(Ali Methodoloav' for Calculation of Pressure Droo in BWR Fuel-Assemblies, Exxon Nuclear Company, Inc., Richland, WA, November 1983, t

t i e I a y_ -. c k -e, k' 4 ( ..., -. ~..

B :c' ~' to GNRO 92/00093 Page 4 of-13 i-CORE OPERATING LIMITS REPORT (Example) - GGNS CYCLE 6 Page 4 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (TS 3.2.1) During two-loop operation, all AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits shown in Figure 3.2.1 1, During single loop operation, the APLHGR for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limit shown in Figure 3.2.1-1 multiplied by 0.86.

~ Attachm:nt 5 to GNRO 92/00093 Pass 5 of 13 4 - CORE OPERATING UMITS REPORT ' Example) - GGNS CYCLE 6 Page 5 { ? k a p e l n \\ s I .{: I\\' j ^' a g -t a 1 l i I, - i e lg t- -1 L1 i i g g =- a Q#AM0 blOKleMN

gy, - g wo ff,yC f?f w Attachm:nt 5 to GNRO 92/00093'

.,, <-c l /p
fr

~- . Page 6 cf 13 '- a y j#. Syd -- CORE.' OPERATING UMITS REPORT (Example) - GGNS CYCLE 6 = - Page 6 v ; 2 kg ~ ~ jys_ r _ (Y . MINIMUM CRITICAL POWER RATIO 1TS 3.2.3). ds? > 1 6: 8 D, i 19 d7 HThe MINiMOM CRITICAL POWER RATIO (MCPR) shall be equal to or greater than the EL y.. : exposu(,re;as shown in Figure lMCPR ~ MCPR, and MCPRglimits at the ind p F +' ja: 3.2.3 1/ 3.2.3 2, and 3.2.3-3. 1 = 9e i 'h C Y' h ';[-) % y __. s r_c e/ f g W, (!. d,_ _. c.y

k Y.

4 ^ ?; .; y s 4lIc5 l .j f s + 1 s 4 - i1 h %+ .r : 9 i@-, 7p) z, a w - g V f[,'. ' ' ~ h_{-[ i, =;6

$W*:*: Attachm:nt 5 to GNRO 92/00093

I Page 7 of 13 -

i-3 CORE OPERATING' LIMITS REPORT (Example) - GGNS CYCLE 6 Page 7 y. I ,L g -= .T g y t g ) n g 9 m I 8 1 L t l t e O 9. 9 -9 9 9 9 d iWdOM d -.m

.] AttachmInt 5 to GNR0 92/00093 h Pags 8 of 13 CORE OPERATING LIMITS REPORT (Example) - GGNS CYCLE 6 Page 8 A. ..g.. ....ee.. ..+.s.g.<4 a**' i 1 i d 1.~~ - -- +- I b I f 4 ~ t a s a I i l g y I I' i 'j i..: a s F g y i~ S .w. N-R N j. dyg i

~. _ ' ;fy; Yo;; ,, to GNRO 92/00093 l' Page 9 of 13

.7

~ CORE, OPERATING LIMITS REPORT (Example) - GGNS_ CYCLE 6 Page 9 6 + dh, ? I e I I e e i g 4 x,- n n -e ,e. r -- SWdOW a. 1

n

  • to GNRO 92/00093

? Page 10 of 13 e CORE OPERAT!NG LIMITS REPORT (Example) - GGNS CYCLE 6 Page 10 LINEAR HEAT GENERATION RATES (TS 3.2.4) The LINEAR HEAT GENERATION RATE (LHGR) shall not exceed the limits shown in Figure 3.2.41 as multiplied by the smaller of either the flow-dependent LHGR factor (LHGRFAC ) of Figure 3.2.4-2, or toe power dependent LHGR factor (LHGRFAC ) of 9 p Figure 3.2.4 3. h

g.. --.-q c- ~ Attachmsnt 5 to GNRO 92/00093-fr Page 11 of 13 ?? ' CORE OPERATING LIMITS REPORT (Example) - GGNS CYCLE 6 Page 11 8 I 4 j i i - [i y-3 1 i 2 _ 7 -- w 1 1 1 l l e L i/ I s [ I \\ n . g.. g_ s/ \\ 1 .9 m 3 C e e 3 d-8 f e 3.- -3 3_ g 3-e e a 4

' '... - o. Attachmsnt 5 to GNRO 92/00093 Pags 12 of 13 , CORF. OPERATING LIMI', REPORT (Example) - GGNS CYCLE 6 Page 12 R I -. ~... - ..9-t l I I 3 i i....g_. + 3 W i 8 ~ S a g i e i R s. 8cvacm

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rz, w '3 Hag 4 -* to GNRO 92/00093 g," Paga 13 of 13 f CORE OPERATING LIMITS REPORT (Example) - GGNS CYCLE 6 Page 13 4 I ? f i 1 I gn QW >= 4 5 B* g. w 5 g M I I aE ? i g s i g N m M i / / l t U d R E h e O.- I +I I I .d DVdWOH1 a}}

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