Text

Forwards Draft Tech Specs for Power Distribution Limits,Rcs, ECCS & Plant Sys as Part of Util Involvement W/Bwr Owners Group as BWR-6 Lead Plant
	ML20043A965
Person / Time
Site:	Grand Gulf
Issue date:	05/17/1990
From:	Cottle W; SYSTEM ENERGY RESOURCES, INC.
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	AECM-90-0089, AECM-90-89, NUDOCS 9005230350
	Download: ML20043A965 (186)
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Port Gibson, MS 39150 Tul 601437 6809 ,

William T. Cottle

%ce Pteudom

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U.S. Nuclear Regulatory Commission Hail Station P1-137 Washington, D.C. 20555 ,

Attention: Document Control Desk Gentlemen:

SUBJECT:

Grand Gulf Nuclear Station Unit 1 Docket No. 50-416

,- License No. NPF-29

! Improved Technical' Specification

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Development Program AECM-90/0089 t

System Energy Resources, Inc. (SERI) is currently involved with the General Electric Boiling Water Reactor (BWR) Owners Group as the BWR-6 lead i plant for the development cf Improved Technical Specifications (ITS). SERI is

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in the process of developing a plant specific technical specification (PSTS)

.and subsequent license amendment application based on the BWR Owners' Group

' Improved-BWR Technical Specifications (NEDC-31681). The process involves review of the PSTS by a team of individuals from nuclear and design engineering, licensing and plant operations organizations.as well as review by tlie Plant . Safety Review Committee and the Safety Review Committee.

In recent discussions with the NRC-0TSB, SERI was requested to provide to

-the Staff preliminary drafts of the PSTS-in order to facilitate the Staff's validation of the BWR Owners' Group ITS. Pursuant to that request, SERI.is providing for your information and preliminary review draft technical specifications for the Power Distribution Limits (3.2), Reactor Coolant System ;

(3.4),' Emergency Core Cooling Systems (3.5), and Plant Systems (3.7) prepared '

under the SERI program for Development of Improved Plant Specific Technical Specifications for Grand Gulf Nuc1 car Station (GGNS).

Along with each Liniting Condition for Operation (LCO), you will find

1) A Revision Summary Sheet which describes the changes from the current GGNS Technical Specification to the PSTS and 2) A draft bases section for each LCO.

i This submittal.is made, of course, with the understanding that the drafts provjded are only for information at this time and that formal review of the license amendment within SERI has not been completed. Changes, therefore, are likely.to occur as the formal application for an amendment is reviewed and certified.

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It is our understanding that SERI and, the NRC staf f wil) meet the week of-

< June-11, 1990 to discuss the results of the NRC-0TSB review.

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Yours truly,

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-MTCimtc; l' Attachment cc: Mr. ' D.iC. . llintz (w/a)

'Mr. T.'ll. Cloninger.(w/a).

(. Mr. R. B. McGohee'.(w/a)

Mr. N.'S, Reynolds (w/a)-

Mr..ll. L. Thomas.(w/o)

L Hr.11. O. Christensen (w/a) b u e Mr. Stewart D. Ebneter'(w/s)

R Regional Administrator

-U.S. Nuclear Regulatory Commission

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Region II E[. ~101-Marietta St., N.W., Suite 2900 Atlanta', Georgia 30323 p .

p Mr.- L. L. Kintner,. Project Manager (w/a)

-Office.of Nuclear Reactor Regulation

i;. .U.S. Nuclear Regulatory Commission

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s. Washlagton, D.C. 20555 .,

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' ' SYSTEM ENERGY RESOURCES,-INC. !

GRAND. GULF NUCLEAR STATION e

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l! TECHNICAL' SPECIFICATIONS: IMPROVEMENT PROGRAM PLANT SPECIFIC TECHNICAL' SPECIFICATIONS 3 L

CHAPTERS 3.2', 3.4, 3.5 AND 3.7 l '- r l

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, REVISION

SUMMARY

SHEET CATEGORY KEY l

A. CATEGORIES 1

a 1. ' ADMINISTRATIVE - a change which is editorial in nature, involves the- movement of requirements within the Technical Specifications without affecting their-technical' content, simply reformats a requirement, or- 1 clarifies the Technical Specification (such as deleting a footnote no longer applicable due to a technical change s to a requirement).

2. RELOCATED - a change which moves requirements from the Technical Specifications to ~.the Bases, Ethe UFSAR, ,

procedures or~other documents.

3A. TECHNICAL CHANGE, NORE RESTRICTIVE - a change which adds a requirement:to the Technical-Specifications or revises ;

an existing requirement to be more stringent.

3B. TECHNICAL CHANGE, LESS RESTRICTIVE - a change which revises an '. existing- requirement such that more ;

. restoration / completion time is- provided .or fewer

compensatory measures are necessary.

4. DELETED - a change which removes requirements-from the

' Technical Specifications without being relocated and ,

without an adequate justification in the BWROG comparison i document. Most of the changes in this category are expected to be GGNS-specific requirements which are not ,'

in the BWR/6 Standard Technical Specifications.

Justification can be provided'to support-deletion of the 'r requirement or a recommendation made to place the requirement back into the Technical Specifications or to i relocate the requirement to another controlled document as discussed in A.2 above. .,.

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. B. CONVENTIONS

1. A change in which a requirement is moved ~from.an LCO to an LCO other than-its associated LCO in the proposed Tech

' Specs ~will be included in two LCO-review packages (e.g., !

a . requirement moved from LCO 3.1.2 to - LCO 3. 3.1 - will-appear in' both packages). If._ the change is .

ADMINISTRATIVE, it will' appear as atCategory 1 change'in !

both packages. If the - change involves a " TECHNICAL CHANGE,:it will appear as a Category 3A'or'3B-change in

-the LCO package associated with its new. location and as a Category 1-change in the LCO package for its' previous' location. ..This convention will result in~ the change only being technically justified one time. -

2. References to-the existing Tech Specs and the proposed-Tech Specs can be distinguished as follows:

a. . CONDITIONS, REQUIRED ACTIONS and COMPLETION TIMES l always refer to the proposed Tech Specs. !

, b. ACTION or ACTIONS always refers to the existing Tech p specs.

.c.. SR 3.x.x always refers to the proposed Tech Specs while SR.4.x.x always refers to the existing Tech.

Specs.

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CHAPTER-3.2 POWER DISTRIBUTION LIMITS i TABLE OF CONTENTS

- 3.2.1 AVERAGE PLANAR LINEAR-HEAT GENERATION RATE 3.2.2 MINIMUM CRITICAL POWER RATION 3.2.3 LINEAR HEAT GENERATION RATE i

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U Grand Gulf: Nuclear Station 4

. Technical Specification Improvement Program Revision Summary Sheet a

Proposed LCO/Section:- 3.2.1 Rev. 1 APLHGR ,

, 11g Chance Descriotion fJLtegory 1 LCO 3.2.1 is reformatted from LIMITING CONDITION: 1 :s FOR OPERATION 3.2.1.

2 'The APLHGR limits are relocated to the Current 2. -

Cycle Safety Analysis.

3 The applicability wording is revised to-remove I the MODE 1 reference.since it is implicitly derived ,

from the power condition.

4 CONDITIONS A and B are reformatted from the 1 ;!

ACTION statement.

5 The 15 minute limit to initiate corrective action 38 ;

specified in the ACTION statement is deleted because the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months limit to restore-the parameter within limits

.is considered to be adequate given the low probability of a. transient or accident occurring during this' interval.

6 SR 3.2.1.1 is. reformatted from SR 4.2.1.a and SR- 3B :'

4.2.1.b except that a surveillance'is required once..

within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after exceeding 25% RTP instead of at 15% power plateaus.

7 SR 4.2.1.c is deleted. Operation with APLHGR 3B equal to its-limit is highly unlikely since margin '

to.the limit is routinely maintained so an increased:

surveillance frequency is unnecessary.

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L 8 SR 4.2.1.d is deleted based upon the "once 1 j within" provision added to SR:3.2.1.1. ~

9 SR 4.2.1.b is deleted. The power increase 38 surveillance has been interpreted differently u throughout the industry. The daily surveillance l 'in conjunction with testing within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after l exceeding 25% RTP is considered adequate monitoring.

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10 A CROSS REFERENCE is added. 1 o . ,

d APLHGR 3.2.1- 1 l

[ 3. 2 POWER DISTRIBUTION LIMITS

-3.2.1 NMEtRE;d.tLMIE l

LCO 3.2.1 All AVERAGE PLANAR LINfAR HEAT GENERATION RATES (APLHGRs) shall be less thta o uual to the limits specified in the- 1 CURRENT-CYCLE SAfliY ANALYSIS.

-APPLICABILITY: THERMAL POWER 2 25% of RTP, .

ACTIONS ,

CONDITION REQUIRED ACTION, COMPLETION TIME

.i A. Any APLHGR greater A.1 Restore-APLHGP. to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months than the required- less than or equal to limits. the required limits.

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B.- Required Action and B.1 Reduce THERMAL POWER to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion < 25% of RTP.

Time of Condition A !

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SURVEILLANCE RE0VIREMENTS SURVEILLANCE FREQUENCY l' ,

SR- 3.2.1.1 Verify all APLHGRs are less than or equal Once within to the required limits. -12 hours after

> 25% of RTP l

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Once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter l.

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APLHGR.

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.3.2.1 CROSS-REFERENCES ,

TITLE NUMBER- ,

Recirculation loops Operating 3.4.1 ,

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APLHGR B 3.2.1 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 AVERA BASES BACKGROUND is a The AVERAGE measure of the average PLANAR LINEAR linear heat HEAT generation rate GENERATION of a RATE (AP fuel rods in a fuel assembly at any axial location. Limits on APLHGR are specified to assure that the fuel design limits identified in Reference I will not be exceeded during anticipated operational occurrences and that the peak cladding temperature (PCT) during the postulated design basis loss of-coolant accident (LOCA) will not exceed the limits specified in 10 CFR 50.46.

APPLICABLE The analytical methods and ascumptions used in evaluating the SAFETY fuel design limits are presented in References 1 and 2. The ANALYSES analytical methods and assumptions used in evcicatiag Design Basis Accidents, anticipated operatto ul transients and normai

, operation that determine the APLHGR limits are presented in References 1, 2, 3 and 4, Fuel design evaluations are perforaed to demonstrate that the cladding 1% plastic strain ar,d other fuel design limits described in Reference 1 are not exceeded during anticipated operational occurrences for operatien with LINEAR HEAT GENERATION Rt.tES (LHGR's) up to the operating limit LHGR.

APLHGR limits are equivalent to the LHGR limit for each fuel rod divided by the local Peaking factor of the fuel assembly.

APLHGR limits are developed as a function of exposure and the various operating core flow and power states to ensure adherence to fuel design limits during the limiting anticipated operational occurrences (Ref. 4). Flow dependent APLGHR limits are determined using the three dimensional BWR simulator code (Ref. 5) to analyze slow flow runout transients. The flow-dependent multiplier, MAPFAC,, is dependent on the maximum core flow runout capability. MAPFAC, curves are based on the maximum credible flow runout transient for Loop Manual and Non loop Manual operation. The result of a single failure or single operator error during loop Manual operation is the runout of only one loop because both recirculation loops are under independent control. Non Loop Manual operational modes allow simultaneous runout of both loops because a single controller regulates core flow.

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Grand Gulf - Unit 1 B 3.2-1 DRAFT B 11/21/89

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APLHGR l B 3.2.1 ;

BASES feontinued)

APPLICABLE Based on analyses of limiting plant transients (other than core !

SAFETY flow increases) over a range of power and flow conditions, ANALYSES power-dependent multipliers (MAPFAC,) are also generated. Due (continued) to the sensitivity of the transient response to initial core flow levels at sower levels below that where turbine stop valve :

closure and tursine control valve fast closurf scram trips are ,

bypassed both high and low core flow "^^T".CW imits areWitPFAC }j:

provided for operation at power levels between 25% of RATE] M >

THERMAL POWER and the previously mentioned bypass power level. 4 i

The exposure dependent APLHGR limits are reduced by MAPFAC and MAPFAC, at various operating conditions to ensure that all, fuel design criteria are met for normal operation and anticipated operational occurrences. A complete discussion of the analysis is provided in References 1 and 4.

LOCA analyses are performed to ensure the above determined APLHGR limits are adequate to meet the PCT and maximum oxidation limits of 10 CFR 50.46. The analysis is performed l using calculational models which are consistent with the r requirements of 10 CFR 50, Appendix K. A cnmplete discussion ;

of the analysis cede uted in the analysis is provided in ,i

, Refsrence 6. The PCT following a postulated LOCA is primarily ,

a function of the averaga heat generation rate of all the rods :

of a fuel assembly at any exial incation and is not atrongly inf1Rn:ed by the rod to rod power distribution within an issembly, ine APLHGR limits specified are equivalent to the -

LHGR of the highest powered fuel rod assumed in the LOCA 1 analysis divided by its local peaking factor. The LOCA i- analysis was perfntmed at conservatively higher APLHGR values - ,

relative to the requirements of 10CFR50.46. Appendix K.

For single recirculat4n loop vperation, the MAPFAC multiplier .

is limited to a maximum of 0.66 (Ref. 3). This is due to the conservative analysis assumption of earlier departure from nucleate =m:r boiling with one recirculation loop available, resulting l in a more severe cladding heatup during a LOCA.

APLHGR satisfies the requirements of Selection Criterion 2 of the.NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 7.

L l c.uracm the sutty Muse .

LCO The APLHGR limits specified in the 00" EPORT are the result of the fuel $es : 0"E"'T:"O Ll", TSign analysis basis accident and transient analysis. For two recirculation ,

loops operating, the limit is determined by multiplying the smaller of the MAPFAC, and MAPFAC, factors times the exposure (continued) l l

7 Grand Gulf Unit 1 B 3.2-2 DRAFT B 11/21/89

APLHGR B 3.2.1 ,

. BASES (continued) ,

LCO dependent APLHGR limits. For operation with only one :

(continued) recirculation loop in operation, in conformance with the requirements of LCO 3.4.1, the limit is determined by f multiplying the exposure dependent APLHGR limit times the ;

or 0.86, where 0.86

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smaller of either MAPFAC,, MAPFAC has been determined by a specific,, single recirculation loop analysis (Ref. 4).

APPLICABILITY The APLHGR limits are primarily derived from fuel design evaluations, LOCA and transient analysis that are assumed to r occur from high power level conditions. Design calculations ,

(Ref. 4) and operating experience have shown that as power is reduced, margin to required APLHGR limits increases. This trend continues down to low power levels where entry into MODE 2 occurs. When in MODE 2, the Intermediate Range Monitor (IRM) scram function will provide prompt scram initiation during any significant transient thereby effectively removing any APLHGR limit compliance concern in MODE 2. Therafore, at THERMAL POWER levels less than or equel to 25% of RTP, the reactor will be operating with substantiel margin to APLHGR ,

limits and the specification is not requir k ,

v ACTIONS. M -

Should any APLHGR exceed the required limits, an initial ;

condition of the design basis tecidert end transient anslyses may not be met. Therefore, prompt action should be taken to restore the APLHGR's to withir, the required limits such that the plant will be operating within analyzed conditions and within design limits of the fuel rods. .

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If the APLHGR cannot be restored to within the required limits ;

in two hours, it is required to reduce THERMAL POWER to < 25%

of RTP. As discussed in the Bases for Applicability, operation below 25% of RTP results in sufficient margin to the required limits.

Comoletion Times The Completion Times are based on industry accepted practice and engineering judgement considering the time to reasonably complete the Required Action. ,

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APLHGR B 3.2.1 :

BASES (continued) l SURVEILLANCE SR 3.2.1.1 REQUIREMENTS l APLHGR's are required to be initially calculated within 12 j hours after THERMAL POWER has exceeded 25% of RTP and then ;

daily thereafter. They are compared to the s)ecified limits to l assure that the reactor is operating within tie assumptions of ;

the safety analysis. The daily requirement for calculating '

APLHGR when THERMAL POWER is greater than or equal to 25% of I RATED THERMAL POWER is sufficient since powm' distribution shifts are very slow when there have not been significant power or J control rod changes. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance after exceeding 25%

of RTP is acceptable given the large inherent margin to operating limits at low power levels.

REFERENCES 1. GGNS 1 Current Cycle Safety Analysis.

2. Grand Gulf FSAR, Chapter 4.

3. Grand Gulf FSAR, Chapter 6.

4. Grand Gulf FSAR, Chapter 15 (including Appendices 150 and :

150, 5, XN NF-8019(P)(A), " Exxon Nuclear Methodology for Boiling Water Reactors", Volume 1, June, 1980. :

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5. XN-NF 8019(A), ' Exxon Nuclear Methodology for Boiling Water Reactors: EXEM ECCS Evaluation Mcdel, " Volume 2, Revision 1,

June, 1981.

7. NEDO 31466, " Technical Specification Screening Criteria Application and Risk Assessment," November 1987.

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Grand Gulf Nuclear Station Tcchnical Specification Itprovement Program :

Revision Summary Sheet Proposed LCO/Section: 3.2.2 Rev. L EP.3 -

lta Chance Descriotion Cateoorv ;

1 LCO 3.2.2 is reformatted from LIMITING CONDITION 1 !

FOR OPERATION 3.2.3 except as noted below. ,

2 The MCPR limits are relocated to the Current Cycle 2 I Sefety Analysis.

3 The applicability wording is revised to remove 1 ,

the MODE 1 reference since it is implicitly derived from the power condition. ,

4 CONDITIONS A and B are reformatted from the 1 ACTION statement except as noted below.

5 The 15 minute limit to initiate corrective action 3B specified in the ACTION statement is deleted because '

the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months limit to restore the limit is considered to be adequate given the icw probability of a transient or accider.t occurring during this 1.terval.

6 SR 3.2.2.1 is reformM ted from SR 4.2.3.a and- 39 SR 4.?.3.6 except that a serveillance is required once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after axceeding 25% RTP instead of at 15% power pisteaus.

7 SR 4.2.3.c is deleted. Operst;or with MOPR equd 3B :

to its limit is highly wiikely since margin to the -

5imit is routinely maintained so en increased surveillance frequency is unnecessary. -

8 SR 4.2.3.d is deleted based upon the "once I within" provision added to SR 3.2.2.1.

9 SR 4.2.3.b is deleted since the power increase 3B surveillance has been interpreted differently throughout the industry. The daily surveillance in conjunction with testing within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after exceeding 25% RTP is considered adequate monitoring.

10 CP.0SS REFERENCES are added. 1 l

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i MCPR 3.2.2 3.2 POWER DISTRIBUTION LIMITS 3.2.2 MINIMUM CRITICAL POWER RATIO I LCO 3.2.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be greater than or equal to the MCPR limit specified in the C^^E ^."E"/T!"C LI"ITC 0:^0%T. CURRENT (NC.t.E MFETy A M ALNS\$.

APPLICABILITY: THERMAL POWER 2 25% of RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. MCPR less than the A.1 Restore MCPR to greater 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months required limit, than or equal to the required limit.

B ., Required Action and B.1 Reduce THERMAL POWER to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Comeletion < 25% ef RTP, ,

Ti:ne af Condition A ;

not met.

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SURVEILLANCE FREQUENCY I

SR 3.2.2.1 Verify MCPR is greater than or equal Once within to the required limit. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after

> 25% of RTP i

8!E Once per

'24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months -

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MCPR 3.2.2 ,

I CROSS-REFERENCES :

TITLE NUMBER :

Control Rod Scram Times 3.1.3 End of Cycle Recirculation Pump Trip Instrumentation 3.3.4.2 ;

4 Recirculation loops Operating 3.4.1 l

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B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 NINIMUM CRITICAL POWER RATIO BASES BACKGROUND The MINIMUM CRITICAL POWER RATIO (MCPR) is a measure of the operating fuel assembly power relative to the fuel assembly i power that would result in the onset of boiling transition.

The Safety Limit MCPR is set such that 99.9% of the fuel rcds ,

will avoid boiling transition if the limit is not violated (refer to the Bases for LCO 2.1.2). For the purpose of establishing reactor operating limits, damage of the fuel rod cladding is assumed to occur, although fuel damage would not i necessarily occur if a fuel rod actually experienced boiling transition (Ref. 1). The operating limit MCPR is established to assure that the safety limit is not exceeded during anticipated operational occurrences. .

The onset of transition boiling is a phenomena that is readily y detecteo during the tssting of various bundle designs. Based on this experimental data, correlations have been developed ;

i.e., tha that bundle arepower used level to predict at th:critical bundle enset of power transition bo (iling) for a given ret of plent parametars (e.g., pressure, mass flux, ;

subcoolir.g, etc.). Since plant operating conditions and bundle i power levels are relatively easily monitored and determined, scnitoring MCPR is a convenient way of ensuring that fuel e failures dus to iaaoequate "coling do not occur. !

APPLICABLE The analytical methods and assumptions used in evaluating :

SAFETY the anticipated operational occurrences to establish the ANALYSES operating limit MCPR are presented in the FSAR, Chapters 4.

6 and 15 and in References 2, 3, 4 and 5. To assure that the !

Safety Limit MCPR is not exceeded during any moderate frequency transient event, limiting transients have been analyzed to determine f a largest reduction in Critical Power Ratio (CPR).

The type or iransients evaluated are loss of flow, increase in pressure ar6 power, positive reactivity insertion and coolant '

temperature Jecrease. The limiting transient yields the largest *CPR. When the largest .CPR is added to the Safety Limit MCPR, the required operating limit MCPR is obtained. t The MCPR operating limits derived from the transient analysis are dependent on the operating core flow and power state (MCPR, -

and MCPR, respectively) to ensure adherence to fuel design *

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MCPR B 3.2.2 BASES (continued)

APPLICABLE limits during the worst moderate frequency transient (Ref. 3, 4 SAFETY and 5). Flow dependent MCPR limits are determined by steady ANALYSES state thermal hydraulic methods, using the three dimensional BWR (continued) simulator code (Ref. 6) and the multi-channel thermal hydraulic

" code (Ref. 7). MCPR, curves are provided based on the maximum credible flow runout transient for Loop Manual and M;;h;; Non La.p Manual Operations. The result of a single failure or operator error during Loop Manual operation is the runout of one loop because both loops are under independent control. Both loops N"*N can runout durinRM;;h;; Manual operation because a single controller regulates core flow.

Power dependent MCPR limits (MCPR ) are determined by the three and the one-dimensional dimensional BWR simulator . Duecode to t (R,ef. 6)he sensitivity of the BWR transient transient code response to (Ref.10)l initia core flow levels at power levels 66 that where the turbine stop valve closure and turbine control r valve fast clos'Jrt, scram trips are bypassed, a high and low flow '

operating limit MCPR is provided for operating between 25% of MTEG THERMAL POWER ,(RTP) and the previodsly nmntioned b) pus power level.

MCPR satisfies the requirements of Selection Criterion 2 of the :

NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 9.

c.uaum C.%sA .5twety AuMMSW f LCO TheMCPRoperatinglimitsspecifiedintheJ^^: OP "AT!%. l LI".!TS "C."0"T are the result of the transient analysis. The operating limit MCPR is determined by the larger of the MCPR, i and MCPR, limits, t

APPLICABILITY The MCPR operating limits are primarily derived from transient t analyses that are assumed to occur from high power level conditions. Below 25% of RTP, the reactor will be operating at '

minimum recirculation pump speed and the moderator void content will be very small. Surveillance of thermal-limits below 25%

of RTP is unnecessary due to the large inherent margin that assures that the Safety Limit MCPR will not be exceeded even if a limiting transient should occur. Statistical analyses documented in Reference 8 indicate that the nominal value of

!- initial MCPR expected at 25% of RTP is in excess of 3.0.

Studies of the variation of limiting transient behavior have been performed over the range of power / flow conditions. These studies (Ref. 5) encompass the range of key actual plant parameter values important to typically limiting transients.

(continued)

Grand Gulf - Unit 1 B 3.2-6 DRAFT B 2/12/90 l

-4 -. , - - , -.- -,-- . .- - - - . .-

. . . - . - ~. . - .

MCPR B 3.2.2 BASES (continued) l l l APPLICABILITY The results of these studies demonstrate that margin is ,

(continued) expected between performance and MCPR requirements, and that j margins increase as power is reduced to 25% of RTP. This trend 1 is expected to continue to the low power range when entry into MODE 2 occurs. When in MODE 2, the Intermediate Rtnge Monitor ]

(IRM) provides rapid scram initiation for any significant power 1 increase transient which effectively eliminates any MCPR compliance concern. Therefore, at THERMAL POWER levels less i' than 25% of RTP, the reactor will be operating with substantial margin to MCPR limits and the specification is not required.

ACTIONS Ad Should any MCPR ba outside the required limits, an initial condition of the design basis transietit analyses may.not be :

met. TLerefore, orompt action should be taken to restore the MCPR's to within the required 11mits such that the plant will be operating within analyzed conditions, j L1 I if the MCPR cannot be rtstored to wi+.hin the required limits in two hents,11 is re:Nired to redute THEP. MAL POWER to < 25% of >

RI P. As discusseci in the Sases for Applicability, operation '

below 25f. of RTP results in sufficient margin to the required ',

limits.

GmDaletion Times Tho Completion Times are based on industry accepted practice ,

and engineering judgement considering the time to reasonably complete the Required Action.

t SURVEILLANCE SR 3.2.2.1 REQUIREMENTS MCPR is required to be initially calculated within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL POWER has exceeded 25% of RTP and then daily thereafter. It is compared to the specified limits to assure that the reactor is operating within the assumptions of the safety analysis. 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 1

(continued)

-Grand' Gulf - Unit 1 B 3.2-7 DRAFT B 2/12/90

i MCPR B 3.2.2 BASES (continued)

I SURVEILLANCE SR 3.2.2.1 REQUIREMENTS (continued) very slow when there have not been significant power or control J rod changes. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance after exceeding 25% of RTP ,

is acceptable given the large inherent margin to operating -

limits at low power levels.

i REFERENCES 1. NUREG-0562, " Fuel Rod Failure as a Consequence of ;

Departure From Nucleate Boiling or Dryout", June 1979.

CecrNs-n

2. W Current 3 Cycle Safety Analysis', (htet '!:rdr)

3. Grand Gulf FSAR, Appendix 150. l

4. Grand Gulf FSAR, Appendix 150. .i

5. Grtnd G:11f FSAR, Appendix 150.

-6. " EXXON Nuclear Methodology for BWRs: Neutronics methods for Desi9n and anlysis", XN NF 80-19 (P)(A), Volume 1, L (as supplemented) !

7. " EXXON Nuclear Methodology for BWRt: THERMEX, lhermal Limits Fethodology Summary Description", XN NF 80 '9(P)(A),

Volume 3, Revtsion 2, January 1987.

8. "BWR/6 Generic Rod Withdrawal Error Analysis", Appendix ISB, General Electric Standard Safety Analysis Report (GESSAR-II).

I

9. NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment", November 1987.

10. " Exxon Nuclear Plant Transient Methodology for BWRs" XN-NF-79-71(P), Revision 2, November 1981.

I .

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! Grand Gulf --Unit 1 B 3.2-8 DRAFT B 2/12/90 !

E

x Grand Gulf Nuclear Station Technical Specification Icprovcment Program Revision Summary Sheet Proposed LCO/Section: 3.2.3 Rev. l_ LtffaB lta Chance Descriotion Cateoorv 1 LCO 3.2.3 is reformatted from LIMITING CONDITION 1 FOR OPERATION 3.2.4.

2 -The LHGR limits are relocated to the Current 2 L Cycle Safety Analysis or comparable document.

3 The applicability wordir'; is revised to remove 1 the MODE 1 reference since it is implicitly derived frem the power condition.

4 CONDIlION3 A and B are reformatted frot the !

ACTION ftstement.

5 The 15 minute limit to initiate corrective action 3B specified in the ACTION statement is deleted because the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months lirait to restore the parameter is considered to be a&qucte given the low probability of a transient or acc1derst occurring during this interval, j 6 SR 3.2.3.1 is refermatted fror.t SR 4.2.4.a and 3B 4 SR 4,2.4.b except that a surveillance'is required once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after exceeding 25% RTP instead of at 15% power plateaus.

7 SR 4.2.4.c is deleted. Operation with LHGR equal 3B to its limit is highly unlikely since margin to the limit is routinely maintained so an increased surveillance frequency is unnecessary, 8 SR 4.2.4.d is deleted based upon the "once I within" provision added to SR 3.2.3.1.

SR 4.2.4.b is deleted. The power increase 9 3B surveillance has been interpreted differently throughout the industry. The daily surveillance in conjunction with testing within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after exceeding 25% RTP is considered adequate monitoring.

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LHGR i 3.2.3 r

3.2 POWER DISTRIBUTION LIMITS :

i 3.2.3 LINEAR HEAT GENERATION RATE -

LCO 3.2.3 The LINEAR HEAT GENERATION RATE (LHGR) shall be less than '

or equal to the limits specified in the CURRENT CYCLE SAFETY ANALYSIS J

APPLICABILITY: THERMAL POWER 1 25% of RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME ;

[ A. Any LHGR greater than A.1 Rertore LHGR to itst 2 teurs the required limits, than br equel to the '

, required limits.

i '

B. Required Action and B.1 Reduce THERMAL POWER to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ,

associated Completion < 25% of RTP, Time of Condition A not met.

(

' SURVEILLANCE REOUIREMENTS SURVEILLANCE FREQUENCY s

l SR 3.2.3.1 Verify all LHGRs are less than or. equal to Once within the required limits. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after

> 25% of RTP l &HQ Once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter l CROSS

REFERENCES:

None l

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Grand Gulf - Unit 1- 3.2-5 DRAFT B 11/21/89'

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LHGR B 3.2.3 8 3.2 POWER DISTRIBUTION LIMITS B 3.2.3 LINEAR HEAT GENERATION RATE l 4

BASES BACKGROUND The LINEAR HEAT GENERATION RATE (LHGR) is a measure of the heat generation rate of a fuel rod in a fuel assembly at an axial location. Limits on LHGR are specified to assure that fuel design limits will not be exceeded anywhere in the core during normal operation including anticipated operational occurrences. -

Exceedly the LHGR limit could potentially result in fuel damage and subsequent release of radioactive materials. Fuel ;

design limits are specified to assure that fuel system damage, f

, fuel rod failure or inability to cool the fuel will not occur during the ant;cf psted operati.ng conditions identified in Ref. "

[ 1. ,

[ l

, APPLICABLE. The analytical methods and assumptions used in evaluating fuel i SAFETY system design are presented in FSTR, Chapter 4 and in ANALYSES (Ref. 1). The fuel assembly it. designed to ensure (in conjunction with the core nuclear and thermal hydraulic design, plant equipment, instrumentatfon kod protection system) that. ,

fuel damage will not result in the release of radioactive materials in excess of the guidelines of 10 CFR 20, 50 and 100.

i The mechanisms transients and whichwhich arecould cause fuel considered in fue dama$e evaluations duringare operational (1) rupture of the fuel rod cladding caused by strain from the relative expansion of the 002 pellet and (2) severe overheating ,

I of the fuel rod cladding caused by inadequate cooling. A value i

! of 1% plastic strain of the zircaloy cladding has been defined as the limit below which fuel damage caused by overstraining of -

the fuel cladding is not expected to occur (Ref. 2). The SafetyLimitMINIMUMCRITICALPOWERRATIO(MCPR)ensuresthat fuel damage caused by severe overheating of the fuel rod cladding is avoided and is discussed separately in the Bases for LCO 3.2.2.

Fuel design evaluations have been performed and demonstrate that the 1% plastic strain fuel design limit is not exceeded during continuous operation with LHGR's up to the operating limit specified in the CURRENT CYCLE SAFETY ANALYSIS. The analysis also includes allowances for short term transient 1 operation above the operating limit to account for anticipated operational occurrences including consideration of densification power spiking.

(continued)

Grand Gulf - Unit 1 8 3.2 9 DRAFT B 11/21/89

LHGR !

B 3.2.3 j BASES (continued)

APPLICABLE LHGR satisfies the requirements of Selection Criterion 2 of the SAFETY NRC Interim Policy Statement on Technical Specification ANALYSES Improvements as documented in Reference 3.

(continued) 1 e

LCO LHGR is a basic assumption in the fuel design analysis. The fuel has been designed to operate at rated core power with ;

sufficient design margin to the LHGR calculated to cause 1%

cladding plastic strain. The operating limit to accomplish j this ob;ective is specified in the CURRENT CYCLE SAFETY t ANALYSIS.

APPLICA0iLITY ~ The LHGR limit it derived from fuel design analysis that is r limiting at hic'n power level conditions. At cora thermal power levels less than 25% of RATED THERMAL POWER 1 (RTP), the reactor will be o>erating with substantial margin to LHGR limits and therefore, tie specification is only required >

, when operating at or above 25% of RTP.

ACTIONS M ;

b Should any LHGR exceed the required limits, an initial

, condition of the fuel design analysis will not be met. There-fore, prompt action should be taken to restore the LHGR to within the required limits such that the plant will be l operating within analyzed conditions.

M If the LHGR cannot be restored to within the required limits in I two hours, it is required to reduce THERMAL POWER to < 25% of 3

RTP. Operation below 25% of RTP results-in sufficient margin to the required limits.

l Comoletion Times L The Completion Times are based on industry accepted practice L and engineering judgement considering the time to reasonably complete the Required Action.

(continued)

Grand Gulf - Unit 1 B 3.2-10 DRAFT B 11/21/89

LHGR B 3.2.3 BASES (continued)

SURVEILLANCE SR 3.2.3.1 !

REQUIREMENTS LHGR is required to be initially calculated within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL POWER has exceeded 25% of RTP and then daily ,

thereafter. It is compared to-the specified limits to assure !

that the reactor is operating within the assumptions of the i safety analysis. The daily requirement for calculatin LHGR :

when THERMAL POWER is greater than or equal to 25% of TED I

. THERMAL' POWER is sufficient since power distribution shifts !

are very slow when there have not been significant power or j

control rod changes. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance after exceeding 25% of RTP is acceptable given the large inherent margin to 1 operating limits at lower power levels.

)

. REFERENCES 1. CGHS-1 Current Cycle Safety Analysis (CCSA) -

2. NUREG 0800, Standard Review Plan 4.2, " Fuel System 4 Design" Section II.A.2(g). 1

3. NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment", November 1987.

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Grand Gulf - Unit 1 B 3.2-11 DRAFT B 11/21/89 i

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CHAPTER 3.4 )

REACTOR COOLANT SYSTEM

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l, CHAPTER 3.4 REACTOR COOLANT SYSTEM 1ABLE OF CONTENTS 3.4.1 Recirculation Loops Operating 3.4.2 Section Deleted 3.4.3 Jet Pumps 3.4.4 Safety / Relief Valves f 3.4.5 Operational Leakage 3 4.6 Specific Activity 344.7 Residual Heat Removal - Shutdown 3.4.8 Reactor Coolant System Pressure / Temperature Limits 3.4.9 Reactor Steam Dome Pressure

+

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v Grand Gulf Nuclear Station :

Technical Specificaticn I: prove:ent program Revision Summary Sheet Proposed LCO/Section: 3.4.1 Rev. ,_L Recire Loons coeratine ;

1133 Chance Descriotion Cateaorv 1 LCO 3.4.1 is reformatted from LIMITING CONDITION 1 FOR OPERATIONS 3.4.1.1 and 3.4.1.3.

2 The LCD statement is revised to only require two 30 ;

recirculation loops in operation or one loop in l operation under specified cotditions.

i 3 DELETED 4 The SLO loop flow limit and the SLO flow control 4 mode requirement is deleted from LCO 3.4.1.1.b. )

5 The applicability is revised to eliminate 3A 1 reference to Special Test Exception 3.10.4. i 6 Footnote '*' to pages 3/4 4-1 and 3/4 4-3 is SA deleted (see Item 5 above).

7 CONDITION B is developed from ACTION i of 3B LCO 3.4.1.1 except that ACTION i provided 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and CONDITION B provides 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

8 CONDITION C is reformated from ACTION a of 1 LCO 3.4.1.1.

9 SR 3.4.1.1 is reformatted from LCO 3.4.1.3 and 1

, SR 4.4.1.3 except_as discussed below.

10 ACTION c is deleted. 4 ,

11 ACTION b is deleted. 4 -

12 ACTION d is deleted. 4 13 ACTION e is deleted. 4 '

14 ACTION f is deleted. 4 15 SR 4.4.1.1.1 is deleted. 4 16 DELETED (See Items 29 and 30) 17 SR 4.4.1.1.3 is deleted. (See Item 24) 4 j

18 SR 4.4.1.1.4 is deleted. (See Item 14) 4 19 SR 4.4.1.1.5 is deleted. (See Item 23) 4 +

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T Grand Gulf Nuclear Station

Technical Specification Improvement Program Revision Summary Sheet Proposed LCO/Section: 3.4.1 Rev. 1 Rectre Loons Ooeratino J.lg Chance Descriotion Cateoorv 20 SR 4.4.1.1.6 is deleted. 4 21 Figure 3.4.1.1-1 is deleted. 4 22 CROSS REFERENCES are added. 1 .

23 Action h is deleted. 4 24 Action g is deleted. 4 25 Statement in LCO 3.4.1.1 specifying when operation 4 is/is not pe nissible (Reference to figure 3.4.1.1-1) is deleted. >

26 ' CONDITION A is developed from ACTION 3.4.1.3. 1 >

27 CONDITION O is added to address MODE 2 operation. 3B+

28 CONDIi!ON E is developed from ACTION b of LCO 3.4.1.3.- 1 29 SR 3.4.1.2 is reformatted from SR 4.4.1.1.2.a. 1 30 SR 3.4.1.3 is reformatted from SR 4.4.1.1.2.b. 1 LCO 3.4.1.3 uses the term " rated recirculation flow" 31 3B as opposed to the PSTS term " rated core flow" in SR 3.4.1.1.

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Recirculation loops Operating 3.4.1 3.4 REACTOR COOLANT SYSTEM 3.4.1 Recirculation Locos Goeratina ,

LCO 3.4.1 Two recirculation loops shall be in operation, E ,

One recirculation loop may be in operation provided the ,

following limits are made applicable:

A. LCO 3.2.1, APLHGR, Single Loop Operation Limits specified in the COR: 0^ RATINC :.!MITS RE"0RT.-

C.ueasuT c.Y cLE. SM ET Y (WAt.h t m .

B. LCO 3.2.2, MCPR, Single Loop Operation Limits specified in the C4RE-OPEMHNG L:"!TS CCNRT.

C.uT.9.ERT C.1 C LE 5 A r:E.T1 Au At.%B.

C. LCO 3.3.1.1, RPS Instrumentation, Function 2.b of .

Table 3.3.1.1-1, Allowable Value is reset for single loop operation.

APPLICABILITY: MODES 1 and 2.

i ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A '. During two loop A.1 Restore the loop jet - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months operation, recircu- pump flows to within the lation loop jet pump specified limits, flow mismatch outside requirements.

B. With one recirculation B.1 Satisfy the single loop. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from !

loop not in o.neration operation requirements of discovery of and limits for single the LCO. loop not in loop operation not met, operation C. No recirculation loops C.1 Place the Reactor Mode Immediately ,

in operation in MODE 1. Switch in the Shutdown position.

(continued) i l

L Grand Gulf - Unit 1 3.4-1 DRAFT B 2/01/90 .

L-l-

Recirculation Loops Operating 3.4.1

' ACTIONS (continuedl CON 0! TION REQUIRED ACTION COMPLETION TIME

0. No recirculation loops D.1 Restore one loop to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months in operation in MODE 2. operation.

E. Required Action and E.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Conditions A, B t

' or 0 not met, i

L Grand Gulf - Unit 1 3,4.la DRAFT B 2/01/90

(f Recirculation loops Operating 3.a.1 SURVEILLANCE RE0VIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1- Verify recirculation loop jet pump flow 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months mismatch is: when both loops are in A. < 10% of rated core flow when operation 3perating at < 70% of rated core flow.

B. < 5% of rated core flow when iperating at > 70% of rated core flow. ,

i SR 3.4.1.2 Demonstrate recirculation Flow Control 18 months Valve (FCV) fails "as is" on loss of ,

hydraulic pressure at the hydraulic unit.

SR 3.4.1.3 Demonstrate average rate of FCV movement is: 18 months A. s 11% of stroke per second opening.

8!!D B. 1 11% of stroke per second closing.

CROSS-REFERENCES TITLE NUMBER AVERAGE PLANAR LINEAR HEAT GENERATION RATE 3.2.1 MINIMUM CRITICAL POWER RATIO 3.2.2 Reactor Protection System Instrumentation 3.3.1.1 Reactor Coolant System Pressure / Temperature Limits 3.4.8 Grand Gulf - Unit 1 3.4-2 DRAFT B 2/01/90 l

Recirculation Loops Operating B 3.4.3 8 3.4 REACTOR COOLANT SYSTEM B 3.4.1 Recirculation loons Doeratina BASES BACKGROUND The reactor recirculation system is designed to provide a forced ;

coolant flow through the core to remove heat from the fuel. The reactor recirculation system consists of two recirculation pump ,

loops external to the reactor vessel. These loops provide the piping path for the driving flow of water to the reactor vessel jet pumps, Each external loop contains one motor driven recirculation pump, a flow control valve and associated piping, valves, and instrumentation. The recirculation loops are part of the reactor coolant pressure boundary and are located inside the drywell structure. The jet pumps are reactor vessel internals.

The recirculated coolant consists of saturated water from the steam separators and dryers that has been subcooled by incoming feedwater. This water passes down the annulus between the '

reactor vessel wall and the core shroud. A portion of the coolant flows from the vessel, through the two external recirculation loops, and becomes the driving flow for the jet

> umps. Each of the two external recirculation loops discharges ,

ligh pressure flow into an external manifold from which individual recirculation inlet lines are routed to the jet pump .

risers within the reactor vessel. The remaining portion of the-coolant mixture in the annulus becomes the suction flow for the jet pumps. This flow enters the jet pump at suction inlets and is accelerated by the driving flow. The drive flow and suction flow are mixed in the jet pump throat section. The total flow then passes through the jet pump diffuser section into the area below the core (lower plenum), gaining sufficient head in the procest to drive the required flow upward through the core, i Each recirculation loop is manually started from the control room. The recirculation flow control valves provide regulation of individual recirculation loop drive flows. The flow in each loop is manually controlled.

(continued)

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t Grand Gulf - Unit 1 B 3.4-1 DRAFT B 2/01/90 i

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y Recirculation Loops Operating ;

l B 3.4.1 !

BASES (continued) l APPLICABLE The operation of the reactor recirculation system is an initial SAFETY condition assumed in the design basis Loss of Coolant Accident t ANALYSES (LOCA) (Ref. 1). During a LOCA caused by a recirculation loop ;

pipe break, the intact loop is-assumed to provide coolant flow L during the first few seconds of the accident. .The initial core flow decrease is rapid because the recirculation pump in the -

broken loop ceases to pump reactor coolant to the vessel almost immediately. The pump in the intact loop coasts down relatively .

slowly. This pump coastdown governs the core flow response for the next several seconds until the jet pump suction is uncovered ,

(Ref. 1). The analyses assume both loops are operating at the !

same flow prior to the accident. Because an initial :

recirculation loop jet pump flow mismatch could affect the ;

transient core flow in the intact loop during pump coastdown, i flow mismatch is required to be maintained within specified limits. The recirculation system is also assumed to have sufficient flow coastdown characteristics to maintain fuel thermal margins during abnormal operational transients (Ref. 6) which are analyzed in Chapter 15 of the FSAR. ;

A plant specific LOCA analysis has been performed for Grand Gulf Unit I assuming only one operating recirculation loop. ,

This analysis has demonstrated that in the event of a LOCA '

caused by a pipe break in the operating recirculation loop, the ECCS response will provide adequate core cooling provided the -

AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) ,

requirements are modified (Ref. 2). The transient analyses of Chapter 15 of the FSAR have also been performed for single recirculation loop operation (Ref. 2) and demonstrate sufficient flow coastdown characteristics to maintain fuel thermal margins during the abnormai operational transients analyzed. During single recirculation loop operation in MODE 1, ,

modifications to the Reactor Protection System (RPS) Average >

Power Range Monitor (APRM) instrument setpoints are also -

required to account fur the different response of the reactor l and different relationships between recirculation drive flow and 1 reactor core flow.

The APLHGR and MCPR requirements for Grand Gulf Unit I also account for the effects of a slow, inadvertent increase in ,

recirculation loop flow to maximum for the two loop as well as the single loop operational conditions. 3 Recirculation loops Operating satisfies the requirements of i Selection Criterion 2 of the NRC Interim Policy Statement on i Technical Specification Improvements as documented in Reference j

3. u (continued) i 1

l Grand Gulf - Unit 1 B 3.4 2 DRAFT B 2/01/90 1

J

Recirculation Loops Operating B 3.4.6 BASES (continued)

LCO Two recirculation loops are required to be in operation with :

their flows matched within the limits specified in SR 3.4.1.1 to ensure during a LOCA, caused by a break of the piping of one recirculation loop, or during a slow runout transient, the assumptions of the applicable analyses are satisfied. With only one recirculation loop in operation, modifications to the required APLHGR limits (LCO 3.2.1), MCPR limits (LCO 3.2.2), and APRM Flow Biased Simulated Thermal Power High setpoint (LCO 3.3.1.1) may be applied to allow continued operation consistent with the assumptions of Reference 2.

APPLICABILITY Requirements for operation of the reactor recirculation system are necessary during MODES I and 2 since there is considerable energy in the reactor core and the limiting design basis transients and accidents are assumed to occur. During other ;

conditions, the consequences of an accident are reduced and the coastdown characteristics of the recirculation loops are not important.

ACTIONS L.1 During two loop operation, recirculation loop jet pump flow mismatch limits are in compliance with ECCS/LOCA and the flow -

runout transient analyses criteria. If the flow mismatch is ,

outside the specified limits, the analyses may no longer be '

bounding. Therefore, only a limited time is allowed to restore >

the flow mismatch to within acceptable limits.

L.1 A recirculation loop is considered not to be in operation when the pump in that loop is idle. Should a LOCA occur with one recirculation loop not in operation, the core flow coastdown and resultant core response may not be bounded by the LOCA .

analysis applicable to two loop o>erations. Therefore, only a limited time is allowed to make tie single loop operation l limits applicable. ,

! L.1 With no recirculation loops in operation in MODE 1, an immediate reactor shutdown is required. This requirement is a core thermal hydraulic stability restriction.

1 (continued) 1 Grand Gulf - Unit 1 B 3.4-3 DRAFT B 2/01/90 l

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Recirculation Loops Operating B 3.4.)

BASES (continued)

ACTIONS D.J (continued) >

With no recirculation loops in operation in MODE 2, a limited ;

time is allowed to restore one loop to operation. i L.1

With the two loop flow mismatch not restored within the Required ,

Com)1etion Time, the single loop requirements of the LCO not met. !

wit 11n the Required Completion Time, or a single loop not restored to operating status within the Required Completion Time, <

the reactor is required to be in MODE 3. In this condition, the i recirculation loops are not required to be operating because of the reduced severity of design basis accidents and minimal '

dependence on the recirculation loop flow characteristics. -

Comoletion Times !

All Completion Times are based on industry accepted practice and engineering judgement considering the number of available systems and the time required to reasonably complete the ,

Required Action. ,

SURVEILLANCE B_3.4.1.1 REQUIREMENTS This surveillance requirement ensures the recirculation loop flows are within the allowable limits for mismatch. At low i core flow (i.e. < 70% rated core flow), the MCPR requirements provide larger margins to the fuel cladding integrity Safety Limit such that the potential adverse effect of boiling transition during a LOCA or a slow flow runout transient is 1 reduced. A larger flow mismatch can therefore be allowed when core flow is < 70% of rated core flow. the recirculation loop jet pump flow, as used 'in this surveillance is the summation of .

l the flows from all of the jet pumps associated with a single recirculation loop. The mismatch is measured in terms of '

percent of rated core flow. This SR is not required when both 1; loops are not in operating since the mismatch limits are ;

l' meaningless during single loop or natural circulation operation. '

l Operating experience has demonstrated that a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months frequency i I for this surveillance is adequate.

SR 3.4.1.2 l 1

Loss of Hydraulic Power Unit Pressure, which provides the motive-force for the FCVs, causes the FCV to lockup in its last demanded position. This surveillance verifies this function. l l

1 (continued)-

Grand Gulf - Unit 1 B 3.4-4 DRAFT B 2/01/90 l

i Recirculation loops Operating i B 3.4.1 ,

BASES (continued) ,

SR 3.4.1,3 :

SURVEILLANCE REQUIREMENTS (continued) This surveillance requirement ensures the overall average rate of FCV movement at all positions is maintained within the analyzed limits.

Surveillance Freauencies In general, surveillance frequencies are based on industry accepted practice and engineering judgement considering the unit !

conditions required to perform the test, the ease of performing ',

the test and a likelihood of a change in the system / component status. ,

REFERENCES 1. Grand Gulf Unit 1 FSAR, Section 6.3.3.7.2.

2. Grand Gulf Unit 1 FSAR, Appendix 150,

3. NEDO 31456, " Technical Specification Screening Criteria !

Application and Risk Assessment", November 1987. >

4. Grand Gulf Unit 1 FSAR, Sectior.15.3.2. f

5. Grand Gulf Unit 1 FSAR, Section 15.4.5. .

6. Grand Gulf Unit 1 FSAR, Section 5.4.1.4.1.

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Grand Gulf Nuclear Station 4

Technical Specification improvement Program

+ Revision Summary Sheet Proposed LC0/Section: 3.4.2 Rev _1_- ELY1 1115 Chanae Descriotion Cateoorv 1 DELETED

^

2- DELETED.

3 DELETED

'4 SR 4.4.1.1.2.a is relocated to SR 3.4.1.2. 2 5 SR 4.4.1.1.2.b is relocated to SR 3.4.1.3. 2-t.

NOTE PSTS LCO 3.4.2 was deleted per CRS 244. The SRs were incorporated into LCO 3.4.1.

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3.4.2-...

'3.4 REACTOR-COOLANT SYSTEM b c4 -

,; 3.4.2-;Section Deleted >

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Grand Gulf - Unit 1 3.4-3 DRAFT B 3/07/90 i

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B 3.4.2.

- B 3.4- REACTOR COOLANT SYSTEM

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B 3.4.2 Section Deleted b

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' Grand Gulf - Unit 1 B 3.4-7 DRAFT B 3/07/90 4,

' Grand Gulf Nuclear Stationz gj Technical Specification Improvement. Program s Revision Summary. Sheet Proposed LCO/Section: 3. 4. 3 ' Rev. _1_ Jet Pumos 1123 '(hance Descriotion CateoorY 1 LCO 3.4.3 is reformatted 'from LIMITING CONDITION' 1-FOR OPERATION 3.4.1.2.

2. CONDITION A is reformatted from the ACTION 1-statement. ,

3 =SR 3.4.3.1 is reformatted from SR 4.4.1.12.1. I t

4 SR 3.4.3.1 criterion c provides a 20% diffuser to 3B+

. lower plenum check in addition to-the 10% jet pump flow check, 5 SR.4.4.1.2.2 is deleted. SR 3.4.3.1 is a "only 38 ,

required" type surveillance.

6 Footnote '*' to page 3/4 3-2 is' deleted. It 4

' involved initial values supplied by Startup test program.

.7 REQUIRED ACTION A.2, to be in MODE 4 in 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months , 3A

.was added.

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s q

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Jet Pumps 3.4.3'-

3.4 REACTOR COOLANT SYSTEM '

3.4.3 Jet Pumos LCO'3.4.3 All jet pumps shall be OPERABLE.

APPLICABILITY: MODES I and 2. 1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME-A. One or more jot pumps A.1 Be in NODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months inoperable.

AND A.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months j I

SURVEILLANCE REOUIREMENTS SURVEILLANCE FREQUENCY-SR 3.4.3.1 Verify at ieast two of the following

, criteria (A, B and C) are satisifed for D each jet pump in each operating loop:

l .....N0TE-----

A. Recirculation loop drive flow versus Only required FCV position differs by $ 10% from when greater established patterns. than 25% RTP.

B. Recirculation loop drive flow versus ,

total core flow differs by 110% from 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months established patterns. ,

L C. _ Each je'. pump diffuser-to-' lower plenum L differential pressure differs by

< 20% from established patterns.-

L 08 Each jet pump flow differs by $ 10%

L from established patterns.

1 CROSS-

REFERENCES:

None Grand Gulf - Unit 1 3.4-4 DRAFT B 2/07/90

, , Jet Pumps ,

l B 3.4.3 ce q

.l B 3.4 REACTOR COOLANT SYSTEM j i

" .^

B 3.4.3 iJet Pumos- J BASES I

BACKGROUND .The reactor recirculation system is described in the Background section of the Bases for LC0 3.4.1.

,. The jet-pumps are part of the reactor recirculation system and.

are designed to provide forced circulation through the core to -

remove heat from the fuel. The jet pumps are lor:ated in the- *

- annular region'between-the core shroud and.the vessel inner wall. Because the jet pump suction elevation is at two-thirds; core height, the vessel can be reflooded and coolant level maintained: at two-thirds core height even with the complete - '

break of a recirculation loop pipe which is located below the jet pump suction elevation.

d

- Each reactor recirculation loop contains 12 jet pumps. .

Recirculated coolant passes down the annulus between the reactor- ,

F vessel wall and the core shroud. A portion of the coolant flows from the vessel,-- through the two external recirculation loops, and becomes the driving flow for the jet pumps. Each of the two external recirculation loops discharges high pressure flow-into j an external manifold from which individual recirculation-inlet. ;

lines are routed to the jet pump risers within the reactor vessel.. The remaining portion of the coolant mixture in the annulus becomes the suction flow for the jet pumps. This flow enters the jet pump at suction inlets and is accelerated'by the-driving - flow. The drive flow and suction flow are mixed in the jet' pump throat section. The total flow then passes through the

- jet pump diffuser section into the area below the core (lower plenum), gaining sufficient head in the process to drive the '

required flow upward through the core.

t (continued) 1 Grand Gulf - Unit 1 B 3.4-8 DRAFT B 2/07/90

____._ ___ _ -___ - _- r-- *e N

i i Jet Pumps

. 'B 3.4.3 l BASES (continued)

'APPLICARLE Jet pump.0PERABILITY is an implicit assumption in the design

-SAFETY basis Loss of Coolant Accident.(LOCA) analysis evaluated in .

ANALYSES Reference 1. If a beam holding a jet pump in place fails, jet .*

pump displacement and performance degradation could occur, !

resulting in a reduction in core flow capacity and a lower core .

flooding elevation. ' Jet pump displacement and performance degradation could adversely affect the water level in the core - j during the reflood phase of a LOCA as well as the assumed 1 blowdown flow. The capability of reflooding the-core to two-thirds core height is dependent upon'the structural integrity of the jet pumps. ;

a Jet' Pumps satisfy the requirements of Selection Criterion 3 of the NRC Interim Policy Statement on Technical Specification .

Improvements as documented in Reference 4. !

.i F

LC0 The structural failure of any of the jet pumps could cause 1 significant degradation in the ability of the jet pumps to allow i reflooding to two-thirds core height during a LOCA. OPERABILITY :

of all jet pumps is required to ensure operation of the I recirculation system will be consistent with the assumptions 1 L used in the-licensing basis analysis (Ref.1). !

L L -;

- APPLICABILITY The jet pumps are required to be OPERABLE in MODES 1 and 2 i K which is consistent with the requirements for operation of the- l recirculation system (Reference LCO 3.4.1). During MODES 3, 4 .;

and 5 the recirculation system is not required to be in ~

operation and insufficient flow is available to evaluate jet pump operability.

s 3

[

1+

ACTIONS A.1 and A.2 i

An inoperable jet pump can increase the blowdown area and- i reduce the capability of reflooding the core during a design i basis LOCA. Therefore with one or more jet pumps inoperable, .

the reactor is required to be shutdown. The Completion Times allow for a controlled and orderly Shutdown of the reactor.

(continued) :

l

. Grand Gulf - Unit 1 B 3.4-9 DRAFT B 2/07/90

w

.. ; Jet Pumps B 3.4.3 I

l BASES (continuedl m

, SURVEILLANCE SR 3.4.3.1 L REQUIREMENTS ~ I M

This surveillance requirement is designed to detect significant- ,

degradation in jet pump performance that precedes jet pump 1

. failure-(Ref. 2-and 3). The jet pump failure of concern is a complete mixer displacement due to jet pump beam failure. - Jet i

pump plugging is also of concern since it adds flow resistance !

to the recirculation loop thereby affecting coastdown flow to the core. Significant degradation is Indicated if more than 1 one of three specified criteria confirms unacceptable -

O

-deviations from established patterns or relationshi)s.. The allowable deviations from the established patterns save been developed based-on the variations experienced at plants during normal. operation and with jet pump assembly failures (Ref. 2 and3).

L The~ recirculation flow control valve operating characteristics 4 (loo ) flow versus flow control valve position) are determined.

by t1e flow resistance from the loop suction through_ the jet i pump-nozzles. A change in the relationship indicates a flow '

restriction, loss in pump hydraulic performance, leak, or new-flow path between the recirculation pump _ discharge and jet pump.

nozzle.

Total core flow can be determined from measurements of the recirculation loop drive flows. Once this relationship has ,

been established, increased or reduced total core flow for the same recirculation loop drive flow may be an indication of-

' failures in one or several jet pumps., o Individual jet pumps in a. recirculation. loop typically do not have the same flow. The unequal flow is due to the drive flow manifold which does not distribute flow equally to all- risers,- 4 individual jet pump; manufacturing and installation tolert.nces which cause different jet pump efficiencies, and the resistance !

the jet pump flow oncounters in the lower plenum and vessel -

annulus. The flow (or jet-pump diffuser-to-lower plenum differential pressure) pattern or relationship of one jet pump- J to the loop average is repeatable. An appreciable' change in- !

this relationship is an indication that increased-(or reduced) resistance has occurred.in one of the jet pumps. This may be i seen as' an increase in the relative flow for a jet pump that has >

experienced beam cracks.

This surveillance requirement is not required to be performed when THERMAL POWER is < 25% of RATED THERMAL POWER because jet pump nohe precludes tee collection of repentable and meaningful data during low flow conditions approaching the threshold response of the associated flow instrumentation.

(continued)

Grand Gulf - Unit 1 B 3.4-10 DRAFT B 2/07/90 ,

" 9 I '

' Jet' Pumps B 3.4.3-BASES (continued) i L' -

SURVEILLANCE SR 3.4.3.1 (continued)

REQUIREMENTS (continued) Also, this surveillance is not applicable for the jet pumps in a loop not operating becuase there is no drive flow in that ;

loop. Operating experience has demonstrated that.a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ^

~ frequency for this surveillance is adequate when > 25% of RTP.-

REFERENCES 1. Grand Gulf FSAR, Section 6.3. l 1

2, GE-Service Information Lettpr No. 330, " Jet Pump Beam Cracks", June 9 1980- ]

1

3. 1NUREG/CR3052, "Closecut of IE Bulletin 80-07: BWR Jet _ Pump

Assembly Failure", November 1984. j I

4. NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment", November-1987.

d j

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1 3

Grand Gulf - Unit 1 B 3.4-11 DRAFT B 2/07/90

. Grand Gulf Nuclear Station-

. Technical--Specification Improvement Program i Revision Summary Sheet Proposed LCO/Section:' 3.4.4 Rev.._L 1 Rya ,

its Chance Descriotion Cateaorv 1

l' LCO 3.4.4 is reformatted from LIMITING CONDITION 1 l t> FOR OPERATION 3.4.2.1 and 3.4.2.2 except as noted '

below. l 1

2 CONDITION A is reformatted from LCO 3.4.2.1 ACTION a. 1 3 SR 3.4.4.1 is developed from LCO 3.4.2.1. '3B+

4 SR 3.4.4.2 is added to check manual operation of 3A+

the SRVs.

. 5 CROSS REFERENCE to related LCO is added. 1 6 ACTION b is deleted. Suppression pool temperature 3B requirements are spectfied in'LCO 3.6.2.1.

7 The_SRV tail-pipe pressure switches (which provide 4 monitor / alarm functions only) in LCO 3.4.2.1 item b, ACTION c- and SR 4.4.2.1.1 are deleted (or considered in the OPERABILITY requirements for LC0 3.4.4).

8 DELETED

9. Footnote '*' to page 3/4 4-5-and 3/4 4-7_is relocated 2 to BASES for SR 3.4.4.1. ,

10 Footnote '#' to page 3/4-4-5 is deleted. SRVlow-low 3B set is handled by LCO.-

11 DELETED 7

.12 Footnote '*' to page 3/4 4-6.is deleted because 1 the NOTE to SR 3.4.4.2 and LCO 3.0.4 perform the intent of the footnote provision.

13 CONDITIONS B and C are reformatted from ACTIONS a and b. 1 of LC0 3.4.2.2. !

14 CONDITION O is reformatted from LCO 3.4.2.1 ACTION d 1 and LC0~3.4.2.2 ACTION c.

~15 CONDITION E is developed frorr LCO 3.4.2.1 ACTION d 3B i

and LCO 3.4.2.2 ACTION c.

-16 SR 3.4.4.3 is added to perform a' CHANNEL CHECK. 3B+ .

1' L 17 SR 3.4.4.4 is developed from SR 4.4.2.1.2.a and 3B L SR 4.4.2.2.1.a. The frequency is reduced to 92 i days from 31 days, m

?

3

Grand- Gul f.. Nuclear Station 1 Technical Specification Improvement Program:

-Revision Summary Sheet i

Proposed LCO/Section: 3.4.4 /Rev. _l_ SEys 1133 Chance Descriotion Cateaorv 18 SR'3.4.4.5-is developed from SR 4.4.2.1.2.a and 38 SR.4.4.2.2.1.a._ The frequency is reduced to 92 4 days from 31~ days..

19 SR 3.4.4.6 is reformatted from SR 4.4.2.1.2.b'and 1 SR 4.4.2.'2.1.b. t

^

20 SR 3.4.4'7 is developed from SR 4.4.2.1.2.b and

. 3B SR 4.4.2.2.1.b. A NOTE is added excluding valve actuation.from the test.

21 Listing of valves in CTS LC0 3.4.2.2 is relocated 2 to BASES. <

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S/RVs. ;

z3.4.4:

13.4 . REACTOR ~ COOLANT SYSTEM.

3.4.4' Safety / Relief Valves LCO 3.4.4 For the.following Safety / Relief Valves (S/RVs):

.The safety function of 1 7.S/RVs shalllbe OPERABLE, ;

AND The relief function of 16 additional S/RVs shall be OPERABLE', _;

AND The Low-Low Set (LLS) function of 6 S/RVs shall be OPERABLE. j Valve Number of Setpoint

-Function S/RVs 'fosio) '

A. Safety 8 1165 'i 11.6 ps; .

6 1180' 11.8 $s; .

6 11901 11.9 pt--

Relief 1103. 115 psi B. 1 10 1113 -15 psi -

9 1123. 15 psi C. Low-Low Set l' Open 1033 115 psi Close 926 -15 psi.

^

l Open 1073 15 psi Close.936 15 psi ,

4 Open-1113 i 15. psi ,

close 946 15 psi' APPLICABILITY: MODES 1, 2, and 3.

i I.

l Grand Gulf - Unit 1 3.4-5 DRAFT B 4/10/90

S/RVs- .

3.4.4-ACTIONS CONDITIONS REQUIRED ACTIONS- COMPLETION TIME

.A. One or.more of the A.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months required S/RVs for the safety or- relief MQ function inoperable A.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months B. One'LLS S/RV B.1 Restore inoperable 14 days from inoperable, valve'to OPERABLE discovery of

status, inoperable valve.

More than one LLS S/RV C .~1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

inoperable.

80 E

C.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Required Action and associated Completion Time of Condition B not met.

D. With either Division D.1 Restore inoperable

'l or 2 -of relief or instrumentation 7 days low-low set actuation to OPERABLE instrumentation status. :

inoperable.

r E. Required' Action and E.1 Be-in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion ,

E Time of Condition D MQ '

not met.

L. .

E.2 Be in MODE'4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months o

g With both Divisions of relief or low-low set

. actuation instrumentation inoperable. ,

( _ _ _ _

l' Grand Gulf - Unit 1 3.4-Sa DRAFT B 4/10/90

S/Ris; 3.4.4

SURVEILLANCE REOUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.4.1 Demonstrate the safety function. lift According to ;

setpoints of the required S/RVs. SR 3.0.5 M

18 months SR 3.4.4.2 Demonstrate each-required S/RV opens when -----NOTE-----

manually actuated. Only required within 12-hours when reactor steam dome pressure is adequate to perform

!he !es}

18 months SR 3.4.4.3- Perform a CHANNEL CHECK. 12-hours a SR 3.4.4.4 Calibrate the trip unit. 92 days .[

3 SR 3.4.4.5 Perform a CHANNEL FUNCTION TEST._ .92 days SR 3.4.4.6 Perform a CHANNEL CALIBRATION. 18 months I I

SR .3.4.4.7 --------------Note--------------

Valve actuation may be excluded.

Perform a LOGIC SYSTEM FUNCTION-TEST. 18 months l

l Grand Gulf - Unit 1 3.4-6 DRAFT B 4/10/90 J

1

.- S/RV s 3.4.4 r CROSS REFERENCES ~ ,

'A TITLE- NUMBER 1

f ECCS - Operating ~ 3.5.1 s

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l Grand Gulf - Unit 1 3.4-6a DRAFT B 4/10/90

LS/RVss q B 3.4.4i

.B 3.4- REACTOR COOLANT! SYSTEM I B 3.4.4. Safetv/ Relief Valves BASES- I j

BACKGROUND' The ASME Boiler and Pressure Vessel Code requires the reactor._

pressure vessel be protected..from overpressure during upset conditions. ' As part of the nuclear pressure relief system, the size and number of- safety / relief valves -(S/RVs) are

- selected such that- peak pressure in the nuclear system will :

not exceed the ASME Code limits for the reactor coolant i pressure boundary.

5 The; S/RVs are located on the main steam lines between the -

reactor vessel and the first isolation. valve within. the drywell. Each S/RV discharges steam through a discharge line ;

to a point below the minimum water level in the suppression- ,

pool. , - The: S/RVs can actuate by either of two modes - the j

. safety mode'or the relief. mode.

In the; safety mode-(or spring mode of operation), the direct:

action of the steam pressure in the main steam lines will act '

against a spring-loaded disk that will pop open when the valve inlet _ pressure _ exceeds the-spring force. ;

For the relief mode of operation, each SRV has two' pressure. ,

actuation trip systems, Division 1 and . Division 2. The

o. Division-11 trip system consists of.an "A"'and'an "E" channel 4 of pressure instrumentation.. The Division 27 trip system consists of' a: "B" and' an "F" _ channel of 5 pressure' instrumentation. : Each channel consists of a separate trip unit' and associated logic such that when RPV pressure (as. sensed '

by a pressure transmitter) reaches the set point of the trip unit, the associated relay logic- energizes and' provides a .

permissive to energize the _ respective divisional solenoid for .

the appropriate SRV; If the other channel in that trip system.

is then placed in, the trip state the solenoid operatedLair-valve will open (solenoid energized) and allow air to port to !

the pneumatic operator _ for the SRV, and the SRV will open. J Eight _of the S/RVs that provide:the relief function.are part l

of the . Automatic. Depressurization System-(ADS) specified in LCO 3.5.1.

Six of the S/RVs provide the Low Low Set (LLS) relief function.

To ensure that no more than one relief valve reopens following a reactor isolation event, two valves are provided with lower opening and closing set points and four valves are provided *

.D- with lower closing setpoints only. These setpoints override the normal relief mode setpoints following the initial opening of _any of the relief valves and act to hold open these valves L . !

(continued)

Grand Gulf - Unit 1 B 3.4-12 DRAFT B 4/10/90 h

l; l '.

i S /RVs :

, B 3.4.4

' BASES'(continued 5 BACKGROUND longer, thus preventing subsequent reopening of more than one

.(continued) valve. The low-low set mode of operation is activated anytime the relief mode of cperation is activated.- Relay logic (two logic channels per division) seals in upon initial _ relief mode activation, providing a permissive to energize the associated .

division's low-low set SRV - solenoids. A subsequent trip ,'

(vessel pressure at or above) of the low-low set. SRV's associated pressure instrumentation will then complete the ;

logic, energize the SRV's solenoid and open-the SRV. The low-low set instrumentation consists of a single pressure channel per trip system for the two valves with the lowest set point '

while the four valves with the highest low low set setting have two channels of instrumentation per trip system.

i APPLICABLE The overpressure protection system must accommodate the eost-L SAFETY severe pressurization transient in order to prevent the reactor- 4 ANALYSES coolant system pressure from reaching the transient pressure safety limit. Evaluations have determined that the most severe transient .is the closure of all-main steam line isolation valves (MSIVs) followed by reactor scram on high neutron flux .

(i.e.. failure of the direct scram associated with MSIV= l position) (Ref. 1). For the purposes of the analyses, six of ,

the S/RVs are assumed to operate in the relief mode, and seven "

y of the S/RVs in the safety mode. The analysis results indicate- a the design S/RV capacity is- capable of maintaining reactor pressure below the ASME Code limit of 110% of vessel design.

pressure (110% x 1250 psig = 1375 psig). Reference 2 discusses additional events- which are expected to - actuate the S/RVs.

l From an overpressure standpoint, these events are bounded by-J the MSIV closure with flux scram event described above.

The LLS relief mode is designed to protect the containment from.

L excessive loads by ensuring that no more than one relief valve- ,

reopera subsequent to the first full blowdown on-an isolation event. The LLS function' minimizes the induced loading on the

lE containment / suppression pool boundary as a result of subsequent S/RV discharge following the first : full blowdown on an isolation event. To remain consistent with the containment' '{

loads analyses, the LLS mode of the S/RVs must ~ function to ensur.e subsequent actuation of only- one SRV.

The relief ins.trument.ation must' function for the S/RVs to l; function to prevent over-pressurization of the nuclear-system j L' and satisfy 'the requirements of the transient analyses -

p presented in Reference 2. l L

S/RVs satisfy the requirements of Selection Criterion 3'of the. I NRC Interim Policy Statement on' Technical Specification !

Improvements as documented in Reference 3.

(continued)

Grand Gulf - Unit 1 B 3.4-13 DRAFT B 4/10/90 1

I Y wm--

S/RVs' B 3.4.4:

W BASES fcontinued)

LCO Seven S/RVs are required to be OPERABLE in. the: safety mode,

? , and an additional 6 S/RVs (other than the 7 S/RVs that satisfy must be OPERABLE in the relief mode. In ,

the safety 1, Reference function)luation an eva was performed to establish the -

. parametric relationship between the peak vessel pressure and L ' the number of OPERABLE-S/RVs. The results show that with a minimum of 7 S/RVs in the safety mode and 6 S/RVs in the relief.

mode OPERABLE, the ASME - Code limit of 1375 - psig is- not' exceeded.

~

The S/RV setpoints are established to ensure the ASME Code '

limit on peak reactor pressure is satisfied. The ASME code'.

, specifications require the lowest. safety valve be set.at or below vessel design pressure (1250 psig) and the highest safety valve be set so the total accumulated pressure does not exceed 110% of the design pressure for overpressurization conditions.

The transient evaluations in Reference 2 are' based on these

setpoints, but also include additional uncertainties to account 1 for potential-setpoint drift.

'TheseVsix S/RVs with the LLS function.must be OPERABLE in the. l.~

relief mode to satisfy the assumptions of the safety analysis >#

(Ref. 1). The six valves that satisfy the LCO, along with their setpoints are listed below. The requirements of. the LCO are- applicable to the mechanical and electrical / pneumatic capability of each valve, and its associated instrumentation, to perform its LLS function.

Setooint Valve No. Qgg Close j m

.F051D 1033 -926 F051B 1073 936 F0470 1113 946 F047G 1113 946 ,

F051A 1113 946 F051F. 1113 946

~

[

Operation with less valves OPERABLE than-specified, or with setpoints greater than specified, could result in a more severe -i reactor response to a transient than predicted, possibly resulting in the ASME Code limit on reactor pressure being ,

exceeded. ,

(continued) i Grand Gulf - Unit 1 B 3.4-14 DRAFT B 4/10/90 x

. -. = .. - .. ..- - - - . - _ - - . -

S/RVs !

B 3.4.4 i BASES'(continued)

APPLICABILITY-The specified number of S/RVs and required . S/RV and LLS _;

pressure actuation instrumentation must be OPERABLE in MODES -l 1,._2 and 3 since there is considerable. energy in the, reactor:

core and the limiting design basis transients _are assumed to- -3 occur. The S/RVs may.be required to provide pressure relief to discharge energy from!the core until . such time that the Residual Heat Removal (RHR)- system is capable of dissipating the heat being generated. - In MODE 4,. decay heat _ levels are low enough such that the RHR system is adequate, and reactor- t pressure levels are low enough such that the overpressure-limit cannot' be challenged by assumed operational transients 'or-accidents.- In MODE 5, the reactor vessel head is unbolted or removed and there is no reactor - coolant pressure boundary - 4 (RCPB). The S/RV ana LLS functions are not needed during these.

conditions.

ACTIONS A.1. A.2 With any of the required S/RVs inoperable, the reactor must be in MODE 3 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in MODE 4 in 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . With less than the minimum number of S/RVs. OPERABLE, a transient may -

result in the violation of the. ASME Code -limit on reactor pressure. It is therefore necessary for the plant to be in -

- a condition where the S/RVs are not rrawwd. The Completion -

Times allow for a controlled and atGA !r Shutdown ofJ thei reactor. :;

L S .'l u C .JW l'

With one of the'6 LLS S/RVs inoperable, the remaining'0PERABLE -i LLS S/RVs are . adequate to perform the designed function, '

o However, the overall reliability is reduced. A limited out of service time (14 days) is.therefore. allowed to restore the~

valve to OPERABLE status. . .

With more than one LLS S/RV inoperable or' Required Action B.1; and associated Completion Time not met, .a single failure can cause loss of designed function. The reactor is required to r be in MODE 3 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in MODE.4 in 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months such that L. the reactor is in a condition where the LLS S/RVs are not i required.

D_d L

With one division of relief or LLS actuation instrumentation inoperable, the capability still exists through the remaining '

division for the required number of S/RVs to function in the (continued) h Grand Gulf - Unit 1 8 3.4-15 DRAFT B 4/10/90 L

i S/RVs B 3.4.4 BASES (continued)

ACTIONS relief and/or LLS mode. An allowable out-of-service time is I

-(continued) permitted to allow restoration of the inoperable Trip System, I based on the reliability and redundancy of. the remaining .

system. )

E.1. E.2

. With both divisions of relief or LLS actuation instrumentation-inoperable or one division inoperable for more than 7 days,-

, the plant must be placed in a condition where the relief and-LLS functions are not required. -l w

SURVEILLANCE SR 3.4.4.1 l REQUIREMENTS j This surveillance requirement demonstrates that the S/RVs will open at the pressures assumed in the safety analysis of Reference 1. The demonstration of the S/RV lift settings must be performed during shutdown and in accordance with the provisions of SR 3.0.5. The lift setting pressure shall correspond to ambient conditions of the valves; at nominal operating temperatures and pressures. '

SR 3.4.4.2 A manual actitation of each relief S/RV, LLS S/RV and safety S/RV is performed to verify the valve is mechanically i i functioning properly, the solenoids '(for the. relief and LLS L S/RVs) are functioning properly and no blockage exists in the valve discharge line. Adequate reactor steam dome pressure

- must be available to perform this test to avoid damaging the valve. Sufficient time is therefore allowed after the required '

pressure is achieved to perform this test. Adequate pressure .

at which this test is to be performed is specified in plant >

procedures. Plant startup is allowed prior to performing this test because valve OPERABILITY and- the setpoints for overpressure protection are verified, per ASME requirements, ,

prior to valve installation. If the valve fails-to actuate due only to the failure of the solenoid but is capable of openi.ng on overpressure, the S/RV can _ still be considered l

OPERABLE for the safety function.

L-SR 3.4.4.3 The performance of a CHANNEL CHECK is the comparison of-the l- indicated parameter values for each of the required OPERABLE i

channels for a Function. It is based on the assumption that all channel indications should be displaying approximately the l (continued) 1 l'

Grand Gulf - Unit 1 B 3.4-15a DRAFT B 4/10/90

S/RVs B'3.4.4- i BASES (continued) i

' SURVEILLANCE SR 3.4.4.3 (continued)

REQUIREMENTS .

(contjnued) same value consistent with expected values for current' plant conditions.- Consistency is determined by the' plant staff and ,

may be based on a combination of - the ~ channel; instrument uncertainties, indication capability- and- readability.

Comparison with other independent instrument channels measuring.

the same parameter may also be used _ for the CHANNEL CHECK.

If a channel- is outside of the-- criteria, it may be - an indication the instrument has drifted outside of its limit or is not functioning.

.SR 3.4.4.4 o

Calibration of . trip units L provides a check of the trip setpoints. If during trip unit calibration the associated trip u setting is discovered to be less conservative than the-Allowable.Value, the channel must'be declared-inoperable.. If the trip setting is discovered to be less conservativeithan

-accounted for in the appropriate setpoint methodology but is not beyond the Allowable - Value, the channel ~ performance is _,

still-within the requirements.of the plant safety analysis. M Under these conditions, the setpoint must be readjusted to be equal- to or more conservative than- accounted for -in the-appropriate setpoint. methodology.

SR 3.4.4.5 Performance' of a CHANNEL FUNCTIONAL TEST demonstrates the-associated channel will function properly when a signal is-injected into the logic indicative of a required trip. If- J' during the CHANNEL FUNCTIONAL TEST, the associated trip setting is discovered to be less conservative than the' Allowable Value- "

o specified, the channel must be declared inoperable.

i SR 3.4'.4.6 ,j Performance of a CHANNEL CALIBRATION provides a complete check-of the channel including the sensor.and trip unit. If during the CHANNEL . CALIBRATION, the associated trip setting 'is discovered to be less conservative than .the Allowable Value, .

the channel'must be declared inoperable.

,SR 3.4.4.7 Performance of a LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required actuation logic for a specific !

channel. The LOGIC SYSTEM' FUNCTIONAL TEST tests all logic' ,

. components, i.e., all relays and contacts, all trip units, 1

(continued) 1 1

Grand Gulf - Unit 1- B 3.4-15b DRAFT B 4/10/90 l

a

~ . . . 6 _ . _ _ - - - _ _ _ _ . ___.__._____.___________________________z

S/RVs -!

B 3.4.4 l BASES (continued)

SURVEILLANCE. SR 3.4'.4.7 '(continued)

REQUIREMENTS (continued). solid' state logic elements, etc., of a logic circuit, from sensor up to the actuated device. The system functional test- '

, of LCOs 3.4.4 and 3.5.1 for S/RVs overlaps the LOGIC SYSTEM FUNCTIONAL TEST to provide complete testing of the associated safety function. t Surveillance Freauencies In general, surveillance frequencies are based on industry-accepted practice and engineering judgement considering the -

unit conditions required to perform the test,. the ease of -

. performing the test and a- likelihood of. a change in the -

system / component status.

'~ -

RiFERENCES 1. Grand Gulf FSAR, Section 5.2.2.

2. Current Cycle Safety Analysis. *

3. NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment", November 1987.

q i

. A i

li l

{

Grand Gulf - Unit 1 B 3.4-15c DRAFT B 4/10/90 i

p i

r. Grand Gulf Nuclear Station 4 Technical'Spacification Improvement Program-Revision Summary Sheet 1

. Proposed LCO/Section: 3.4.5 Rev. _.L. Ooerational Leakace itam Chance Descrjotion Catecorv

'l LCO 3.4.5 is reformatted from LIMITING CONDITION 1 t FOR OPERATION 3.4.3.2. >

2 DELETED (Ref. CRS 174)_.

3 LC0 3.4.3.2 item d is moved to LCO 3.6.1.6. .1 4 4 A NOTE is added stating that CONDITIONS A through C 1 can be concurrently ec.' e ad.

5- ' CONDITION A is reformatted from ACTION b except 1

-the shutdown requirements are in CONDITION C (seeItem7).

6 CONDITION B is reformatted from ACTION e except -1 the shutdown requirements are in CONDITION C (see Item 7). ,

7 CONDITION C is reformatted from ACTION a and the 1 shutdown' requirements from ACTIONS b and e.

8 SR 3.4.5.1 provides monitoring leakage on a 12 3B a hour frequency.- SR 4.4.3.2.1 provided frequencies '

ranging from-4 hours to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

9 ACTION c is moved to LCO 3.6.1.6. 1 10 ACTION d is deleted. JJustification-in comparison 4 document addressed only-the monitors and not the '

interlocks. ;

11 The' specific leakage detection methods are 3B deleted from SR 4.4.3.2.1. . Leak Detection System requirements are provided'by LCO 3.3.4.1 and the surveillance frequency specified by SR 3.4.5.1.

12- SR 4.4.3.2.2 is moved to LCO 3.6.1.6. 1 '

i.

13 SR~4.4.3.2.3 is deleted. Justification in comparison 4 document addressed only the monitors and not the interlocks (see Item 10).

14 Table'3.4.3.2-2 is deleted. This change is considered 1 administrative because it addresses an action statement which has been deleted (see Item 10).

l' l

L

Grand Gulf Nuclear Station s Technical Specification Improvement Program i Revision Summary Sheet Proposed LC0/Section: 3.4.5 Rev. _1_ Ooerational Lrakage i 1133 Chance Descrintion Cateaorv 15 LCO item D is made applicable in MODE 1;only. 3B' 16 Table 3.4.3.2-1 is removed from the Tech Specs 2 and relocated based upon guidance in the proposed generic letter.

17- Table 3.4.3.2-3 is deleted from Tech Specs and not 4 relocated (see Item 10).

?

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+

Operational Leakage !

g 3.4.5- !

3.4- REACTOR COOLANT SYSTEM.-

3;4.5 Doerational-Leakane

I LCO 3.4.5 Reactor coolant system LEAKAGE shall be limited to:

A. No Pressure Boundary Leakage, E

B. 15 gpm total Unidentified Leakage, s-,

M

-C. 5 30 gpm' Total Leakage averaged over any 24 hour-period, ,

M- .,

D. < 2_ gpm-increase in Unidentified Leakage witnin any r 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months period in MODE 1. ,

APPLICABILITY: MODES 1, 2, and 3.

.........................N0TE-------------------------

Conditions A through C may be concurrently applicable.

.t ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME-A. Unidentified Leakage A.1 Reduce leakage to within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months

> 5 gpm. the limits..

M Total ' Leakage

> 30.gpm.

-M Unidentified Leakage

> 5 gpm and Total Leakage > 30 gpm.

(continued)

' Grand Gulf - Unit 1 3.4-7 DRAFT B 11/22/89

.i

~

Operational Leakage- ;

3.4.5

? ACTIONS'icontinued)~

CONDITION. REQUIRED' ACTION COMPLETION TIllE B. Unidentifiid Leakage B.1 Verify source of leakage 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months increase > 2 gpm in increase is not service t any 4: hour period, sensitive Type 304 or 316 austenitic stainless steel.

t C. Required Actions and C.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Times of Condition A AtiQ or B not met.

C.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months ,

08 Any Pres::ure Boundary Leakage.

SURVEILLANCE REOUIREMENTS SURVEILLANCE FREQUENCY-SR 3.4.5.1 Verify the reactor coolant system LEAKAGE 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,

is less than or equal to the required limits.

i' CROSS-

REFERENCES:

None l

=

l 1

l l l I l g Grand Gulf - Unit 1 3.4-8 DRAFT B 11/22/89 1

I b

t Operational Leakage !

B 3.4.5 !

l B 3.4 REACTOR COOLANT SYSTEM l B 3.4.5 Onorational Leakane )

BASES j i

r BACKGROUND The reactor coolant system includes systems and components that !

contain or transport fluids to or from the reactor core. These ;

systems form a major portion of the reactor coolant pressure boundary. The pressure containing components of the reactor ,

coolant system, including the portions of the system out to and :

including isolation valves, are defined as the Reactor Coolant ?

Pressure Boundary (RCPB). Limits on leakage from the RCPB are ;

required to ensure appropriate action can be taken before the l integrity of the nuclear system process barrier is impaired.

The safety significance of leaks from the RCPB can vary widely ,

depending on the source of the leak as well as the leakage rate !

and duration. Therefore, detection of leakage in the drywell is necessary. Identified Leakage is defined as the leakage :

into closed systems, such as pump seal or valve packing leaks ,

that are captured, flow metered and pi >ed to a sump or collecting tank. Also, leakage into tie drywell atmosphere from sources that are specifically located and known not to I interfere with the operation of Unidentified Leakage detection or not to be a flaw in the RCPB are considered Identified Leakage. Unidentified Leakage is collected in the drywell i floor. drain sump. Methods for se)arating the Indentified ,

Leakage from the Unidentified Leacage are necessary to provide prompt and quantitative information to the operators to permit '

them to take corrective action.

A limited amount of leakage is expected from auxiliary systems within the drywell that cannot be made 100% leaktight. If l 1eakage occurs from these paths, it should be detectable and isolated from the drywell atmosphere if possible, so as not to ,

l mask any potentially serious leak should it occur.

(continued)

L i

l L Grand Gulf - Unit 1 B 3.4-16 DRAFT B 11/22/89

Operaticnal Leakage i B 3.4.5 i

)

BASES (continued)

APPLICABLE 'The allowable leakage rates from the reactor coolant system .

SAFETY have been based on the predicted and experimentally observed l ANALYSES behavior of pipe cracks. The normally expected background '

leakage due to equipment design and the detection capability of the instrumentation for determining system leakage were also ,

considered. The evidence obtained from experiments suggests

for leakage somewhat greater than s>ecified for Unidentified i Leakage, the probability is small tie imperfection or crack j associated with such leakage would grow rapidly. The '

Unidentified Leakage rate limit is established at 5 gpm to allow time for corrective action before the process barrier could be significantly compromised. This limit is a small fraction of the calculated flow from a critical crack in the -

primary system piping. Based on crack behavior from experimental programs (Ref. 2 and 3) it is estimated that leak l rates of hundreds of gpm will precede crack instability (Ref.

4).

There are no applicable safety analyses that assume the Total Leakage limit. The Tot 71 Leakage limit is specified based on- ;

consideration of inventory makeup capability and sump capacities.

Operational Leakage satisfies the requirements of Selection Criterion 1 of the NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 5. ,

t LCO No Pressure Boundary Leakage is allowed since the potential

exists for a break in the RCPB and a loss of substantial :

inventory.

The Total Leakage rate consists of.all leakage, identified and ,

unidentified, that flows to the drywell floor drain and equipment drain sumps. The Unidentified Leakage rate limit is ,

established to allow time for corrective action before the reactor coolant pressure boundary is significantly compromised and based on a crack size large enough to propagate rapidly.

t (continued) l' Grand Gulf - Unit 1 B 3.4-17 DRAFT B 11/22/89 L

1

=

Operational Leakage B 3.4.5 BASES (continued)

, LCO In addition to the absolute limits discussed above, a limit is (continued)- placed on the increase in Unidentified Leakage over a specified period. An increase of 2 gpm in any 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months period in Unidentified Leakage is indicative of a potential flaw in the RCPB and it should be promptly evaluated to determine the source and extent of the increased leakage. The 2 gpm increase is measured relative to the steady state Leakage value. This allows :

for temporary changes in leakage that are the expected result of transientconditions(e.g.,startup). As such, the 2 gpm increase limit is only applicable in MODE 1 where operating -

pressures and temperatures have been established. ;

APPLICABILITY The potential for RCP8 leakage is greatest when the reactor is pressurized. Under these' conditions, high stresses are applied to the system piping resulting in the potential for crack growth and possible failure of the RCPB. Therefore, detection and measurement of RCPB leakage is required during )

MODES 1, 2 and 3. In MODES 4 and 5, operational leakage limits are not required, since the reactor is not pressurized and the potential for leakage and possible pressure boundary failure is reduced. ,

ACTIONS &J With either the Total Leakage Unidentified Leakage, or both greater than the required limits, actions should be taken to identify the source of the leak and determine the_ significance.

Because the leakage limits are conservatively below the leakage that would constitute a critical crack size, a limited time is allowed to evaluate the situation. If a change in Unidentified '

Leakage has been adequately identified and quantified, it may be reclassified and considered as Identified Leakage. However, the Total Leakage limit would remain unchanged.

(continued)

I l

l 4

Grand Gulf - Unit 1 B 3.4-18 DRAFT B 11/22/89 l

\ i

~

i_ _Operatlonal Leakage l B 3.4.5 L

BASES (continued)

ACTIONS JL1 (continued)

An increase of 2 gpm in Unidentified Leakage in a 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months period is an indication of a potential flaw in the RCPB and should be promptly evaluated. Although the increase does not necessarily violate the absolute Unidentified Lenkage limits, certain susceptible components should be determined to not be the source of the leaks. Reactor coolant system service sensitive Type 304 and 316 austenitic stainless steel piping should be evaluated and eliminated as a potential source of the increased leakage. These components are especially susceptible to intergranular stress corrosion cracking.

C.1. C.2 If the LEAKAGE cannot be restored to within the required limits, the reactor should be in MODE 3 and subsequently in MODE 4. If Pressure Boundary leakage occurs there is the potential that the flaw in the RCPB could eventually result in a pipe break or other LOCA. Since the area being monitored is

' inaccessible, the reactor must be in MODE 3 and subsequently in MODE 4 to allow a visual inspection to determine the source of the leak, Comoletion Times All Completion Times are based on industry accepted practice and engineering judgement considering the number of available systems and the time required to reasonably complete the Required Action.

SURVEILLANCE SR 3.4.5.1 REQUIREMENTS The reactor coolant system LEAKAGE is monitored by a variety of systems designed to provide alarms when leakage is indicated and to quantify the various types of leakage. Leakage detection is discussed in more detail in the Bases for LCO 3.3.4.1. Changes in sump levels and measured flow rates are monitored to determine actual leakage rates. However, additional methods may be used which quantify leakage within the quidelines of Reference 1. Operating experience has demonstrated that a 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months frequency for this surveillance is adequate.

(continued)

Grand Gulf - Unit 1 B 3.4 19 DRAFT B 11/22/89

Operational Leakage B 3.4.5 BASES (continued)

REFERENCES 1. Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems", May 1973.

2. GEAP 5620, " Failure Behavior in ASTM A106B Pipes Containing Axial Through-Wall Flaws," April 1968.

3. NUREG-76/067, " Investigation and Evaluation of Cracking in Austenitic Stainless Steel Piping of Boiling Water Reactor Plants,' October 1975.

4. Grand Gulf FSAR, Section 5.2.5.5.3.

5. NE00 31466, ' Technical Specification Screening Criteria Application and Risk Assessment *, November 1987.

Grand Gulf - Unit 1 B 3.4 20 DRAFT B 11/22/89

Grand Gulf Nuclear Station Technical Specification ICprovement Program Revision Summary Sheet Proposed LC0/Section: 3.4.6 Rev. 1. Soecific Activity lig Chance Descriotion Cateaorv 1 LCO 3.4.6 is reformatted from LIMITING CONDITION 1 FOR OPERATION 3.4.5.

2 The 100/E microcuries per gram criterion of LCO 3B >

3.4.5 is deleted because iodine monitoring is considered most limiting. '

3 The applicability in MODE 4 is deleted because no 3B pressure or steam exists to provide a force or medium to transport activity beyond the vessel.

8 The applicability in MODES 2 and 3 is limited to 3B !

when any main steam line is not isolated because activity cannot escape a breach outside containment with the MSIVs closed.

5 CONDITION A is reformatted from ACTIONS a.1 and b 1 -

excep*. as modified per Items 6 and 7.

6 REQUIRED ACTIONS A.1 and B.1 are reformatted from 1 ACTION b and Table 4.4.5-1-item 4(a).

7 REQUIRED ACTIONS A.2 and B.2 are reformatted from 3B-ACTION a.1 except the requirement for HOT SHUTDOWN is removed.

8 SR 3.4.6.1 is reformatted from Table 4.4<5-1 1 ;

item 2.

9 ACTION a.2 is deleted (see Item 2). 3B l 10- ACTION b is revised to delete reference to 100/E 3B l- microcuries per gram (see Item 2).

l 11 ACTION c is deleted. 4 ;

i 12 Footnote '*' to page 3/4 4-17 is deleted. It I applied to the startup test program and is no longer applicable.

13 Table 4.4.5-1 items 1 and 5 are relocated. 2 14 Table 4.4.5-1 item 3 is deleted (see Item 2). 38 -l 15 Table 4.4.5-1 item 4(b) is deleted. 4 j 16 Footnote '*' to page 3/4 4-18 is deleted (see Item 2). -38

r .

Grand Gulf Nuclear Station Technical Specification Irprovement Prograa t

Revision Summary Sheet Proposed LCO/Section: 3.4.6 - Rev. _1_ Soecific Activity ;

1113 Chance Descriotion Cateoorv 17 Footnote '#' to page 3/4 4-18 is deleted. This 1 provision is addressed by REQUIRED ACTIONS A.1 and B.1 (as reformatted).

4 1,8 Table 4.4.5-1 is deleted. 4

[

P 4

t B

t A

P P

f y

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)

Specific Ac%ietty 3.4.6 3.4 REACTOR COOLANT SYSTEM

{

3.4.6 $necific Activity -

LC0 3.4.6 The specific activity of the primary coolant shall be

< 0.2 microcuries per gram DOSE EQUIVALENT l 131. '

APPLICABILITY: MODE 1, f MODES 2 and 3 with any main steam line not isolated. ;

ACTIONS CONDITION REQUIRED ACTION COMPLETION T!ME A. Primary coolant A.1 Perform an isotopic Once per specific activity analysis for lodine. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months

> 0.2 but < 4.0 .

sci per gram DOSE 6RQ EQUIVALENT I-131.

A.2 Restore specific 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months

' activity to within limits.

B. Required Actions and B.1 Perform an isotopic Once per associated CMpletion analysis for lodine. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Times of Condition A not met. AN.Q ;

QB B.2 1solate all main steam 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months lines.

Primary coolant specific activity

'> 4.0 uti per gram DOSE EQUIVALENT I-131.

Grand Gulf - Unit 1 3.4-9 DRAFT B 2/13/90

e Specific Activity

, 3.4.6 SURVEILLANCE RE0VIREMENTS SURVEILLANCE FREQUENCY SR 3.4.6.1 .- -....--- ---NOTE---------- - - -

Only required in MODE 1.

Demonstrate specific activity of primary 31 days ;

coolant is < 0.2 sci per gram DOSE EQUIVALENT T.131.

CROSS-

REFERENCES:

None f

)

i i

)

Grand Gulf - Unit 1 3.4-10 DRAFT B 2/13/90

L , -

Specific Activity B 3.4,6 B 3.4 REACTOR COOLANT SYSTEM B 3.4.6 soecific Activity BASES BACKGROUND Ouring circulation, the reactor coolant acquires radioactive material due to release of fission products into the coolant and activation of crud particles in the roactor coolant. These '

radioactive materials in the reactor coolant could contribute to release _of radioactive materials into the environment during i

design basis accidents. :

Limits on the maximum allowable level of radioactivity in the ,

reactor coolant are established to assure, in the event of a release of any radioactive material to the environment during a '

design basis accident, radiation doses are maintained within the limits of 10 CFR 100. ,

APPLICABLE Analytical methods and assumptions involving radioactive i SAFETY material in the primary coolant are presented in FSAR Chapter ANALYSES 15, Accident Analyses. The specific activity in the reactor coolant (source term) is an initial condition assumed for evaluation of the consequences of an accident due to a main

steam line break (MSLB) outside the containment. No fuel .

damage is postulated in the MSLB accident, and the release of radioactive material to the environment is postulated to be terniinated by complete closure of the main steam line isolation ,

valves (MSIVs). This release forms the basis for determining t off site doses (Ref. 1). The limitations on the specific activity of the primary coolant ensure the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thyroid and whole body doses resulting from a main steam line failure ;

outside the containment during steady state operation will not exceed 10% of the dose guidelines of 10 CFR 100. The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. .

These values are conservative in that specific site parameters, such as site boundary location and meteorological conditions, were not considered in this evaluation.

Specific Activity satisfies the requirements of Selection i Criterion 2 of the NRC Interim Policy Statement on Technical l Specification improvements as documented in Reference 2.

L (continued)

Grand Gulf - Unit 1 B 3.4-21 ORAFT B 2/13/90 i

Specific Activity B 3.4.5 ,

i BASES'(continued)

LCO The primary coolant specific activity level of 5 0.2 l microcuries per gram DOSE EQUIVALENT l-131 is required to '

ensure the source term assumed in the safety analysis of the MSLB is not exceeded such that any release of radioactive ;

material to the environment does not exceed 10 CFR 100 limits.

APPLICABILITY Limitations on levels of primary coolant radioactivity are applicable during MODE I and MODES 2 and 3 with any main steam ,

line not isolated since there is an escape path for release of !'

radioactive material from the coolant to the environment in the event of a MSLB outside of the primary containment, buring l MODES 4 and 5, no limits are required since the reactor is not i pressurized and the potential for leskage is reduced. ,

ACTIONS A.I. A.2. B.I. B.2 A primary coolant specific activity level > 0.2 nCi por i gram DOSE EQUIVALENT I-131 indicates the presence of some '

abnormality in plant operations. The range between 0.2 uCi -

and 4.0 aCi is acceptable for up to 48 continuous hours to account for potential iodine spiking that may occur following '

changes in THERMAL POWER. Increased surveillance of the reactor coolant specific activity during this period is required to closely monitor the condition and determine if additional limits are exceeded. ,

If coolant specific activity cannot be restored < 0.2 5C1 within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , or when coolant activity is > 470 ,Ci, the main steam lines are required to be isolated. This action precludes the possibility of release of radioactive material to the environment in excess of the requirements of 10 CFR 100, during a postulated MSLB accident.

Comoletion Times 4 All Completion Times are based on ? .dustry accepted practice and engineering judgement considering the number of available -

systems and the time required to reasonably complete the Required Action.

(continued)

\

Grand Gulf - Unit 1 B 3.4-22 DRAFT B 2/13/90 4e+ -,- w - - - , - - - . - - ,

Specific Activity B 3.4.6 ,

BASES (continued)

SURVEILLANCE SR 3.4.6.1 j REQUIREMENTS Isotopic analysis for DOSE EQUIVALENT I-131 concentration is necessary to determine that the specified maximum primary coolant activity is < 0.2 microcuries per gram DOSE

EQUIVALENT I-131 durTng steady state operation. Operating experience has demonstrated that a '31 day frequency for this -

surveillance is adequate.

Surveillance Frecuencies i 1

In general, surveillance frequencies are based on industry ,

accepted practice and engineering judgement considering the i unit conditions required to )erform the test, the ease of ,

performing the test and a lice 11 hood of a change in the system / component status, t

REFERENCES 1. Grand Gulf, FSAR Section 15.6.4. ,

t

2. NE00-31466, " Technical Specification Screening Criteria Application and Risk Assessment" November 1987.

P i

i Grand Gulf - Unit 1 B 3.4-23 DRAFT B 2/13/90

, t

~

Grand Gulf Nuclear Station Technical Specification Improvement Program Revision Summary Sheet Proposed LCO/Section: 3.4.7 Rev.._L RHR-Shutdown 11gm Chance Descriotion C_ateoorv 1 LCO 3.4.7 is reformatted from LIMITING CONDITIONS 1 FOR OPERATION 3.4.9.1 and 3.4.9.2.

2 The LCO statement is revised to only require that 3B two RHR shutdown cooling subsystems be OPERABLE. The requirement for at least one RHR SDC loop to be in operation or one recirculation pump to be running is deleted.

3 The reference to one RHR pump and heat exchanger 2 in the LCO statement for LCOs 3.4.9.1 and 3.4.9.2 is relocated to the BASES.

4 The applicability in MODE 4 is revised to a 3B condition based upon decay heat generation and heat removal loads.

5 CONDITION A is developed from ACTION a for 3B LCO 3.4.9.1.

6 CONDITION B and C are developed from ACTION a for LCOs 3B 3.4.9.1 and 3.4.9.2.

7 The requirement to demonstrate the alternate 2 method of decay heat removal is deleted. Alternate capability car be provided by plant specific administrative controls.

8 ACTION b for LCOs 3.4.9.1 and 3.4.9.2 is deleted. 2 Coolant circulation is available without forced circulation and can be administrative 1y controlled.

9 SR 3.4.7.1 is added'to verify the capability to 3A+

establish the correct valve alignment for the shutdown cooling mode of RHR.

10 SRs 4.4.9.1 and 4.4.9.2 are deleted. The 3B revision to the LCO statement (see Item 2) removes !

the need to periodically verify a SDC loop is operating, f

l i

g:r i Grand Gulf Nuclear Station l Technical Specification IGprovement Program j Revision Summary Sheet I

Proposed LCO/Section: 3.4.7 Rev. _1_ RHR-Shutdown i 11gm Chance Descriotion Cateaorv 'i 11 Footnote '#' to pages 3/4 4-26 and 3/4 4-27 is 3B l deleted. The revision to the LCO statement (see Item 2) eliminates the need for the other loop to be in operation during this time. ,

4 12 Footnote '*' to pages 3/4 4-26 and 3/4_4-27 is 3B l deleted. The revision to the 1.00 statement (see i Item 2) removes the requi ement for a RHR SDC loop to always be in operation.

13 Footnote '##' to pages 3/4 4-26 and 3/4 4-27 is 3B deleted. The revision to the LCO statement (see Item 2) removes the requirement for a RHR SDC loop to -

always be in operation. 1 14 Footnote '**' to page 3/4 4-26 is deleted. .This 3B provision is addressed in CONDITION C.

15 Footnote '**' to page_3/4 4-27 is deleted. It 1 ,

applied to RF03 only. -

16 Footnote '***' is deleted. It applied to RF03 1 only.

17 CROSS REFERENCES are added. I ;

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. I RHR o Shutdown 7 3.4.7 3.4 REACTOR COOLANT SYSTEM 3.4.7 Residual Heat Removal - Shutdown ,

LCO 3.4.7 Two Residual Heat Removal (RHR) shutdown cooling subsystems shall be OPERABLE.

APPLICABILITY: MODE 3 with- reactor steam dome pressure < the RHR cut-in ;

permissive pressure, !

MODE 4 with heat losses to ambient not sufficient to maintain .

average reactor coolant temperature i 200'F. l ACTIONS CONDITIONS REQUIRE 0 ACTION COMPLETION TIME A. One of the required RHR A.1 Restore the required '

shutdown cooling RHR subsystem to subsystems inoperable. OPERABLE status. -

Q3 A.2 Provide an alternate As soon as method capable of practicable ;

decay heat removal t for the inoperable "

subsystems.

B. No RHR shutdown cooling B.1 Restore at least one As soon as subsystem OPERABLE. RHR subsystems to practicable OPERABLE status.

L C. Required Action and C.1 Provide an alternate As soon as '

associated Completion method capable of practicable Time of Condition B not decay heat removal met, for each inoperable subsystem, i

3 F

Grand Gulf - Unit 1 3.4-11 B 3/21/90

' RHR o Shutdown 3 . 4 ~. 7 i

SURVEILLANCE RE0VIREMENTS SURVEILLANCE FREQUENCY SR 3.4.7.1 Verify for the required RHR shutdown 31 days cooling subsystem (s) each manual, power operated, or automatic valve in.the flow QB path, not locked, sealed or otherwise secured in position,' is in the correct 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when position or is capable of being . reactor steam manually aligned in the correct position. dome pressure is < the RHR cut-in permissive pressure.

CROSS-REFERENCES TITLE NUMBER ECCS - Operating 3.5.1 ECCS - Shutdown 3.5.2 Residual Heat Removal Suppression Pool Cooling 3.6.2.3 i .

l +

l l1.

1 i

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Grand Gulf - Unit 1 3.4-12 DRAFT B 3/21/90 l

l

RHR o Shutdown L B 3.4.7 -

B 3.4 REACTOR COOLANT SYSTEM B 3.4.7 Re'sidual Heat Removal - Shutdown BASES BACKGROUND Irradiated fuel in the reactor pressure vessel (RPV) generates decay heat during normal and abnormal shutdown conditions, potentially resulting in an increase in the temperature of the reactor coolant. The reactor vessel and internals also contain sensible or stored heat energy. This decay and sensible heat

, is required to be removed such that the. reactor coolant

- temperature can be reduced to or maintained at i 200*F. A system capable of removing decay heat is therefore required to perform this function.

The two shutdown cooling loops of the Residual Heat Removal (RHR) system provide decay heat removal. Each loop coasists of a motor driven pump, two heat exchangers in series,'and the associated piping and valves. Both loops have a common suction from the same recirculation loop. Each pump discharges the reactor coolant, after it has been cooled by circulation through the respective heat exchangers, to the reactor. The RHR heat exchangers transfer heat to the Standby Service Water System (LCO 3.7.1 and LCO 3.7.2). The RHR shutdown cooling mode is a manually controlled system.

APPLICABLE Residual heat removal through operation of the shutdown cooling SAFETY mode of the RHR system is not required for mitigation of any.

ANALYSES postulated transients or accidents evaluated in the safety analyses. However, the NRC Interim Policy Statement (Ref.1) requires the. RHR system be retained in the Technical Specifications even though the shutdown cooling mode of RHR did not satisfy any of the selection criteria (Ref. 2)..

(continued) l

. Grand Gulf - Unit 1 B 3.4 24 DRAFT B 3/21/90

--s__ mama.m._m_.___-A-mm.a- ____.--____=.m.._a._ -

_2._____2 - ._ __ __ .m m .-_

< f RHR-Shutdhwn B3.4.7 i BASES (continued) !

t LC0 Two shutdown cooling subsystems are required to be OPERABLE.

An OPERABLE RHR shutdown cooling subsystem consists of one RHR pump, two heat exchangers in series, and the associated piping ;

and valves. Additionally, each shutdown cooling subsystem is ~

considered OPERABLE if it can be manually aligned (remote or :

local) in the shutdown cooling mode for removal of decay heat. '

In MODES 3 and 4, one shutdown cooling subsystem of the RHR can provide the required cooling to maintain the desired >

temperature. Two subsystems are required to be OPERABLE to .

provide redundancy. Operation (either continuous or intermittent) of one subsystem can maintain and reduce the reactor coolaat temperature as required.

APPLICABILITY Decay heat removal at reactor pressures above the RHR cut-in ,

permissive pressure is typically accomplished by condensing ,

steam from the RPV in the main condenser. When the reactor pressure is below the RHR cut-in permissive pressure, the RHR system may be operated in the shutdown cooling mode. Operation of the RHR system in the shutdown cooling mode above this pressure is not allowed because the coolant pressure may exceed ,

the design pressure of the shutdown cooling piping. In MODE 3 operation of a subsystem to remove the decay heat may be required to either reduce or maintain coolant temperature. .,

If ambient losses are insufficient continued operation of a '

shutdown cooling subsystem may be required to reach and maintain reactor coolant temperatures s 200*F, which i

corresponds to MODE 4. The requirements for decay heat removal in MODE 5 are discussed in LC0 3.9.8 and LCO 3.9.9.

(continued)

L L

Grand Gulf - Unit 1 B 3.4-25 DRAFT B 3/21/90 .

RHR - Shutdown l B3.4.7 BASES (cod inued) l ACTIONS- A.1. A.2. B.1. C.1 With one RHR shutdown cooling subsystem inoperable for decay .l heat removal, the remaining OPERABLE subsystem or aa alternate ~

method of decay heat removal can provide the necessary decay heat removal. The required cooling capacity of the alternate method should be ' ensured by verifying (by calculation or .

demonstration) its capability to maintain or reduce -

temperature. Decay heat removal by ambient losses can be considered as contributing to the alternate method capability.

Loss of one RHR shutdown cooling subsystem reduces the overall- l system reliability, therefore the subsystem should be restored ,

or an alternate method of decay heat removal should be provided 4 as soon as sracticable. If one inoperable subsystem cannot be restorec to OPERABLE status or with both subsystems -r inoperable, an alternate method of decay heat removal is required to be made available for each inoperable subsystem -

to restore cooling capability as soon as practicable.

d P SURVEILLANCE SR 3.4.7.1 RE0VIREMENTS Verification that all valves of the required RHR shutdown cooling subsystem are in the correct position ensures the '

proper flow path. Valves not in the correct position must be :

capable of manual realignment either from the control room or -l at the valve location. Verification of this capability is >

provided by actuation of the valve from the control room or the current inservice inspection reports. Operating experience ,

has demonstrated that a 31 day frequency for this surveillance is adequate. Because some of the required valves. are interlocked- closed when above the -RHR cut-in permissive pressure, an allowance is-provided to test the valves within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after pressure has been reduced - below. the cut-in permissive pressure. This allows conditions to be established .

under which the test may be performed. ;

(continued)

(..

1 H

Grand Gulf - Unit 1 B 3.4-26 DRAFT B 3/21/90

RHR - Shutdown .)

B 3.4.7 ;

BASES (continued)

REFERENCES 1. 52FR3788, Commission Policy Statement on Technical I Specification Improvements for Nuclear Power Reactors,.

February 6, 1987, i

i

2. NED0-31466, " Technical Specification Screening Criteria l Application and Risk Assessment," November 1987.

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Grand Gulf - Unit 1 B 3.4-27 DRAFT B 3/21/90

t

C Grar.d Gulf Nuclear Station i Technical Specification Improvement Program l 1

Revision Summary Sheet Proposed LCO/Section: 3.4.8 Rev. _1. P/T Limits !

1133 Chance Descriotion Cateaorv 1 LCO 3.4.8 is reformatted from LIMITING CONDITION 1 ;

FOR OPERATION 3.4.6.1. '

2 Figure 3.4.6.1-1 and the RCS temperature and 2 pressure limitations are relocated to the UFSAR.

3 The RCS heatup and cooldown rate limitations are 2 I relocated to the UFSAR.

l 4 LCO 3.4.6.1 Item c is deleted. 3B I l'

5 LCO 3.4.6.1 item d is relocated. 2 6 CONDITION A is reformatted from the ACTION 1 statement except as modified per Item 7.

7 REQUIRED ACTION A.2 has 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months specified for its 3B COMPLETION TIME. The time was unstated previously, 8 CONDITION B is reformatted from the ACTION 1 .

statement.

9 SR 3.4.8.1 is reformatted from SR 4.4.6.1.1. 1 ,

10 SR 3.4.8.2 is reformatted from SR 4.4.6.1.1. I 11 SR 3.4.8.5 is reformatted from LCO 3.4.1.4 1 and SR 4.4.1.4.

i 12 SR 3.4.8.4 is reformatted from LCO 3.4.1.4 1 and SR 4.4.1.4.

13 SR 3.4.8.5 is reformatted from SR 4.4.6.1.3. I 14 SR 4.4.6.1.4 is relocated. 2 15 SR 4.4.6.1.5 is deleted. The flux wire specimens -3B were removed during a previous outage.

16 Table 4.4.6.1.3-1 is relocated (see Item 14). 2 -

17 CROSS REFERENCE is added. 1 ,

RCS Pressure / Temperature Lioits 3.4.8 3.4 REACTOR COOLANT SYSTEM 3.4.8 Reactor Coolant System Pressure /Teenerature Limits LC0 3.4.8 The Reactor Coolant System (RCS) temperature and reactor ves:e1 pressure shall be maintained within the Pressure / Temperature (PT) Limits.

APPLICABILITY: At all times.

ACTIONS CON 0! TION REQUIRED ACTION COMPLETION TIME A. --------NOTE- ------ A.1 Restore RCS temperature 30 minutes Required Actions A.1 and reactor vessel and A.2 must be pressure to within the completed whenever PT Limits.

this Condition is entered. MlQ A.2 Determine RCS is 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 1 Operation outside the acceptable for continued PT Limits. operation.

B. Required Actions and B.1 Be ir: MOOL 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Times of Condition A 6@

not met. ..

B.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months !

l L Grand Gulf - Unit 1 3.4-13 DRAFT B 2/13/90

RCS Pressure / Temperature Limits 3.4.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.8.1 --

NOTE- ---------- ----

On1 required during system heatup, coo down and inservice leak and hydrostatic testing.

Verify RCS pressure and temperature are 30 minutes within the PT Limits.

SR 3.4.8.2 Verify RCS pressure and temperature are Once within within the PT Limit Curve criticality 15 minutes limit. prior to initial control rod withdrawal for the purpose of .

achieving criticality F

SR 3.4.8.3 ---

---.----..----NOTE----------------

On1 required in MODES 1,2,3, and 4 wit reactor steam dome pressure 1 25 psig.

l.

Verify the difference between the bottom Once within-head coolant temperature and the reactor 15 minutes pressure vessel coolant temperature is prior to each s 100'F. startup of a recirculation loop pump (continued)

Grand Gulf - Unit 1 3.4-14 ORAFT B 2/13/90

RCS Pressure / Temperature Limits 3.4.8 SURVEILLANCE RE0VIREMENTS fcontinued)

SURVEILLANCE FREQUENCY SR 3.4.8.4 ----------------- NOTE----------------

Only required in MODES 1,2,3, and 4.

Verify the difference between the reactor Once within coolant temperature within the recirculation 15 minutes loop to be started and the reactor pressure prior to each vessel coolant temperature is 150'F. startup of a recirculation loop pump SR 3.4.8.5 Verify the reactor vessel flange and head 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when flange temperature are 170'F. in MODE 4 with reactor coolant system temperature s 100*F .-

M 30 minutes when in MODE 4 with reactor coolant system temperature 1 80'F g.

30 minutes when tensioning the reactor vessel head bolting studs Grand Gulf - Unit 1 3.4-15 DRAFT B 2/13/90 l

i RCS Pressure / Temperature liaits 3.4.8 i CROSS-REFERENCES ~ !

)

TITLE NUMBER i Rectreulation loops Operating 3.4.1

-I*eeev4:: L::h ::d "ydr::t:th T::tir;; ^;;;r:ti;r, 3.10.1 l ]

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L. . Grand Gulf - Unit 1 3.4-16 ORAFT B 2/13/90 a

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- . . .. . - - ~_ . .- . . .- - . . - . - . .

l RCS Pressure / Temperature Linits '

B 3.4.8 8 3.4 REACTOR COOLANT SYSTEM !

B 3.4.8 Reactor Coolant System Pressure /Temnerature Limits

BASES B

BACKGROUND All components in the reactor coolant system (RCS) are designed to withstand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are '

introduced by normal load transients, reactor trips, and i startup-and shutdown operations. During startup and shutdown, i the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are- ,

consistent with the design assumptions and satisfy the stress 4 limits for cyclic operation. ;

The purpose of this specification is to establish operating limits that provide a conservative margin to brittle failure of major piping and pressure vessel components of the Reactor Coolant Pressure Boundary (RCPB). Of the major components within the RCPB, the reactor vessel (including feedwater nozzles, vessel flanges, shell and closure studs) is the component most . .

subject to brittle failure and therefore the component for which the technical specification limits is most pertinent.

The basis of the pressure and temperature (PT) limits is found in Appendix G to 10 CFR 50 (Ref. 1). Appendix G requires that limits be established, and that the limits be based on specific ,

fracture toughness requirements for RCPB materials such that an adequate margin to brittle failure will be provided during '

operational occurrences. 10 CFR 50 Appendix G mandates the use '

of ASME Section III, Appendix G (Ref. 2).

The concern addressed by Appendix G is that undetected flaws could exist in the RCPB components, which if subjected to unusual pressure and/or thermal stresses, could result in ,

non-ductile (brittle) failure. Certain reactor coolant system

PT combinations can cause stress concentrations at flaw ,

locations which in turn could cause flaw growth, resulting in

, failure before the ultimate strength of the material is attained. Flaw growth is resisted by the material toughness. '

Toughness is a material property which depends on the alloy microstructure. Toughness of steels vary with ambient temperature, and is lower at room temperature than at reactor operation temperature. Furthermore, toughness is negatively affected by neutron irradiation. The cumulative effect of neutron irradiation (fluence) causes the toughness to decrease with exposure. The region of the reactor vessel exposed to high neutron irradiation is defined as the reactor vessel beltline or ,

e (continued) ,

i Grand Gulf - Unit 1 B 3.4-28 DRAFT B 2/13/90 t

RCS Pressure / Temperature Limits B 3.4.8 BASES fcontinued)

BACKGROUND Beltline. This is comprised of the region of the reactor vessel (continued) that directly surrounds the effective height of the active core and adjacent regions that are predicted to experience high neutron irradiation.

Linear elastic fracture mechanics (LEFM) methodology, following the guidance given by 10 CFR 50 Appendix G, ASME Section III Appendix G, and Regulatcry Guide 1.99, is used to determine the

-stresses and material toughness at locations within the RCPB.

Although any region within the pressure boundary is subject to non ductile failure, the regions that provide the most 'i restrictive limits are the vessel closure head flange, the feedwater nozzles, the control rod drive nozzles, and the '

vessel beltline.

One indicator used to indicate the temperature effect on ductility is the nil-ductility transition (NDT) temperature.

The NOT temperature is a temperature below which it can be said i that brittle fracture may occur. Ductile failure may occur above the NDT temperature. The NDT temperature is integrated into a reference temperature (RT,,o,) by testing. RT,,oris a key,"

indicator of ductility that is used for steels in pressure vessel construction. The neutron embrittlement effect on the material toughness is reflected by increasing the RT,,,, as exposure to neutron fluence increases. In effect the temperature at which brittle failure can occur increases.

of the vessel material will be The actual shift established in RT,,E during operation by removing and periodicall evaluating in accordance with ASTM E185-73 and 10 CFR 50, -

Appendix H, irradiated reactor vessel material specimens installed near the inside wall of the reactor vessel in the core area. The operating limit curves shall be adjusted, as required, on the basis of the specimen data and recommendations of Regulatory Guide 1.99 (Ref. 3 and 4).

l The PT curves are composite curves established by superimposing L limits derived from stress analyses of those portions of the reactor vessel and head that are the most restrictive. At any specific pressure, temperature, and temperature rate of change, one location within the geometry of the reactor vessel will dictate the most restrictive limit. Across the entire r pressure / temperature span of the limit curves, some locations i

are more restrictive, and thus the curves are composites of the most restrictive regions.

(continued) l Grand Gulf - Unit 1 B 3.4-29 DRAFT B 2/13/90 L

L 0 .

RCS Pressure / Temperature Linits :

B 3.4.8 l J

BASES (continued)

BACKGROUND The curves have been developed for heatup, in-service leak and !

(continued) hydrostatic testing, and cooldown in conjunction with stress .

analyses for a large number of operating cycles and )rovide a l conservative margin to non ductile failure. .Althoug) they have I been created to provide limits for these specific normal ;

operations, they also can be used as a basis for determining if '

evaluations are necessary for abnormal transients which can e begin from power operation. '

~ APPLICABLE The limits are not derived from design basis accident analyses SAFETY presented in the FSAR, but are prescribed as limits to be ANALYSES used during normal operation to avoid encountering pressure, .

temperature, and temperature rate-of change conditions which might cause undetected flaws to propagate, in turn causing non ductile failure of the RCPB.

RCS Pressure / Temperature Limits satisfies the requirements of the NRC Interim Policy Statement on Technical Specification Improvements for Nuclear Power Reactors (Ref. 5). While none ,

of the three Selection Criteria (Ref. 7) directly apply, this l specification preserves limits defining important boundaries for safe operation derived from the RCS stress analysis.

Criterton 2 is the most appropriate criterion because operation outside of these boundaries is unanalyzed and may result in RCPB failure.

F LC0 Compliance with the following PT limits is required by this LCO: ,

1. Operation within the PT limit curves specified in Figure B 3.4.8 1, ,

2. A maximum reactor coolant heatup or cooldown of 100'F in -

any one hour period, l

l 3. A maximum temperature change of s 10'F in any one hour I period during inservice hydrostatic and leak testing ,

operations above the heatup and cooldown limit curves in l Figure B 3.4.8 1, l (continued) l .

Grand Gulf - Unit 1 B 3.4-30 DRAFT B 2/13/90 >

{

RCS Pressure / Temperature Limits .

B 3.4.8 .

BASES (continued). .

LCO 4. The reactor vessel flange and head flange temperature :

4 (continued)~ > 70'F when reactor vessel head bolting studs are i iinder tension,

5. A temperature difference between the bottom h'ead coolant temperature and the reactor pressure vessel coolant temperature of < 100'F during recirculation pump !

startup, and

6. A temperature difference between the reactor coolant temperature within the recirculation loop to be started and the reactor pressure vessel coolant of < 50'F ~

s

-ourir.g recirculation pump startup. .

The above limits define allowable operating regions and permit a large number of operating cycles while also providing a wide-margin to non-ductile failure. :

APPLICABILITY The potential for violating the PT limits exists at all times' . .

when the reactor coolant system can be pressurized. The temperature rate of change limit can be potentially v41sted any time the reactor vessel is a-different temperature from a cooling' source.

ACTIONS 1 1. A.2

. As noted, Required Actions A.1 and A.2 must be comaleted whanever Condition A is entered. The purpose of t1e Note is to give additional emphasis to.the need to restore operation to the allowable condition and to also perform an evaluation of

'the effects of any excursion.outside of the allowable limits. 3

' Restoration alone is insufficient because higher than . analyzed -

t stresses may have occurred and may have affected the RCPB ]

integrity.

Restoration within the limits is appropriate because the action

$ is'in the proper direction to reduce RCPB stress.

(continued) b L

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p Grand Gulf - Unit 1 B 3.4-31 DRAFT B 2/13/90

RCS Pressure / Temperature Limits B-3.4.8 BASES feontinued) _ . _ . . , _.

ACTIONS - . -A.J. A.2 (continucil (continued)

The Completion Time mt .,f 30 minutes is based on engineering judgement. Most violations.will not be so severe that the activity cannot be accomplished in this time in a controlled manner; however, if the activity cannot be accomplished, then a controlled shutdown must be initiated per Required Actions B.1 and B.2.

-In addition to restoration, an evaluation to determine if RCS operation may proceed is. required. The purpose of the evaluation is to determine if RCPB integrity is acceptable and must be accomplished before the event is reconciled.

If the evaluation cannot be accomplished in 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , cr if the-results of the evaluation are indeterminate or unfavorable, then the next appropriate action is to further reduce pressure and temperature as required in Condition B.

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is based on engineering judgement-and is reasonable to accomplish the activities necessary. For r, a mild violation the evaluation should be possible within'tnis time. More severe violations may require special, event specific stress analyses and/or inspections, which are appropriately carried out while the RCS is in a reduced pressure and temperature condition as required by Condition B.

B.1. B.E If the-Required Actions and associated Completion Times are not

-met, a controlled shutdown must be initiated. This is a prudent action when the RCS-remained in an unacceptable region-for an extended period of increased stress or a sufficient severe event caused entry into an unacceptable region. Either

" possibility indicates a need for more careful examination of the event, which is best accomplished while the RCS is in a low pressure and temperature state. With the plant at reduced pressure conditions the possibility of propagation of undetected flaws is reduced. The times allowed for a controlled shutdown to MODE 4 are reasonable and avoid placing undue stress on' plant operators or plant systems.

(continued).

Grand Gulf - Unit 1 B 3.4-32 DRAFT B 2/13/90

RCS Pressure / Temperature Limits -l B 3.4.8 _

j

l JASES (continued) 1 SURVEILLANCE SR 3.4.8.1 i REQUIREMENTS Verification that operation is within limits is an appropriate surveillance when RCS temperature and pressure conditions are undergoing planned changes. .The time period of 30 minutes is

-based on engineering judgement. Since temperature rate of change limits are specified in hourly increments, a half hour 1 time period permits assessment and correction of minor l deviations within a reasonabis time. 1 SR 3.4.8.2 :

A separate limit is used when the reactor is critical.

Consequently, it is appropriate to verify that the RCS pressure and temperature-are within the appropriate limit prior to ther withdrawal of control rods that will make the reactor critical.

SR 3.4.8.3. SR 3.4.8.4 Differential ~ temperatures within the limits of.these .

-surveillances will ensure that thermal stresses resulting from an idle recirculation pump startup will not exceed design allowances. In addition, compliance with the limit stated in SR 3.4.8.4 ensures that the assumptions of the idle ,

recirculation loop startup analysis (Ref. 6) are satisfied. -

Performing the surveillance within 15 minutes-before starting the idle recirculation pump provides adequate assurance that the limits will not be exceeded between the time of-the surveillance- :

- and the time of.the idle pum) start. An acceptable means of demonstrating compliance wit) the 50'F-requirement in SR 3.4 8.4' .

is to compare the temperatures between the operating recirculation loop and the idle loop.

SR 3.4.8.5 '

Limits on-the reactor vessel flange and head flange temperature 1 (required when.the vessel head in tensioned) are sometimes i bounded by the other PT limits during system heatup and cooldown.'.However, during operation in M00E 4, with RCS temperature less than or equal to 100 F, surveillance of the <

flange temperatures is required to ensure the 70 F temperature limit-is not violated. With RCS temperature less than.or-equal to 80+F, a more frequent check of the flange temperatures is required because of the reduced margin to the limit. The flange temperatures must also be verified to be above the limit

' prior-to and during tensioning of the vessel head bolting studs to ensure that once the head is tensioned the limit is satisfied.

(continued)

Grand Gulf - Unit 1 B 3.4-33 DRAFT B 2/13/90

- , - * < , ~

RCS Pressure / Temperature Liaits B 3.4.8

BASES (continued)

SURVEILLANCE- Frecuencies REQUIREMENTS-(costinued) In general, survelliance frequencies are based on industry accepted practice and engineering judgement considering the unit conditions required to aerform the test, the ease of performing the test and a licelihood of a change in the systeWcomponent status.

REFERENCES 1. . Code of Federal Regulations, Title 10, Part 50, Appendix G, " Fracture Toughness Requirements."

2. American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code, Section III, Appendix G,

" Protection Against Non-Ductile Failure."

3. USNRC Regulatory Guide 1.99, Revision 1, " Effects of Residual Elements on Predicated Radiation Damage to Reactor Vessel Materials," April 1977.

~

4. USNRC Regulatory Guide 1.99, Revision 2, " Radiation Embritticent of Reactor Vessel Materials," May 1988.

5. 52FR3788, Interim Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, USNRC, February 6, 1987.

6. Grand Gulf Unit 1 FSAR, Section 15.4.4.

7. NED0-31466, " Technical Specification Screening Criteria Application and Risk- Assessment," November 1987..

l l

Grand Gulf - Unit 1 B 3.4-34 DRAFT B 2/13/90 1

~-

. RCS Pressure / Temperature Limits' B 3.4.8 l I 4

A BB' CC'

.1400

~ '

7 r, r A - INITIAL SYSTEM HYDROTEST LIMIT ETRA 0 $ ""

[ B - INITIAL NON. NUCLEAR HEATING LIMIT l , , _, C* INITIAL' NUCLEAR (CORECRITICAL).

~

I 1200 A'.B'.C'-A.B.C LIMITS AFTER AN i

__ ASSUMED 26*F CORE DELTLINE TEMP.

j j- SHIFT FROM AN INITIAL SHELL PLATE OF 0*F. A' !$ NOT SHOWN

, gf f_ RT NDT

' '/ f f ~ (NOTLIMITING)

/ f [ B' and C' are coincident 1000 f g g with 8 and C. respectively, Core Belt I I

=

' I ~Line after f . Shift i / '

/ '

I 300

.g. / Curves A. 8 and C are predicted to r be applicable for service periods up to 32 EFPY, g -}_

A a % FEEDWATER r i N0ZZLE 600 '

$- I- LIMITS b

7 3'

g. 400 [k ~I

-- ~ -~

l.

Acceptable region of operation is to the right of the a-A applicable curve.

312 Psig f /--

80LTup

/i / l i 200' ?N FEEDWATER O/ N0ZILE ~

> . / g LIMITS - ' . 1

. l 0 100 200 300 400 500 600 RPV Metal Temperature '('F) -

MINIMUM REACTOR PRESSURE-VESSEL METAL TEMPERATURE VS. REACTOR VESSEL PRESSURE Figure B 3.4.8-1 GRAND GULF - UNIT 1- B 3.4-34a DRAFT B'2/13/90

1

. Grand Gulf Nuclear Station 1 o Technical Specification Improvement Program i Revision' Summary Sheet n l

. Proposed LCO/Section: 3.4.9 Rev. _1_ Steam Dome Pressure e 1133 Chanae Descriotion .Cateoorv I 1 LCO 3.4.9 and its applicability are reformatted 1 from LIMITING CONDITION FOR OPERATION 3.4.6.2 excepti as discussed below. ;

o~ i 2 CONDITIONS A and B are reformatted from the ACTION 1 statement except as discussed below.

3 SR 3.4.9.1 is reformatted from SR 4.4.6.2 except as 1 discussed below.

4 The '*' footnote to page 3/4 4-23 is deleted I because the applicability for LC0 3.4.9 includes this provision.

^

5 The CTS LCO 3.4.6.2 does not allow the pressure to be 38 equal to 1045 psig whereas the PSTS LCO will. The REQUIRED ACTION A.1 and SR 3.4.9.1 are similarly affected..

k i

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i

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Reactor Steam Dome Pressure a 3.4.9 i 3.4 REACTOR COOLANT SYSTEM e 3.4.9 . Reactor Steam' Dome Pressure LCO 3.4.9 The reactor steam dome pressure shall be $ 1045 psig.

APPLICABILITY: MODES 1 and 2, except during anticipated transients. l i

I ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME ;

-i A. . Reactor steam dome A.I Reduce reactor steam 15 minutes pressure. dome pressure to ;

> 1045 psig. 1 1045 psig.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A-not met.

SURVEILLANCE REOUIREMENTS SURVEILLANCE FREQUENCY-SR 3.4.9.1 Verify reactor steam dome pressure is 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

$ 1045 psig.

CROSS-

REFERENCES:

None Grand Gulf - Unit 1 3.4-17 DRAFT B 11/22/89

Reactor Steam Dome Pressure B 3.4.9 j B 3.4 _ REACTOR COOLANT SYSTEM-D B 3.4.9 Reactor Stear Dome Pressure 1 1

BASES l BACKGROUND The reactor steam dome pressure is an assumed initial condition l of design basis accidents and transients.and is also assumed in l the determination of compliance with reactor pressure vessel overpressure protection criteria.- l APPLICABLE' .The reactor steam dome pressure is an initial condition of the SAFETY vessel overpressure protection analysis of Reference 1. This ANALYSES analysis assumes an initial maximum reactor steam dome pressure and evaluates the response of the pressure relief system--

(primarily the safety / relief valves) during the limiting ,

pressurization transient. The determination of compliance with the overpressure criteria is dependent on the initial reactor steam dome pressure and therefore, the limit on this pressure ensures that the assumptions of the overpressure protection analysis are conserved. Reference 2 a' Iso assumes an initial reactor steam dome pressure for the analysis of design basis accidents and transients used to determine the limits-for fuel cladding integrity (MINIMUM CRITICAL POWER RATIO, see Bases for LCO 3.2.2 and 1% cladding plastic strain see Bases for'LCO 3.2.1 and LCO 3.2.3).

Reactor Steam Dome Pressure satisfies the requirements of Selection Criterion 2 of the NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 3.

LC0 - -The specified reactor steam dome pressure limit assures the plant is operated within the assumptions of the transient :

analyses. ' Operation above the limit may result in a transient s response more severe than analyzed, q (continued) t Grand Gulf - Unit 1 B 3.4-35 DRAFT B 11/22/89

Reactor Steam Dome Pressure B 3.4.9 BASESicontinuedii APPLICABILITY 1The reactor steam dome pressure is required to be less than or equal to the limit in MODES 1 and 2 where the reactor is generating significant steam and the design basis transients and accidents are bounding.- The limit may be exceeded during anticipated transients since the evaluations of References 1 and 2 demonstrate that appropriate reactor and fuel limits are not exceeded.

The' limit is not applicable in MODES 3, 4 and 5, because in these modes the reactor is shutdown. The reactor pressure is well below the required limit and no anticipated events will .

challenge _ the overpressure limits.

ACTIONS A.1. B.1 If the reactor steam dome pressure is greater than the limit, prompt action should be taken to reduce the pressure to below the limit. If the operator is unable to reduce the reactor steam dome pressure to the limit, then the reactor is required f to be in MODE 3.

Comoletion Times All Completion Times are based on industry accepted practice and engineering judgement considering the number of available systems and the time required to reasonably complete the Required Action.

SVRVEILLANCE SR 3.4.9.1 REQUIREMENTS The reactor-steam dome pressure is verified to be less than or equal to the limit every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This ensures that the initial conditions of the design basis accidents and transients are met. Operating experience-has demonstrated that the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months frequency for this surveillance'is adequate.

(continued) 4 Grand Gulf - Unit 1 B 3.4-36 DRAFT B 11/22/89 :

-Reactor Steam Dome Pressure--

B-3.4.9'

BASES (continued)

REFERENCES- ~ 1. - Grand Gulf FSAR, Section 5.2.2.

2; Grand Gulf FSAR, Section 15, l

3. .- NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment," November 1987.-

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-Grand Gulf - Unit 1 B 3.4-37 DRAFT B 11/22/89

N

'i CHAPTER 3.5 ECCS AND-RCIC fi

.c 1

i

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n.

CHAPTER 3.5~

+ ECCS and RCIC

! TABLE OF CONTENTS

.3.5.1 'ECCS . Operating

-:3 . 5 . 21 ECCS - Shutdown 3.5.3 Reactor Core Isolation Cooling System 0

--__--___n- - _ . - - - - . - _ . - - . = _ - - - _ - - _ _ _ _ _ _ _ _ _ _ _ _ - . - _ _ _ _ _ . _ - _ - - - - _ _ _ _ _ _ _ _ . . - _ _ _ _ _ _ _ - _ _ _ - _ _ - _ _ _ _ - - _ - - . - _ - _ - - - - ~ - _ - _ - _ . - -

Grand Gulf Nuclear Station

-Technica1' Specification Improvement Program Revision Summary Sheet Proposed LCO/Section: 3.5.1 Rev _1_ ECCS-Ocerating 1115 Chance Descriotion Cateoorv 1 LCO'3.5.1 is' reformatted from LIMITING CONDITION 1 FOR OPERATION-3.5.1.

2 - The details of system operability requirements are 2

-relocated to the BASES.

3 DELETED 4 The footnote '#' to page 3/4 5-1 for applicability 3A in MODE 2 is deleted. Special Test Exception 3.10.5 will be reviewed as a supplemental specification.

5 CONDITION A is reformatted from ACTIONS a.1, a.2, and 1-b.1.

6 CONDITION B was developed from ACTIONS a.3, b.2, d.1 3B and d.2 with the added flexibility that any two ECCS injection / spray systems may be inoperable as long as they are not LPCS and HPCS simultaneously.

7 REQUIRED ACTION B.4 is added to require the remaining 3A+

. inoperable ECCS subsystem to be restored within 7 days.

8 CONDITION C is reformatted from ACTION col. I 9 REQUIRED ACTION C.1 limits RCIC OPERABILITY to when 3B the system is required.(LCO 3.5.3).;

10 CONDITION D is reformatted from ACTIONS a.4, b.3, c.2 1 and d.3.

11 CONDITION E is reformatted from ACTION e.l. 1 12 CONDITION G is reformatted from ACTIONS e.1 and e.2. I 13 SR 3.5.1.1 is reformatted from SR 4.5.1.a.1. I 14 The method of verifying that system piping is 2 filled in SR 3.5.1.1 is relocated.

15 SR 3.5.1.2 is reformatted from SR 4.5.1.a.3. 1 16 A NOTE is added to SR 3.5.1.2 to provide for LPCI 3B+

OPERABILITY when in SDC mode under specified conditions.

i y

Grand Gulf Nuclear Station *

, . Technical Specification IGprovement Program-Revision Summary Sheet Proposed LCO/Section: 3.5.1 Rev. _1_ ECCS-Oneratino !

lj;.gg . Chance Descriotion Cateaorv 17 SR 3.5.1.$ is added to define surveillance 3A+

-requirements for ADS. #

~18 SR 3.5.1.4 is reformatted from SR 4.5.1.b. 1 19 SR.3.5;1.5 is reformatted from SR 4.5.1.c.1 1 20 SR 3.5.1.6 is reformatted from SR 4.5.1.d.1. 1 21 SR-3.5.1.7 is reformatted from SR 4.5.1'.d.2 l' and footnote '*' to page 3/4 5-5. <

22 The method of verifying ADS valve operation in 2-SR 3.5.1.7 is relocated.

23~ CROSS REFERENCES are added. 1 :

24 Footnote '**' to page 3/4'5-1 and footnote '*' to 2 pages 3/4 5-2 and 3/4 5-3 are deleted. The

" low as practical" provision is relocated to the <

BASES.

25 LCO 3.0.2 does not perm;. more than one CONDITION- 1 :

for LC0'3.5.12to be entered at a time. . Therefore, the '

_-provisions in ACTIONS a, b, c, d.and e to have the remaining ECCS subsystem (s) OPERABLE is not. required in the CONDITIONS for LCO 3.5.1.

26 ACTION f is deleted. 4 :

27 ACTION g is deleted. 4 1 l

.28 ACTION h is moved.to Section 5. 1 29 ACTION i is deleted. 4 30 SR 4.5.1.a.2 is deleted. This change is considered 4 administrative because the " keep-filled" and delta P ,

I. I instrumentation ACTIONS are deleted (see Items 26 and 27).

31- SR 4.5.1.c.2 is deleted. This change is considered 4 ,

administrative because the " keep filled" and delta P !

instrumentation ACTIONS are' deleted (see Items 26-and 27).

L 32 SR 4.1.5.c.3 is covered by SR 3.5.1.5 and is 2 explicitly described in the BASES for SR 3.5.1.5.

Grand Gulf Nuclear Station.

' - Technical Specification Improvoment Program Revision Summary Sheet Proposed LCO/Sectioni 3.5.1 Rev. _1_ ELCS-Oneratina-ligg Chance Descriotion Cateoorv 33 SR 4.1.5.c.4 is deleted. 4 SR 4.1.5 d.3 is deleted. 4 35- SR 4.1.5.d.4 and SR 4.1.5.e are relocated. Alarm 2

only instrumentation (except those required by RG 1.97) are not included in the Improved Tech Specs.

36 The '*' footnote to page 3/4 5-1 is deleted since it 1 is now included in the applicability statement.

37. REQUIRED ACTION B.2 is added to' require RCIC to be 1+

OPERABLE if HPCS is one of the ECCS injection / spray syrtems that is inoperable. -(See item 6),

38 REQUIRED ACTION B.1:is added.to ensure that LPCS and 1+

HPCS are not simultaneously inoperable (see item 6).

39 CONDITION F and REQUIRED ACTIONS F.1, F.2 and F.3 3B+

are added to allow one ADS valve and one ECCS injection / spray system to be inoperable.

1 n

(

ECCS - Operating 3.5.1 3.5~ ECCS AND RCIC--

3.5.1 ECCS - Operatine

! LCO 3.5.1 All ECCS injection / spray systems shall be OPERABLE, 8!!D 8' ADS valves shall be OPERABLE.

APPLICABILITY:- MODE 1, '

MODES 2 and 3 except ADS.is not required to be OPERABLE with reactor steam dome pressure $ 135 psig.

' ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME

.A. One low pressure A.1 Restore inoperable 7 days from ECCS injection / spray system to OPERABLE . discovery of system inoperable, status, inoperable system

-B. Two ECCS injection / spray B.1 Verify at least one ECCS Immediately systems inoperable. spray system to be-OPERABLE.

8!!Q

. B.2 ----------NOTE--------- Immediately Required-Action B.2-applicable only when HPCS system is operable.

Verify RCIC is OPERABLE Immediately when required to be OPERABLE.

A!!D B.3 Restore at least one 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable system to OPERABLE status.

8!{Q (continued)

Grand Gulf - Unit 1 3.5-1 DRAFT B 5/15/90 i

l

ECCS - Operating.

3.5.1 ACTIONS (continued) ,

i' C0hDITION REQUIRED. ACTION COMPLETION TIME B.4 Restore the remaining -7 days'from inoperable system to discovery of OPERABLE status, initial

. i_noperabl e -

system l

C. HPCS inoperable. C.1 Verify RCIC is . Immediately OPERABLE when required '

to be OPERABLE. -!

1 M

C.2 Restore HPCS to OPERABLE 14 days from status, discovery of .

inoperable l system

-1 >

D. Required Actions and D.1 Be in H0DE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Times of Condition A, B or C not met.. E J 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months D.2 Be in MODE 4. ,

E. One ADS valve E.1 Restore inoperable ADS- 14 days from inoperable, valve to OPERABLE discovery of status, inoperable valve 1

F. One ADS valve F.1 ---------NOTE---------- u 3- .

inoperable. Required Action F.1 '

applicable only when' '

E HPCS system is inoperable.

One ECCS injection / spray -----------------------

system inoperable.

Verify RCIC is OPERABLE Immediately when required to be OPERABLE.

t E (continued)

Grand' Gulf - Unit 1 3.5-2 DRAFT B 5/15/90 b______'__ _--_.m-__[ - - - - - - _ _ m. ----- - _ -_ _- _ _ _

y--- ,

~

ECCS -~ Operating .

3.5.1 .i

ACTIONS (continued)- '

_CONDITIONL REQUIRED ACTION COMPLETION TIME.

F.2-- Restore _ the ' inoperable 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months #

injection / spray system- '

or ADS Valve to OPERABLE status AND F.3 Restore the remaining 7 days from- [

inoperable injection /. discovery of spray system or' ADS initial-

valve to'0PERABLE- inoperable-status, system / valve

>t il a G. -Two or more' ADS G1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months '

valves inoperable.-

'AND QR G.2 Reduce reactor steam: '36 hour4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months s-Required Action and dome pressure to associated Completion 5 135 psig.

Time of Condition-E or ,

'F notimet.

I.

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l Grand Gulf - Unit 1 3.5-2a DRAFT B 5/15/90 1

i

^ ~ '

5 ECCS - Operating. ,

3.5.1 SURVEILLANCE RE0VIREMENTS

~

SURVElLLANCE- FREQUENCY _- g SR~ 3.5.1.1 Demonstrate for each ECCS injection / spray 31 days i system the system piping is filled with water from the pump discharge valve to the system isolation valve. '

i SR' 3.5.1.2 -----------------NOTE-----------------

LPCI subsystems may be considered .

OPERABLE during alignment to and .?

operation in- the RHR shutdown cooling i mode ~when below the RHR cut-in permissive pressure in MODE 3, if capable of being manually realigned and not otherwise inoperable.

Verify for-each ECCS injection / spray 31 days system each manual, power operated or-automatic' valve in the flow path not )

locked, sealed or otherwise secured in position is in its correct position.

SR 3.5.1.3- Verify ADS air receiver. pressure is 31 days 2_150 psig.

'SR-.3.5;1.4 Demonstrate the following ECCS pumps According to develop the specified flow rates with SR 3.0.5 the specified total developed head:

t N0.- TOTAL OF DEVELOPED SYSTEM FLOW RATE PUMPS HEAD _

LPCS 2 7115 gpm 1 2 290 psid 3 LPCI 2 7450 gpm 3 2 125 psid HPCS 2 7115 gpm 1 2 445 psid i

(continued)

Grand Gulf - Unit 1 3.5-3 DRAFT B 5/15/90

< s Y 'ECCS - Operating 3.5.1

~ SURVEILLANCE REOUIREMENTS (continued)-

SURVEILLANCE FREQUENCY SR 3.5.1.5- -----------------NOTE ----------------

Vessel injection may be excluded.

Perform a system functional test for .18 months each ECCS injection / spray system,

! including simulated automatic actuation

- of the system throughout its emergency operating sequence,'to verify each automatic valve in the flow path actuates to its correct position.

SR 3.5.1.6 ----------------NOTE-----------------

Valve, actuation may be excluded..

Perform a system functional test for ADS, 18 months including simulated automatic actuation, throughout its emergency operating sequence.

SR 3.5.1.7 Demonstrate each ADS valve opens when manually actuated at reactor steam dome pressure > 100 psig.

NOTE------

Only required within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when reactor steam dome pressure is adequate to perform the test.

18 months Grand Gulf - Unit 1 3.5 4 DRAFT B 5/15/90

ECCS - Operating

~3.5.1 CROSS-REFERENCES TITLE NUMBER.

ECCS Actuation Instrumentation 3.3.5.1

-Safety / Relief Valves 3.4.4 Residual Heat Removal - Shutdown 3.4.7 Reactor Core Isolation Cooling System 3.5.3 Residual Heat Removal Suppression Pool Cooling 3.6.2.3 Grand Gulf - Unit 1 3.5-5 DRAFT B 5/15/90 3

g -

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ECCS.- Operating '

B 3.5.1

'B-3.5 ECCS AND RCIC- ;

Y B 3.5.1 ECCS - Operatina BMES' BACKGROUND. TheECCS is designed, in conjunction with the primary cnd secondary containment to limit the release _of radioactive i materials to the environment following a Loss-of-Coolant .

-Accident (LOCA). The ECCS uses two independent methods (flooding and spraying). to cool ~the core during a LOCA. The-ECCS injection / spray network is comprised of the High_ Pressure Core Spray (HPCS) system, the Low. Pressure Core Spray (LPCS) -!

system and the Low Pressure Coolant injection (LPCI) mode of- i the Residual Heat Removal .(RHR). system. The ECCS also consists of- the Automatic Depressurization System (ADS). The 1, supprestion pool provides the source of water for:the ECCS.

Although no credit is taken in the safety analyses for the' '

Condensate Storage Tank (CST) it is capable of providing a :

source of water for the HPCS system. ',

All ECCS systems are designed to ensure no single active-component failure in any system will prevent automatic

. in;tiation and successful operation of the minimum. required l- ECCS systems.

1 The LPCS system (Ref.1) consists of a motor-driven pump, ~a i spray sparger above the core, piping and valves to transfer 3 water from the suppression pool to the. sparger. The LPCS.

t- system is designed to provide cooling to the reactor core when ;

/ the reactor pressure-is low. _Upon receipt of an initiation l ' signal, the LPCS pump is automatically started'(from normal-l A. C. power if available, otherwise, the pump starts after t emergency A.C. power becomes available). When the reactor vessel pressure drops sufficiently, the injection valve opens i- and LPCS flow to the -reactor vessel begins. A full-flow test _

line is provided to route water from and to the suppression L

E pool to allow testing, when required,- of the LPCSJ system; .

l without spraying water- into the reactor vessel.- The test return valve automatically closes on an initiation signal. .t LPCI is an independent operating mode of the RHR system. There:

are three LPCI subsystems. Each LPCI subsystem (Ref. 2)- j consists of a motor-driven pump, piping and valves to transfer- a water from the suppression pool to the core. Each'LPCI' ,

i l',

[

(continued)

Grand Gulf - Unit 1 B 3.5-1 DRAFT B 5/15/90

ECCS - Operating B 3.5.1 BASES (continued)

BACKGROUND subsystem has its own suction and discharge piping and separate (continued) vessel nozzle which connects with the core shroud through internal piping. The LPCI subsystems are designed to provide core cooling at low reactor vessel pressure. Upon receipt of an initiation signal, each LPCI pump is automatically started (from normal A.C. power if available, otherwise, the pumps start after emergency A.C. power becomes available). When the reactor vessel pressure drops sufficiently, the injection valve opens and LPCI flow to the reactor vessel begins. With suction aligned to the suppression pool, the required RHR system valves in the LPCI flow path are automatically positioned to ensure the proper flow path for water from the suppression pool to injection into the core. The shutdown cooling and suppression pool suction valves will not automatically realign to the LPCI mode. A discharge test line is provided to route water from and to the suppressior. pool to allow testing, when required, of each LPCI pump without injecting water into the reactor vessel. The test return valves automatically close on an initiation signal.

The HPCS system (Ref. 3) consists of a single motor driven pump, a spray sparger above the core, and piping and valves to transfer water from the suction source to the sparger.

Suction piping is provided from the CST and the suppression pool. Pump suction is normally aligned to the CST, which is the preferred source of water for injection into the RPV when HPCS functions to backup RCIC. However, if the CST water supply is low or the suppression pool level is high, an automatic transfer to the suppression pool water source ensures a continuous water supply for continuous operation of the HPCS system. The HPCS system is designed to provide core cooling over a wide range of reactor vessel pressures 0 to 1177 psid, vessel to suction source. Upon receipt of an initiation signal, the HPCS pump automatically starts (from normal A.C.

power if available, otherwise, the pump starts after emergency A.C. power becomes available) and valves in the flow path begin to align to the positions required for injeci. ion. Since the HPCS system is designed to operate over the full range of expected reactor vessel pressures, HPCS flow begins as soon as the necessary valves are open. A full-flow test line is provided to route water from and to the CST to allow testing, when required, of the HPCS system during normal operation without spraying water in the reactor vessel.

The ECCS pumps are provided with minimum flow bypass lines which discharge to the suppression pool. The valves in these lines automatically open to prevent pump damage due to overheating when other discharge line valves are closed or reactor vessel pressure is greater than the LPCS or LPCI pump discharge pressures following system initiation. To ensure (continued)

Grand Gulf - Unit 1 8 3.5-2 DRAFT B 5/15/90

M, j ECCS'- Operating- ,

B.3.5.1 ,

BASES (continued)- 4 BACKGROUND! ' rapid delivery of water to the reactor vessel and to minimize (continued) waterhammer effects, the ECCS discharge line keep fill systems are designed to maintain all pump discharge lines filled with water. ~ .

The ADS (Ref. 4) consists of 8 of the 20 safety / relief valves j (S/RVs). It is designcd to perde depressurization:of the !

primary system during a small break LOCA if-HPCS fails or is !

" unable to maintain required water level in the reactor vessel. "

ADS-operation reduces the reactor vessel pressure to within i.

the operating pressure range of the low pressure ECCS systems.

(LPCS and LPCI), so that these- systems can - provideL eore-cooling. Each ADS valve is supplied with pneumatic power from an air storage system which consists of accumulators and large air flask located in the Drywell.

APPLICABLE. The ECCS performance is evaluated for the entire. spectrum of SAFETY break sizes for a postulated LOCA. The accidents for which ANALYSES ECCS operation is required are specifically listed in .FSAR Sections 15.2.8,15.6.4 and 15.6.5. The' required analyses and - :

assumptions are-defined-in Reference 5. The results of these a analyses are described in Reference 6.

The. ECCS system design requirements ensure. the criteria. of Reference 7.are satisfied under all postulated LOCA conditions assuming the worst single active component; failure in the ECCS.

The limiting single failures' are discussed; in Reference 8.

For a large break LOCA, failure-of ECCS systems in Division -

1 (LPCS and LPCI-A) or 2 (LPCI-B and LPCI-C) due to failure <

of its- associated-diesel- generator is, ;in general, the'most ;

severe failure.- For a small break LOCA, HPCS failure is the: ,

most-severe. failure. One' ADS' valve failure is analyzed as a ,

limiting single failure ~ for events requiring ADS operation. 1 The remaining OPERABLE ECCS systems provide the capability to adequately cool the core and prevent excessive fuel damage.

ECCS-Operating - satisfies the requirements of Selection Criterion 3 of NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference'll.- J (continued)

-I f

m Grand Gulf - Unit 1 B 3.5-3 DRAFT B 5/15/90

ECCS -_0perating B 3.5.1:

BASESicontinued)

^

- LCO All ECCS injection / spray systems and 8 ADS. valves are required to be OPERABLE. The ECCS injection / spray systems are defined as the_ three LPCI subsystems, the LPCS system and the HPCS system. The low pressure ECCS injection / spray. systems are:. .

g _ defined as the LPCS system and the three LPCI subsystems. A.

description of what is- required for the ECCS systems to be considered OPERABLE is provided in the Background.section.

With less than the required number of ECCS systems OPERABLE, the potential exists that during.a limiting design basis LOCA-t concurrent with the worst single failure, the limits specified-in Reference - 7 ~ could be exceeded. . . All- ECCS systems- must

.therefore be OPERABLE to satisfy the single failure criterion.

. required by Reference 7.

A LPCI subsystem may be considered OPERABLE during alignment-to and operation in the RHR shutdown cooling mode when helow-the RHR cut-in permissive pressure -in MODE 3, if capable of being manually realigned to the LPCI mode and not otherwise inoperable. At these low pressures and decay heat levels (reactor is shutdown.in MODE 3) a reduced complement of ECCS systems can provide the required core cooling thereby allowing--

operation of an RHR shutdown cooling loop when necessary.

APPLICABILITY All ECCS' systems are required-to be OPERABLE during MODES 1, 2.and 3 when there is considerable. energy in the reactor core and_ core cooling would be. required to prevent fuel-damage in-the event of'a break in the primary system piping. In MODES 2 and 3 the ADS function is not required when pressure is i 135 psig because the low pressure ~ECCS systems (LPCS, LPCI).

are capable of providing. flow into the reactor vessel ~ below -

this pressure. ECCS requirements- for MODES 4 and-~ 5 are specified in LC0 3.5.2.

(continued) l Grand Gulf - Unit 1 B 3.5-4 DRAFT B 5/15/90 jq 4

?M ECCS - Operating j B 3. 5.1' BASES (continued) ,

y

ACTIONS A.I. B.1. B.2. B.3. B.4. C.I. C.2 'I

'With one or any two .(except both LPCSL and HPCS)2 ECCS injection / spray systems inoperable, the remaining OPERABLE systems provide adequate core cooling during a LOCA. With the-HPCS system -inoperable, adequate core cooling. is- assured- by -

the OPERABILITY of the redundant and diverse low pressure ECCS .

systems in conjunction with ADS. Also, the Reactor Core:-

i

~'-

system, for which no credit is taken Isolation in the safety Cooling analysis, (RCIC) wi ll automatically provide makeup wate:l at most- reactor operating. pressures. - Verification of RCIC OPERABILITY is therefore. required when HPCS is inoperable, j This may be performed by an administrative ' check, by examining-

' logs or other information, to determine if RCIC'isi out - of i service'for maintenance or other reasons. It does'notLmean to perform the surveillance requirements needed to demonstrate -i the OPERABILITY of RCIC. However, with less than the minimum number of required ECCS injection / spray systems OPERABLE, the :

overall ECCS reliability is reduced because a single failure--

in one of the remaining subsystems concurrent with a LOCA, may -i result ~ in the ECCS not being able to perform;its = intended safety function. Therefore, continued operation is- only.

' allowed for a limited time, ,

D.1. D.2 Should the Required Actions and associated Completionlimes - !

of Condition A, B or C not be met, the reactor is required to be in MODE 3 and subsequently in MODE 4; In-MODE 4, the ECCS : .

requirements are specified'in LCO 3.5.2. If unable to attain :

MODE 4, the reactor coolant' temperature.should be maintained

, as low as practicable by' use of alternate decay heat removal methods. -

l (continued)l 7

.L Grahd Gulf - Unit 1 B 3.5-5 DRAFT B 5/15/90 l

I -

, ECCS -Operating B - 3. 5.1 '

BASES-(continued)

E.1. F.1. F.2. F.3. G.I. G.2 ACTIONS ~ 1 (continued)' ;

The LC0 ' requires 8 ADS valves to be OPERABLE to provide the l ADS function as designed. Reference 9 contains the results i of an analysis which evaluated the'effect of one ADS valve out- "

of service. Per this analysis,' operation of only 7 ADS valves will provide the required depressurization. However,' with one -

ADS valve inoperable,-the-overall reliability of the ADS is- -

? reduced and operation is only allowed for a limited time. With one ADS valve and one ECCS system inoperable, the overall ECCSl ,]

reliability is reduced because a single failure in one of.the -

remaining systems concurrent with a LOCA may result'in the ECCS not being. able to perform its-' intended safety function.- i Therefore, continued operation is allowed for only a limited' '

time. When more than _ one ADS - valve is ! inoperable, system capability may not be sufficient to .provided- the designed function. Therefore, if the one-inoperable ADS valve cannot' be made OPERABLE within the allowed Completion Time or with more than one' ADS valve inoperable, the reactor is required.

to be in MODE 3 and the reactor pressure reduced to 1135 psig.

At' these conditions the~ ADS function is no longer required 1.

since'the reactor pressure is low enough such that the: low-pressure ECCS subsystems can perform their ' designed safety function.

Comoletion Times The' ECCS Completion Times are based-on the.results of a -study-which evaluated the impact on ECCS unavailability assuming-various-components'and subsystems'were taken-out of service.

The results were used to calculate the average unavailability of ECCS equipment needed to mitigate the consequences of a LOCA-as a function of allowable outage times (A0T). A0Ts were then

. chosen ~ to provide comparable levels._ of ECCS availability throughout the industry (Ref.'12)'.

i SURVEILLANCE SR 3.5.1.1 1 REQUIREMENTS' The pump discharge lines of the HPCS, LPCS and LPCI systems are required to be kept full with water to minimize _ potential >

waterhammer effects when the systems are initiated. j Additionally, the = lag between the receipt of the initiation signal and the actual injection into the reactor vessel is minimized. One acceptable method of ensuring the lines are

" full" is to vent at the high points.

(continued)

Grand Gulf - Unit 1 B 3.5-6 DRAFT B 5/15/90

> 1 7

ECCS - Operating' B 3.5.1-l BASES'(continued)^

1

- SURVEILLANCE- SR 3.5.1.2 LREQVIREMENTS

.(continued) , Verification that all - valves are in the required position '

ensures proper flow paths for ECCS. However, a valve that is

, y, _ capable of automatic return to its ECCS position when an ECCS signal is present, can be: in position for another mode ~ of:

operation. This is applicable only ~1f the valve . auto- ,

repositions and fully opens within the time required for,its ECCS. function. As noted, a LPCI subsystem may be considered -

OPERABLE during alignment to and operation in the RHR shutdown - l cooling mode when below the RHR cut-in permissive pressure in MODE 3 if capable of being manually realigned to the LPCI mode 1.

and the system in not otherwise inoperable. This allows operation in the RHR shutdown cooling mode during MODE 3 if necessary.

SR 3.5.1.3 ,

The accumulator on each ADS valve provides pneumatic pressure-for valve' actuation. -The designed pneumatic supply pressure requirements for the accumulator are such that following a failure of _the pneumatic supply to the accumulator, at least.

two valve actuations can occur with the drywell at 70% of design pressure (Ref. 10). .The ECCS safety analysis assumes only one- actuation to achieve- the depressurization required for . requirement provides sufficient - margin to satisfy ' the, assumptions of the safety -analyses. .This minimum required

- pressure of 150 psig is provided by the Instrument Air System.

SR 3.5.1.4 ,

The performance requirements of the ECCS pumps are determined through application of the 10 CFR 50 Appendix K criteria (Ref, a 5). The pump flow rates, as determined by analysis, ensure -

that : adequate core cooling is provided to satisfy the acceptance criteria of Reference' 7. Periodic surveillance is

performed -(in accordance with ASME Section XI requirements) '

to verify these flow rates; The pump flow rates are verified- .

against a system head that is equivalent to the reactor vessel l pressure expected during a LOCA. The total system head -

developed is adequate to overcome the elevation differences. '

between the suction source and the vessel, friction losses and pressure differences present during LOCA. These' values were i established during preoperational testing.

~

(continued)

Grand Gulf - Unit 1 B 3.5-7 DRAFT B 5/15/90

ECCS - Operating B 3.5.1 BASES (contj3y g t __

n +

SURVEILLANCE SR 3.5.1.5 REQUIREMENIS ,

(continued) The ECCS systems are required to actuate automatically to '

perform their design function. These surveillance tests !

demonstrate that with the required system initiation sii1nals. '

the automatic initiation logic of HPCS, LPCS, and LPC'. will !

cause them to operate as designed, including actuation of all

~

automatic valves to their required position. This test also ensures the HPCS system will automatically restart on a reactor ve',sel low water level (Level 2) signal received subsequent tr, reactor vessel high water level (Level 8) trip and that the fuction is automatically transferred from the CST to the ,

suppression pool. Since all active components are testable and full flow can be demonstrated by recirculation through the -

test line, coolant injection into the reactor vessel is not required during the tests. !

SR 3.5.1.6 f The ADS designated S/RVs are required to actuate automatically upon receipt of specific initiation signals. A system i functional test (logic only) is performed to demonstrate that j the ADS logic operates as designed when initiated, causing !

proper actuation of the required components. Actual ADS valve actuation is excluded to prevent a reactor pressure vessel <

blowdown. <

1R 3.5.LJ A manual af.tuation of eact. ADS valve is performed to verify the valve and solenoids are functioning properly and no :

blockage exists in the S/RV discharge lines. This is demonstrated by the response of the turbine control or bypass valve or by a change in the measured steam flow or any other '

method suitable to verify steam flow. Adequate reactor steam dome pressure must be available to perform this test to avoid i damaging the valve. - Sufficient time 'is therefore allowed, l after the required pressure is achieved, to perform this test.

Reactor startup is allowed prior to performing this test .

because valve OPERABILITY and the set)oints for overpressure l protection are verified, per ASME requ' rements, prior to valve installation.

(continued)

~

l l Grand Gulf - Unit 1 B 3.5-8 DRAFT B 5/15/90

F 1 d

ECCS - Operating i B 3.5.1 '

BASES (continued) !

SURVEILLANCE Surveillance Frecuencies REQUIREMENTS (continued) In general, surveillance frequencies are based on industry accepted practice and engineering judgement considering the unit conditions required to perform the test, the ease of ,

performing the test and a likelihood of a change in the system / component status. ,

REFERENCES 1. Grand Gulf FSAR, Section 6.3.2.2.3.

2. Grand Gulf FSAR, Section 6.3.2.2.4.

3. Grand Gulf FSAR, Section 6.3.2.2.1.

1

4. Grand Gulf FSAR, Section 6.3.2.2.2. -

5. 10CFR50, Appendix K, "ECCS Evaluation Models".

6. Grand Gulf FSAR, Section 6.3.3.

7. 10CFR50.46, " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors".

8. Grand Gulf FSAR, Section 6.3.3.3. [

9. Grand Gulf FSAR, Section 6.3.3.7.8.

10. Grand Gulf FSAR, Section 7.3.1.1.).4.2.

11. NED0 31466, " Technical Specification Screening Criteria Application and Risk Assessment", November 1987. .

12. Memo, R. L. Baer (NRC) to V. Stello, Jr. (NRC), ;

" Recommended Interim Revisions to LCO's for ECCS .!

Components", December 1, 1975. !

)

Grand' Gulf - Unit 1 B 3.5 9 DRAFT B 5/15/90 em t-

V Grand Culf Nuclear Statien )

Technical Specification Itprove ent Prograa !

4 Revision Summary Sheet l l

Proposed LCO/Section: 3. 5.2 . Rev. 1_ ECCS-Shutdown ;

, 11gg Chance Descrintion Cateaorv

]

1 LCO 3.5.2 is reformatted from LIMITING CONDITION 1 I FOR OPERATION 3.5.2.

2 The details of subsystem operability requireraents 2 are relocated to the BASES.

3 The applicability in MODE 5 is revised to eliminate 38 ,

the requirement to remove the transfer canal gate and :

to specify the water level requirement. .

4 CONDITION A is reformatted from ACTION a. 1 i

5 CONDITION B is reformatted from ACTION a except as 1 :

discussed below, t

6 CONDITION C is reformatted from ACTION b except as 1 !

discussed below.

7 CONDITION D is reformatted from ACTION b except as I h discussed below.

8 CONDITION D requires secondary containment to be 3A !

made OPERABLE as soon as practicable rather than within i 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months per ACTION b. '

9 SR 3.5.2.1 is developed from LCO 3.5.3, item b, 3B and.SR 4.5.3.1.b.

10 SR 3.5.2.2 is developed from SR 4.5.2.2, LCO 3.5.3 3B item b, and SR 4.5.3.1.

11 SR 3.5.2.3 is reformatted from SR 4.5.2.1 and 1 SR 4.5.1.a.1 except as discussed below.

12 SR 3.5.2.4 is reformatted from SR 4.5.2.1 and 1 SR 4.5.1.a.3. ,

13 A NOTE to SR 3.5.2.4 is added to allow a LPCI 38+

subsystem in SDC mode to be considered OpEP.ABLE for ,

ECCS if capable of being manually aligned. l 14 SR 3.5.2.5 is reformatted from SR 4.5.2.1 and 4.5.1.b. I 1 15 SR 3.5.2.6 is reformatted from SR 4.5.2.1 and 1 4.5.1.c.1. ,

16 CROSS REFERENCES are added. 1 l-i

_______6_-- --t _ m____a __-_m_ _ _ -__ _ _ _ - _ _ _ _ _ _ - _ - - _

I Grand Gulf Nuclear Statten l Technical Specificatien I provement Pr:gran Revision Summary Sheet l

J Proposed LCO/Section: 3.5.2 Rev. 1. ECCS-Shutdown 1 ISg Chance Descrintion Cateaorv 17 The 3.0.4 sentence in ACTION a, along with footnote 1 !

'#' to page 3/4 5-6 are deleted. Applicable only until '

startup from RF03.

18 CONDITIONS B, C and 0 permit suspension of operations 3B with a potential for draining the reactor vessel as soon .

as practicable versus within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and immediately of l CTS ACTIONS a and b respectively. j 19 Suspension of CORE ALTERATIONS is deleted from ACTION b. 3B 20 The REQUIRED ACTIONS D.2, D.3 and D.4 are not 3B equivalent to the CTS definition of SECONDARY CONTAINMENT IllTEGRITY. <

21 The method of testing in SR 3.5.2.3 is relocated 38 !

(venting at the high point vent).

I 22 A note is added to SR 3.5.2.2 to indicate that the SR 1 is only required when HPCS system is required to be ,

OPERABLE.

b i.

i r

i

ECCS o Shutdown '

3.5.2 3.5 EC*R AND RCIC t 3.5.2 fees - shutdown LCO 3.5.2 Two ECCS injection / spray systems shall be OPERABLE. .

APPLICABILITY: MODE 4, MODE 5 except with the upper containment cavity to dryer pool gate removed and water level 2 22'8" over the top of the ,

RPV flange.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME

==

A. One of the required A.1 Restore the required 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from ECCS systems inoperable. ECCS systems to discovery of OPERABLE status, inoperable system B .. Required Action and B.1 Suspend operations As soon as associated Completion with a potential for practicable Time of Condition A dre.ining the reactor not met, vessel.

C. Both of the required C.1 Suspend operations As soon as ECCS systems with a potential for practicable inoperable, draining the reactor vessel.

AND >

C.2 Restore at least one 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ECCS system to OPERABLE status.

(continued)

I Grand- Gulf - Unit 1 3.5-6 DRAFT B 5/15/90

ECCS - Shutdown 3.5.2 ACTIONS fcontinued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action C.2 0.1 Sussend operations and associated wit 1 a potential for

[

Completion Time not draining the reactor met, vessel.

1 l

MD F ,

D.2 Ensure Secondary &

Containment is OPERABLE. -

MQ D.3 Ensure the SGTS is in As soon as compliance with the practicable requirements of Specification 3.6.4.3. ;

MD D.4 Ensure Secondary I Containment Isolation ,

Valves are in .

compliance with the requirements of Specification 3.6.4.2 and Secondary Containment Actuation Instrumentation is in compliance with the requirements of Specification 3.3.6.2.

I l

1 j

i Grand Gulf - Unit 1 3.5-7 DRAFT B 5/15/90 l I

ECCS - Shutdocn 3.5.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY t

SR 3.5.2.1 Verify the suppression pool water level 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is 12.67' when a low pressure ECCS system is required to be OPERABLE.

SR 3.5.2.2 --------------

NOTE------------ -----

Only required when HPCS system is required to be OPERABLE.

Verify for the HPCS system the: 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months A. Suppression pool water level is 2 12.67',

QB B. CST water level is 118'.

SR 3.5.2.3 Demonstrate for each required ECCS 31 days injection / spray system the system piping is filled with water from the pump discharge valve to the system isolation valve.

SR 3.5.2.4 -------------------NOTE------------------

LPCI subsystems may be considered OPERABLE during alignment to and operation in the RHR shutdown cooling mode if capable of being manually realigned and not otherwise inoperable.

Verify for each required ECCS 31 days injection / spray system each manual, power operated or automatic valve in the flow path not locked, sealed or otherwise secured in position is in its correct position.

(continued)

Grand Gulf - Unit 1 3.5-8 DRAFT B 5/15/90 1

F t i

ECCS - Shutdown ;

3.5.2 .

! t SURVEILLANCE REOUIREMENTS (continuedi ,

i SURVEILLANCE FREQUENCY SR ' 3.5.2.5 Demonstrate each required ECCS pump According to develops the specified flow rates with SR 3.0.5 the specified total developed head. '

NO. TOTAL OF DEVELOPED SYSTEM FLOW RATE PUMPS HEAD LPCS 2 7115 gpm 1 1 290 psid LPCI 2 7450 gpm 3 1 125 psid HPCS 2 7115 gpm 1 1 445 psid SR 3.5.2.6 ----- ------ -------NOTE--- ----- --------

Vessel injection may be excluded.

Perform a system functional test for each 18 months required ECCS injection / spray system, including simulated automatic actuation of' the system throughout its emergency operating sequence, to verify each ,

automatic valve in the flow path actuates to its correct position.

t b

F i

l Grand Gulf - Unit 1 3.5 9 DRAFT B 5/15/90

i j

ECCS - Shutdown 3.5.2 i CROSS-REFERENCES TITLE NUMBER ,

4 ECCS Actuation Instrumentation 3.3.5.1 l Secondary Containment Actuation Instrumentation 3.3.6.2 Residual' Heat Removal - Shutdown 3.4.7 )

Secondary Containment 3.6.4.1 1

Secondary Containment Isolation Valves 3.6.4.2 Standby Gas Treatment System 3.6.4.3 Residual Heat Removal - High Water Level 3.9.8 ;

Residual Heat Removal - Low Water Level 3.9.9

(

4 l

l?

1 I

Grand Gulf - Unit 1 3.5-10 DRAFT B 5/15/90 l~

l

I ECCS - Shutdown B 3.5.2 -

B 3.5 ECCS AND RCIC B 3.5.2 ECCS - Shutdown BASES BACKGROUND A description of the High Pressure Core Spray (HPCS), Low Pressure Core Spray (LPCS) system and the Low Pressure Coolant Injection (LPCI) subsystems of the Residual Heat Removal (RHR) .

system is provided in the Bases for LCO 3.5.1.

APPLICABLE For MODES 1, 2 and 3 ECCS performance is evaluated for the SAFETY entire spectrum of break sizes for a postulated Loss of Coolant ANALYSIS Accident (LOCA). The long-term cooling analysis following a design basis LOCA (Ref.1) demonstrates that only one ECCS ,

system is. required, post-LOCA, to maintain the peak cladding ,

temperature below the allowable limit. In MODES 4 and 5, two OPERABLE ECCS systems ensure adequate vessel inventory makeup i in the event of an inadvertent vessel draindown, i

ECCS-Shutdown satisfies the requirements of Selection Criterion of the NRC Interim Policy Statement on Technical 3

Specification Improvements as documented in Reference 3.

LCO Two ECCS injection / spray systems are required to be OPERABLE.

The ECCS injection / spray systems are defined as the three LPCI l subsystems, the LPCS system and the HPCS system. The LPCS systeni and each LPCI subsystem consists of one motor driven pump, piping and valves to transfer water from the suppression pool to the reactor vessel. The HPCS system consists of one motor driven pump, piping and valves to transfer water from the suppression pool or Condensate Storage Tank (CST).to the j reactor vessel. Any LPCI subsystem (A or B) that may be '

aligned in the shutdown cooling mode of the RHR system in MODE 4 or 5 is considered OPERABLE for the ECCS function, if it can be manually realigned (remote or manual) to the LPCI mode and j is not otherwise inoperable. Because of low pressure and /

temperature conditions in MODES 4 and 5, sufficient time will )

be available to manually align and initiate LPCI subsystem ;

operation to provide core cooling prior to postulated fuel l uncovery.

I (continued)

L 1

Grand Gulf - Unit 1 B 3.5-10 DRAFT B 5/15/90

i

~

ECCS - Shutdown B 3.5.2 BASES (continued)

APPLICABILITY ECCS OPERABILITY is required in MODES 4 and 5 to assure i

-adequate coolant inventory and sufficient heat removal capability for the irradiated fuel in the core in case of an inadvertent draindown of the vessel. Requirements for ECCS OPERABILITY during MODES 1, 2. and 3 are discussed in the Applicability section of the Bases for LCO 3.5.1.1:CCS systems -

are not required to be OPERABLE during MODE 5 with the upper

containment cavity to dryer pool gate removed and the water '

level maintained greater than or equal to 22'8" feet above the Reactor Pressure Vessel (RPV) flange. This provides sufficient coolant inventory to allow operator action to terminate the inventory loss prior to fuel uncovery in case of an inadvertent-draindown. ,

The Automatic Depressurization System (ADS) is not reqdired to be OPERABLE during MODES 4 and 5 because the reactor vessel pressure is < 135 psig and the LPCS system, HPCS system and '

LPCI subsystems can provide core cooling without any

depressurization of the primary system being required.

ACTIONS A.I. B.1 d-With one of the two required ECCS systems inoperable, the

. remaining OPERABLE system can provide sufficient vessel flooding capability to recover from an inadvertent vessel draindown. . However, the overall system reliability is reduced 1-because a single failure in the remaining system concurrent with a vessel draindown could result in the ECCS not being able to perform its intended function. Therefore, continued operation is only allowed for a limited time. With the L inoperable system not restored to OPERABLE status within the l required Completion Time, operations that have the potential

for draining the ' reactor vessel must' be suspended. This L minimizes the probability of a vessel draindown and the

L subsequent potential for ECCS actuation.

L (continued) l l^-

l~ ' Grand Gulf - Unit 1 B 3.5-11 DRAFT B 5/15/90 L

E !

ECCS - Shutdswn B-3.5.2 j t

BASES (continued)

ACTIONS C.I. C.2. D.1. 0.2. D.3. 0.4 (continued) -

l With both of the required ECCS systems inoperable, all coolant ,

inventory makeup capability may be unavailable and operations

that have a potential for draining the reactor vessel must be :

suspended. If a least one- ECCS system is not restored to '

OPERABLE status within the required Completion Times, i additional actions are required to minimize any potential- !

release of radioactive materials to the environment. This includes ensuring Secondary Containment is OPERABLE (LC0 3.6.4.1), the Standby Gas Treatment System (SGTS) is in ;

compliance with its Specification (LC0 3.6.4.3) and the i Secondary Containment Isolation Valves and Secondary :

Containment Actuation Instrumentation are in compliance with their_ Specifications (LCO 3.6.4.2 and 3.3.6.2 respectively).

This may be performed by an administrative check, by examining i logs or other information, to determine if the components are ,

out of service for maintenance or other reasons. It does not mean to perform the surveillances needed to demonstrate .

OPERABILITY of the components. If however, any - required component is inoperable, it must be restored to OPERABLE -

status. In this case, surveillance requirements may need to :

be performed to restore the component to OPERABLE status.

Comoletion Times !

~

All Completion Times are based on industry accepted practice and engineering judgement considering the number of available systems and the time required to reasonably complete the :

. Required Action. l SURVEILLANCE SR 3.5.2.1. SR 3.5.2.2 -

l- REQUIREMENTS -

l The minimum water level 12.67' required for,the suppression L pool is verified to ensure the suppression pool will provide -

adequatenetpositivesuctionhead(NPSH)fortheECCSpumps, recirculation volume, vortex prevention and a safety margin for conservatism. With the suppression pool water level less than the required limit, all ECCS systems are inoperable unless they are aligned to an OPERABLE CST.

(continued)

Grand Gulf - Unit 1 B 3.5-12 DRAFT B 5/15/90 6

- . . - # , ,,. , _ - , ,,- ~-

L ECCS - Shutdown i B 3.5.2 l BASES (continued)

SURVEILLANCE SR 3.5.2_.1. SR 3.5.2.2 (continued)

REQUIREMENTS ;

(continued) When suppression pool level is less than 12.67', HPCS is !

. considered OPERABLE only if it can take-suction from the CST and the CST water level is sufficient to provide the required ;

NPSH for the HPCS pump. Therefore, a verification that either i the suppression pool water level is 112.67' or HPCS is aligned to take suction from the CST and the CST contains 1 170,000 l gallons of water, equivalent to 18', ensures HPCS can supply l makeup water to the reactor vessel. I 1

SR 3.5.2.3. SR 3.5.2.5. SR 3.5.2.6 The bases provided for SR 3.5.1.1, SR 3.5.1.4 and SR 3.5.1.5 are applicable.

1 SR 3.5.2.4 Verification that all applicable valves are in the required position ensures proper flow paths for ECCS. However, a valve that is capable of automatic return to its ECCS position, when and ECCS signal is present, can be in position for another mode of operation. I i

In MODES 4 and 5, the RHR system may operate in the shutdown !

cooling mode to remove decay heat and sensible heat from the reactor. Therefore, during MODE 4 and 5, RHR valves that are i required for LPCI subsystem operation may be aligned for the ;

shutdown cooling mode. The LPCI mode of the RHR however, may be considered OPERABLE for the ECCS function if all the required valves in the LPCI flow path can be manually realigned to allow injection into the RPV and the system is not otherwise inoperable. This will ensure adequate core cooling if an -

inadvertent vessel draindown should occur.

Surveillance Freauencies In general, surveillance frequencies - are based on industry-accepted practice and engineering judgement considering the-unit conditions required to perform the test, the ease of performing the test and a likelihood of a change in the systeWcomponent status.

(continued) b G

Grand Gulf' - Unit 1 B 3.5-13 DRAFT B 5/15/90 i

~3

r e

ECCS -' Shutdown B 3.5.2 BASES (continued)

REFERENCES 1. Grand Gulf FSAR, Section 6.3.3.4.

2. 10CFR50.46, " Acceptance Criteria for Emergency Core Cooling Systems for Light-Water-Cooled Nuclear Power Reactors".

3. NEDO-31466, " Technical Specification Screening Criteria Application and Risk Assessment", November 1987.

i i

Grand Gulf - Unit 1 B 3.5-14 DRAFT B 5/15/90 j

e

[ Grand Gulf Nuclear Station .

Technical Sp;cification Irprovement Program j o i Revision Summary Sheet !

i Proposed LCO/Section: 3.5.3 Rev. 1 fLCIC 1115 Chance Descrintion Cateoorv l l

1 LCO 3.5.3 and applicability are reformatted 1 from LIMITING CONDITION FOR OPERATION 3.7.3 and its t applicability. .

2 The details of system operability requirements 2 are relocated to the BASES.

3 DELETED 4 CONDITIONS A and B are reformatted from the ACTION 1 statement.

5 SR 3.5.3.1 is reformatted from SR 4.7.3.a.1. 1 6 The method of verifying that system piping is 2 filled in SR 3.5.3.1 is relocated.

7 SR 3.5.3.2 is reformatted from SR 4.7.3.a.2. 1 8 SR 3.5.3.3 is reformatted from SR 4.7.3.b and 3B footnote '*' to page 3/4 7-7. The pressure conditions are restated.

9 SR 3.5.3.4 is reformatted from SR 4.7.3.c.2 and 3B footnote '*' to page 3/4 7-8. The pressure conditions are restated.

10 SR 3.5.3.5 is reformatted from SR 4.7.3.c.1 1 except as discussed below.

11 CROSS REFERENCES are added. 1r

'12 SR 4.7.3.a.3 is deleted. This is considered 2 in the system lineup of SR 3.5.3.2 and is !

explicitly described in the BASES for SR 3.5.3.2.

13 SR 4.7.3.c.3 is deleted. This is considered in 2 the testing of SR 3.5.3.5 and is explicitly described in the BASES for SR 3.5.3.5.

14 The restart requirement from SR 4.7.3.c.1 is deleted. 3B 15 The words "throughout its emergency operating 3A sequence" are added in SR 3.5.3.5.

RCIC 3.5.3 3.5 ECCS AND RCIC 3.5.3 Reactor core Isolation coolina system LCO 3.5.3 The Reactor Core Isolation Cooling (RCIC) system shall be OPERABLE.

APPLICABILITY: H0DE 1.

H0 DES 2 and 3 with reactor steam dome pressure

> 135 psig.

i ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. RCIC inoperable. A.1 Verify HPCS is OPERABLE. Immediately

&MQ A.2 Restore the system to 14 days from OPERABLE status. discovery of '

inoperable system <

B. Required Actions and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion ;

Times of Condition A AND not met.

B.2 Reduce reactor steam 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months dome pressure to

$ 135 psig.

l t

f Grand Gulf - Unit 1- 3.5-11 DRAFT B 2/13/90 l

I RCIC 3.5.3 i

SURVEILLANCE RE0VIREMENTS l

SURVEILLANCE FREQUENCY SR 3.5.3.1 Demonstrate RCIC system piping is filled 31 days with water from the pump discharge valve to <

the system isolation valve. '

SR 3. 5.3.2 - Verify each manual, ower operated or 31 days

~

automatic valve in t e flow path not i locked, sealed or otherwise secured in position is in its correct position. !

1 I

SR 3.5.3.3 Demonstrate, with reactor pressure -----NOTE-----

$ 1045 psig, the RCIC pump can develop Only required a flow rate 1800 gpm against a system within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months head corresponding to a reactor pressure when reactor 1 945 psig, steam dome pressure is i 1 945 psig 92 days SR 3.5.3.4 Demonstrate, with reactor pressure -----NOTE-----

l $ 165 psig, the RCIC pump can develop Only required a flow rate 1800 gpm against a within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months system head corresponding to a reactor when reactor

pressure 1 150 psig, steam dome l pressure is 1 135 psig 18 months (continued) i Grand Gulf - Unit 1 3.5-12 DRAFT B 2/13/90 i

RCIC j 3.5.3 SURVEILLANCE REOUIREMENTS fcontinued)

SURVEILLANCE FREQUENCY SR 3.5.3.5 -----------------NOTE-----------------

Vessel injection may be excluded.

Perform a system functional test for 18 months the RCIC system, including simulated automatic actuation throughout its emergency operating sequence, to verify each automatic valve in the flow path actuates to its correct position.

CROSS-REFERENCES TITLE NUMBER Reactor Core Isolation Cooling Actuation Instrumentation 3.3.5.2 ECCS - Operating 3.5.1

' Grand Gulf - Unit 1 3.5 13 DRAFT B 2/13/90

7 RCIC B 3.5.3 c

B 3.5 ECCS AND RCIC B 3.5.3 Reactor Core Isolation Coolina system BASES BACKGROUND NOTE: The Reactor Core Isolation Cooling (RCIC) system is not andEmergencyCoreCoolingSystem(ECCS). The RCIC system is included with the ECCS section because of similar functions during certain plant transients.

The RCIC system is designed to operate following Reactor Pressure Vessel (RPV) isolation accompanied by a loss of coolant flow from the feedwater system to provide adequate core cooling and control of reactor vessel water level. Under these conditions, the High Pressure Core Spray (HPCS) and RCIC systems perform similar functions.

The RCIC system (Ref.1) consists of a steam driven-turbine-pump unit, piping and valves to provide steam to the turbine, and piping and valves to transfer water from the suction source to the core via the feedwater system line. Suction piping is provided from the Condensate Storage Tank (CST) and the suppression pool. Pump suction is normally aligned to the CST to minimize injection of suppression pool water into the RPV.

However, if the CST water supply is low or the suppression pool level high, an automatic transfer to the suppression pool assures a continuous water supply for the RCIC system. The steam supply to the turbine is piped from the main steam line A upstream of the inboard main steam line isolation valve.

The RCIC system is designed to provide core cooling over a wide ;

range of reactor pressures, 150 to 1177 psig. Upon receipt !

of an initiation signal, the RCIC turbine accelerates at a specified rate. As the RCIC flow increases, the turbine ;

control valve is automatically adjusted to maintain design - s

! flow. Exhaust steam from the ROIC turbine is discharged to

! the suppression pool. A full flow line is provided to route t water from and to the CST to allow testing of the RCIC system during ~ normal operation without injecting water into the L reactor vessel. '

l l- The RCIC pump is provided with a minimum flow bypass line which discharges to the suppression pool. The valve in this line automatically opens to prevent pump damage due to overheating at reduced RCIC pump discharge flow. Low flow combined with high pump discharge pressure opens the valve.

l To ensure rapid delivery of water to the reactor vessel and to minimize waterhammer effects, the RCIC system discharge line is maintained filled by the static head of water from the Condensate Storage Tank.

(continued)

Grand Gulf - Unit 1 B 3.5-15 DRAFT B 2/13/90 ,

_ _ - _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ . _w . , . - . . _ m ,,,-w__,-w,- , , - - - , . , , - , ..m.

RCIC !

L B 3.5.3 l BASES (continued)

APPLICABLE The ability of the RCIC system to provide makeup coolant to SAFETY the reactor is used to respond to transient events. However,

' ANALYSES the RCIC system is not an Engineered Safety Feature -(ESF) 1 system and no credit is taken in the safety analyses for RCIC j system operation. However, based on its contribution to the 1 F

reduction of overall plant risk, Reference 2 requires that the i RCIC system be included in the technical specifications even I though none of the Selection Criteria were satisfied (Ref. 3). .J LCO RCIC is required to be OPERABLE to provide makeup coolant to l the reactor in the event of reactor isolation accompanied by f a loss of feedwater flow. A description of what is required ,

for the RCIC system to be considered OPERABLE is provided in ;

the Background section. '

L

' APPLICABILITY The RCIC system is required to be OPERABLE in MODES 1, 2 and 3 with reactor steam dome pressure > 135 psig since RCIC'is the primary non ECCS water source for core cooling when the -

reactor is isolated and pressurized. In MODES 2 and 3 with l reactor steam dome pressure 1135 psig, and in MODES 4 and 5, RCIC is not required to be OPERABLE since the ECCS subsystems can provide sufficient flow to the vessel.

ACTIONS A.I. A.2 During MODES 1, 2 or 3 with reactor steam dome pressure > 135 psig, loss of RCIC will not affect the overall plant capability to provide makeup coolant during transients at high reactor pressure since either HPCS or RCIC is assumed to be available ,

during plant transient analyses. OPERABILITY of HPCS is :

therefore verified when the RCIC system is inoperable. This l may be performed by an administrative check, by examining logs or other information, to-determine if the HPCS system is out of service for maintenance or other reasons. It does not mean to perform the surveillance requirements needed to demonstrate -

the OPERABILITY of the HPCS system. For transients and certain abnormal events with no LOCA, RCIC (as opposed to HPCS) is the >

preferred source of makeup coolant because of its relatively small capacity which allows easier control of reactor vessel water level. Therefore, continued operation is only permitted for a limited time.

(continued)

Grand Gulf - Unit 1 B 3.5-16 DRAFT B 2/13/90

RCIC B 3.5.3 ,

BASES (continued)

ACTIONS B.1. B.2 (continued)

Should the Required Actions and associated Completion Times of Condition A not be met, the reactor is required to be in MODE 3 and the reactor pressure reduced to s 135 psig. At these conditions, RCIC is no longer required. a Comoletion Times +

The ECCS Completion times are based on the results of a study !

which evaluated the impact on ECCS unavailability assuming '

various components and subsystems were taken out of service. 1 The results were used to calculate the average unavailability '

of ECCS equipment needed to mitigate the consequences of a LOCA as a function of allowable outage time (A0T). A0Ts were then chosen to provide comparable levels of ECCS availability throughout the industry (Ref. 4). Because of the similar ;

functions of HPCS and RCIC, the A0Ts determined for HPCS are also applied to RCIC.

SURVEILLANCE SR 3.5.3.1 REQUIREMENTS The pump discharge line of the RCIC system is required to be kept full with water to minimize potential waterhammer effects when the system is initiated. Additionally, the lag between the receipt of the initiation signal and the actual injection into the reactor vessel is minimized. One acceptable method of ensuring the lines are " full" is to vent at the high points.

SR 3.5.3.2 Verification that all applicable valves are in the required position ensures proper flow paths for RCIC. For the RCIC system, this also includes the steam flow path for the turbine L and the flow controller position.

SR 3.5.3.3. SR 3.5.3.4 l The RCIC pump flow rates ensure that the system can maintain ,

reactor coolant inventory during pressurized conditions with L the RPV isolated and the reactor slutdown. The flow tests for .

L the RCIC system are performed at two different pressure ranges such that system capability to provide rated flow is tested both at the higher and lower operating ranges of the system.

Since the required reactor steam dome pressure must be available to perform SR 3.5.3.3 and SR 3.5.3.4, sufficient time :

is allowed after adequate pressure is achieved to perform these t tests.

(continued)

Grand Gulf - Unit 1 B 3.5-17 DRAFT B 2/13/90 j

RClc B 3.5.3 BASES feontinued)

SURVEILLANCE SR 3.5.3.5 REQUIREMENTS (continued) The RCIC system is required to actuate automatically to perform its designed function. This surveillance test demonstrates with the required system initiation signals, the automatic initiation logic of RCIC will cause the system to operate as designed, including actuation of all autcr.atic valves to their required positions. The test also ensures the RCIC system will automatically restart on a reactor vessel low water level (Level 2 signal received subsequent to a reactor vessel high water le)elv (Level 8) trip and that suction is automatically transferred from the CST to the suppression pool. Since all~

active components are testable and full flow can be demonstrated by recirculation through the test line, coolant injection into the reactor vessel is not required during the test.

Igrve111ance Frecuencies In general, surveillance frequencies are based on industry accepted practice and engineering judgement considering the unit conditions required to perform the test, the ease of performing the test an a likelihood of a change in the system / component status.

REFERENCES 1. Grand Gulf FSAR, Section 5.4.6.2.

2. 52FR3788, Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, February 6, 1987.

3. NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment", November 1987.

4. Memo, R. L. Baer (NRC) tu V. Stello, Jr. (NRC),

" Recommended Interim. Revisions to LCO's for ECCS Components," December 1, 1975.

Grand Gulf - Unit 1 B 3.5-18 DRAFT B 2/13/90 4

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l 1

l 81 . y

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?

CHAPTER 3.7 PLANT SYSTEMS i

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5 l'v. t 1:

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+.

l, CHAPTER 3.7 :

PLANT SYSTEMS !

TABLE OF CONTENTS 3.7.1 Standby Service Water System - Operating 3.7.2 : Standby Service Water System - Shutdown 3.7.3 HPCS Service Water System i 3.7.4 . Control Room Fresh Air System 3.7.5 Main Condenser Offgas r

i

,s ,-.

Grand Gulf Nuclear. Station Technical Specification Improvement Pregram l Revision Summary Sheet j Proposed LCO/Section: 3.7.1 Rev.__L SSW-Oceratina 1tg Chance.Qtserietion Cateaorv i

I LCO 3.7.1 is reformatted from LIMITING CONDITIONS 1

, FOR OPERATION 3.7.1.1 and 3.7.1.3.

2 Details of system operability requirements are 2 relocated to the Bases.

3 CONDITIONS A and B are reformatted from LCO 3.7.1.1 1 ACTION a except as discussed below.

4 REQUIRED ACTION A.1 is added to permit an individual 3B+

system receiving water from SSW to be declared '

inoperable rather than declaring S$W in it entirety inoperable based upon the River Bend Te'ch Specs, t 5 ' LCO 3.7.1.1 ACTION b for MODE 3 is deleted. 4 6 LCO 3.7.1.1 Item b.2, ACTIONS b and c, footnotes '*' 1 and '#' to page 3/4 7-1 ACTIONS d, e, f (modes 4 and S) and footnote '#' to page 3/4 7-2 are moved to-LCO 3.7.2.

7 LCO 3.7.1.1 ACTION f for MODES 1,2 and 3 is deleted. 4 8 SR 3.7.1.1 is reformatted from LCO 3.7.1.3 item a 1 ,

and SR 4.7.1.3.a.

9 ,

DELETED 10 SR 3.7.1.2 is reformatted from SR 4.7.1.1.a.- 1 L

t 11 SR 3.7.1.3 is reformatted from SR 4.7.1.3.b. 1 12 The provision to operate the fan from the control 3B room in SR 4.7.1.3.b is deleted. The intent is to test the fan, not the control room switch.

13 SR 3.7.1.4 is reformatted from SR 4.7.1.1.b and 1-SR 4.7.1.3.c.

14 CROSS REFERENCES are added. I ;

15 SR 3.7.1.2 has been limited to valves in lines 3B servicing only safety related systems or components.

l 16 Technical Specification Position Statement 101 is not 4 l- j incorporated into LCO 3.7.1. ,

SSW System - Operating I 3.7.1 J 3.7 PLANT SYSTEMS 3.7.1 Standby service Water system - Doeratino LCO 3.7.1 The Division 1 and 2 Standby Service Water (SSW) subsystems shall be OPERABLE. ;

APPLICABILITY: MODES 1, 2, and 3.

l ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 1 A. One of the required A.1 Declare affected system Immediately SSW subsystems or component inoperable, upon inoperable, discovery of inoperable component E

A.2 Restore the inoperable 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from i SSW subsystem to discovery of 1-OPERABLE status. inoperable component B. Required Actions and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Times of Condition A MQ not met.

B.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months M

Both SSW subsystems inoperable.

-Grand Gulf - Unit 1 3.7-1 DRAFT B 2/13/90

-i

SSW System - Operating 3.7.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR- 3.7.1.1 Verifytheultipateheatsinkbasinwater 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months level is 2 7.25.he+. I

< SR' 3.7.1.2 Verify for each required SSW subsystem 31 days each manual, power operated or automatic valve in SSW flow paths servicing safety related systems or components not locked, sealed or otherwise secured in position is in its correct position.

SR 3.7.1.3 Demonstate cach cooling tower fan operates 31 days for 1 15 minutes.

SR 3.7.1.4 Perform a system functional test for 18 months !

each requ. ired SSW subsystem including '-

simulated automatic actuation of the subsystem.

l l

CROSS-REFERENCES

TITLE NUMBER Residual Heat Removal - Shutdown 3.4.7 l

ECCS - Operating 3.5.1 Reactor Core Isolation Cooling System 3.5.3 l Residual Heat Removal Containment Spray 3.6.1.9 Residual Heat Removal Suppression Pool Cooling 3.6.2.3 !

A.C. Sources - Operating 3.8.1 l t i

i Grand Gulf - Unit 1 3.7-2 DRAFT B 2/13/90 I

SSW System - Operating l B 3,7.1 8 3.7 PLANT SYSTEMS i B 3.7.1 Standbv service Water system - Doeratine !

BASES l

BACKGROUND TheStandbyServiceWater(SSW)Systemisdesignedtoprovide !

cooling water for the removal of heat from slant auxiliaries, '

such as Residual Heat Removal (RHR) system seat exchangers, standby diesel generators, and room coolers for Emergency Core !

Cooling System (ECCS) equipment, required for a safe reactor ;

shutdown following a design basis transient or accident. The. j

, SSW system also provides cooling to plant components, as i required, during normal shutdown and reactor isolation modes. i During a design basis accident, the equipment required for ,

normal operation only, is isolated from the SSW system and i cooling is directed only to safety related equipment.

For the purposo cf this technical specification the SSW system ;

consists of the ultimate heat sink (VHS), two independent 3 cooling water headers (subsystems A and B) and their associated '

pumps, piping, valves and instrumentation. Subsystems A and B- i are redundant and service equipment in Division I and II, j respectively, j J The UHS is tw'o concrete makeup water basins each comprised of ;

one cooling tower with four independent fan cells (two fan cells per basin). The combined basin volume is sized such that '

sufficient water inventory is available for all SSW system post-LOCA cooling requirements for a 30 day period with no ,

external makeup water source available (Ref.1). Normal makeup for each basin is provided-automatically by the Plant Service Water (PSW) System.

Cooling water is pumped from the cooling tower basins by the ,

two SSW pumps to the essential components through the two main :

redundant supply headers (Subsystems A and B). After removing- 1 heat from the components, the water is discharged-to the :

cooling towers where the heat is rejected through direct ;

contact with ambient air.

- (continued) ,

I t

t l

Grand Gulf - Unit 1 B 3.7-1 DRAFT B 2/13/90 1

SSW System J 0perating B 3.7.1

' BASES'Icontinued)

BACKGROUND Subsystems A and B supply cooling water to redundant equipment

=(continued) required for a safe reactor shutdown. The specific equipment for which the SSW system supplies cooling water is listed in Reference 2. Subsystem A and B pumps are sized such that the o)eration of both of them or one'of them in conjunction with tie HPCS Service Water System (LCO 3.7.3) pump will provide adequate cooling water to the equipment required for safe shutdown. The SSW system is designed to withstand a single active or passive failure coincident with a loss of offsite power without losing the capability to supply adequate cooling

' water to equipment required for safe reactor-shutdown (Ref.:3).

Following a design basis accident or transient, t.he SSW system

-will operate automatically and without operator action. Manual initiation of supported systems, e.g. suppression pool cooling, is however, performed for long term cooling operations.

APPLICABLE - The ability of the SSW system to support long term cooling of

" ' SAFETY the reactor or containment is evaluated in FSAR Chapters 6 ANALYSTS. (Engineered Safety Features),'9 (Auxiliary Systems) and 15 (Accident Analyses). These analyses explicitly assume the SSW will provide adequate cooling support to the equipment required,.

~'

for safe reactor shutdown. These analyses include the -

t evaluation of the long term containment response after a design basis-Loss Of Coolant Accident (LOCA). The SSW system provides cooling water for the RHR suppression pool cooling mode to-limit the suppression pool temperature and containment pressure following a LOCA. This ensures the containment can perform its intended > function of limiting the release ~of radioactive materials to the environment following a LOCA. The SSW system also provides cooling to other components assumed to function during a LOCA.

The safety analyses for -long term cooling were performed (Ref.

4 and M for a LOCA concurrent with a loss of offsite power, and n.Himum available diesel generator power. The worst case single failure which would affect the performance of- the SSW subsystems A or B is the failure of one of the two standby diesel-generators which would affect one subsystem of the SSW system. The SSW flow assumed in the analyses is 7900 gpm per pump (Ref. 6).

SSW System - Operating satisfies the requirements of Selection Criterion 3 of the NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 6.

(continueo)

Grand Gulf - Unit 1 B 3.7-2 DRg r 5 2/13/90 l

e SSW System -_ Operating B 3.7.l' BASES (continued) -

LCO The OPERABILITY of Division I (subsystem A) and Division !!

(subsystem B) of the SSW system is required to ensure the effective operation of the RHR system in removing heat from the reactor and the effective operation of other safety related equipment during_a design basis accident or transient. The OPERABILITY of each independent subsystem of the.SSW system is based on having an OPERABLE UHS, the pump in the subsystem-OPERABLE and an OPERABLE flow path capable of taking suction from the associated SSW cooling basin and transferring-the water to the appropriate plant equipment, as required. Requiring both

, subsystems OPERABLE assures either subsystem A or B subsystems-A and B together will be available to provide adequate capability to meet cooling requirements of the equipment.

required for safe shutdown.

The OPEk/,BILITY of the-UHS is based on having a minimum basin water level at or above elevation 130' 3" mean sea level

, (which is equivalent to an indicated level of > 7.25 feet),

and having-four OPERABLE cooling tower fans.

APPLICABILITY The requirements for OPERABILITY of. the SSW system including the cooling tower basins in MODES 1,' 2,. and 3 are governed- by l'-

the required OPERABILITY of the equipment serviced by the SSW system in those MODES. SSW system requirements for other--

operating modes are covered in- LCO 3.7.2.

ACTIONS A'.1. A.2 With SSW subsystem A or B inoperable either due to an inoperable pump or inoperable flow path,. sufficient' cooling water can be supplied by the remaining OPERABLE subsystem should a reactor shutdown be:necessary. However, if an-additional. single failure in the SSW system were'to occur, the system would not be capable of performing its intended-

. function. Therefore, only'a limited time is allowed' to restore the inoperable subsystem to OPERABLE status. Alternatively, when the .SSW flow path to any safety related system or.

component is inoperable, the affected system may be declared-inoperable and the applicable LC0 is entered to determine the appropriate action. The SSW subsystem may still be capable of providing cooling water to all other associated systems.

(continued)

Grand Gulf - Unit 1 B 3.7-3 DRAFT B 2/13/90 I

1

s SSW-System - Operating

- + 8 3.7.1

.' BASES (continued)

' ACTIONS B.1. B.2

, (continued)

If the Required Actions and associated Completion Times of n , Condition A cannot-be met, the reactor is required to be in MODE 3 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in MODE 4 in the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . In-MODE 4 the system requirements are reduced.as specified in LC0 3.7.2.

Additionally, with both subsystems of the SSW system' inoperable, the associated ecuipment cannot perform the intended function. Continuec operation _in these MODES cannot'

' be justified. .Therefore, the reactor is required.to be in MODE

-3 and subsequently in MODE-4. If MODE 4 cannot be achieved because of the inoperable SSW subsystems, the _ reactor coolant temperature should be maintained as-low as practicable using an alternate decay heat removal. method.

Comoletion Times All-Completion Times are based on industry accepted practice and engineering judgement considering the number of available systems and the time required to reasonably complete the Required Action.

[ <

. SURVEILLANCE SR 3.7.1.1 REQUIREMENTS This surveillance verifies the cooling tower basins have sufficient: cooling water (as measured by basin water level) to satisfy the design basis.of 30 day cooling capability with= no external makeup source. With the ultimate' heat sink inoperable, the affected SSW subsystems must be declared

' inoperable.

SR 3.7.1.2 Verification of the correct alignment of-all applicable valves is essential to ensure the proper flow paths servicing safety related systems or components for the SSW subsystems.

(continued)

Grand Gulf - Unit 1 B 3.7-4 DRAFT B 2/13/90 i

M -' ' _ _ _ _

SSWl System---Operating ^

B 3.7;1.

. BASES (continued)

, SURVEILLANCE SR 3.7.1.3 REQUIREMENTS o

(continued): This surveillance verifies the OPERABILITY of the SSW cooling.

.- tower fans. The 15 minute duration for fan operation is sufficient-to monitor the steady state performance of the fans.

SR 3.7.1.4 This surveillance verifles the automatic. isolation valves of

-the SSW system will. automatically switch to the safety or 4

emergency position to provide cooling water exclusively to the:

safety related equipment during an accident event. - This surveillance also verifies the automatic start capability _of the SSW cooling tower fans. The fans are required to start-automatically whenever the- associated SSW subsystem is started.

Surveillance Freauenqigi In general, Surveillance Frequencies are based on industry.

accepted practice and engineering judgement-considering the s unit conditions required to perform the test,_the ease of-performing the test and a likelihood of a change in the system / component status.

l REFERENCES 1. Grand Gulf FSAR, Section 9.2.5.1.1.

2. Grand Gulf FSAR, Table 9.2-3.

3. Grand Gulf FSAR, Section 9.2.1.1.1.a. and 9.2.1.1.1.d.

4. Grand Gulf FSAR, Section 6.2.1.1.3.3.1.6.

5. Grand Gulf FSAR, Section 6.2:.2.3.

6. NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment", November 1987.

' i

. Grand Gulf - Unit 1 B 3.7-5 DRAFT B 2/13/90 ,

2

I Grand Gulf Nuclear Station !

e Technical Specification' Improvement Program-Revision Summary Sheet

. Proposed LC0/Section: 3.7.2 SSW-Shutdown Rev. _1_

1133 Chance Descriotion Cateaorv .

s. - 1 LCO 13.7.2 is reformatted from LIMITING CONDITIONS 1-FOR OPERATION 3.7.1.1 and 3.7.1.3.

2 Details of system operability. requirements are 2 .

relocated to the Bases, t

=

3 CONDITION A is reformatted from LCO 3.7.1.1 ACTIONS 1 i b,-c, d, e and f and LCO 3.7.1.3 ACTION c. -

4 4 SR 3.7.2.1 is reformatted from SR 4.7.1.3.6 and 1 LCO 3.7.1.3 item a.

5- DELETED j 6 SR 3.7.2.2 is reformatted from SR'4.7.1.1.a except 1 as discussed below.

7 SR 3.7.2.3 is reformatted from SR 4.7.1.3.b except I as discussed below.

8 The provision to start the-fan froris the control 38 coom in SR 4.7.1.3.b is deleted. Intent is to test-the fan not the control room handswitch.

9 SR 3.7.2.4 is reformatted from SR 4.7.1.1.b and 1 SR 4.7.1.3.c. !

10 CROSS REFERENCES are added. 1 J

11 The Specification 3.0.4 exceptions in'LCO 3.7.1.1 1 ACTIONS b, c, and d, the Specification 3.0.3 exception in ACTION e and the '#' footnotes to pages 3/4 7-1 and 3/4 7-2 are deleted as they are.no longer applicable.

12 Footnote '**' to page 3/4 7-1 is deleted.- The . 2 alternate MODE 4 provision is relocated to the Bases.

13 Footnote '*' to page 3/4 7-1 and 3/4 7-4 is relocated 2' to PSTS 3.6.4.1.

14 The Specification 3.0.4 exception in LCO 3.7.1.3 1 ACTION a, footnote '**' and Specification 3.0.3

a. exception in Action c are deleted as they are no .,

'. longer applicable.

15 Footnote '##' to page 3/4 7-4 is relocated to the 2 e Bases.

r

p ,

I *

' Grand Gulf- Nuclear Station t Technical' Specification-Ieprovement' Program Revision Summary Sheet Proposed LC0/Section: 3.7.2 Rev. _1_ SSW-Shutdown.

11AE Chanae'Descriotion' .Cateoorv 3 16 SR 3.7.2.2.has been limited to valves in lines 3B '!

, servicing only safety related systems or components.- !

l,

] 17~ Footnote '#' to page 3/4 7-1 is relocated to Bases 2 >

t t 3.7.3.-

p 5

t l

5 L

4 x

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-l l'

.,, SSW System - Shutdown 3.7.2 3.7 PLANT SYSTEMS-3.7.2 standby service Water system - shutdown LCO 3.7.2 The Division 1 or 2 Standby Service Water (SSW) subsystem shall be OPERABLE.

APPLICABILITY: MODES 4 and 5, .

When associated systems and components are required to be OPERABLE.

ACTIONS CONDITInN REQUIRED ACTION COMPLETION TIME A.- Required SSW subsystem A.1 Declare affected system Immediately inoperable. or component inoperable, r.

1 Grand Gulf - Unit 1 3.7-3 DRAFT B 2/13/90

P ,

'SSW System Shutdown 3.7.2-t SURVEILLANCE RE0VIREMENTS

.SVRVEILLANCE FREQUENCY.

SR'-3.7.2.1 Verify' the ultimate heat sink basin water 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months level is 2 7.25.M. l SR 3.7.2.2 Verify for each required SSW subsystem 31 days-each. manual, power operated or automatic valve in each required SSW flow path servicing safety related; systems or components not locked, sealed or otherwise secured in position is in its correct position. .

SR 3.7.2.3 Demonstrate each cooling tower fan operates 31 days y

.for 2 15 minutes.

SR 3.7.2.4 Perform a system functional test for 18 months ,

each required SSW subsystem -including i simulated automatic actuation of.the '

subsystem.-

I 1

.- j CROSS-REFERENCES TITLE NUMBER l

Residual Heat Removal - Shutdown 3.4.7 ]

I

.ECCS - Shutdown 3.5.2-l A.C. Sources - Shutdown 3.8.2 Residual Heat-Removal - High Water Level 3.9.8 Residual- Heat Removal - Low Water Level 3.9.9 !

-1 i

i Grand Gulf - Unit 1 3.7-4 DRAFT B 2/13/90 j

. _ _ _ _ . _ _ _ _ _ . . . . _ . . . . ""'"1 .

- SSW System - Shutdown B 3.7.2 B 3.7 PLANT SYSTEMS-B 3.7.2. itandby Service Water System - Shutdown -

-BASES BACKGROUND _The Standby Service Water.(SSW) System-is described in the Bases for LCO 3.7.1.

APPLICABLE .The ability of the SSW system to support long term cooling'of SAFETY the reactor or containmen*. is evaluated in FSAR Chapters 6 ANALYSES. (Engineered Safety Features), 9 (Auxiliary Systems) and 15 (Accident Analyses).. These analyses explicitly assume that-the SSW will provide adequate cooling . support to the equipment required for safe reactor shutdown. These analyses include the evaluation of the long term containment response after a design basis accident (DBA).

During shutdown, refueling, or fuel handling conditions equipment cooling support may be required. Safety analyses assume that the SSW system will support components and systems required to maintain core cooling following a loss of offsite power or loss of normal shutdown cooling capability occurring while in MODES 4' or 5. During fuel handling or handling of loads which, if dropped, could result in release of radioactive material, SSW.

must also be available to provide cooling to the Standby Diesel Generations in the event of a loss of offsite-power. This is required to maintain the Standby Gas Treatment system in operation for secondary containment integrity.

Should -the reactor vessel inadvertently draindown,- cooling

' support will also be required for ECCS operation.

SSW System - Shutdown satisfies the requirements of Selection Criterion 3 of the.NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 1.

(continued)

Grand Gulf - Unit 1 B 3.7-6 DRAFT B 2/13/90

) SSW System --Shutdown l L? B 3.7.2 i j

=

BASES'(continued)'

'] 4

._' LCO The OPERABILITY of the SSW system is required to ensure the. l

l. effective operation of the Residual Heat Removal (RHR) system l

i. '

, in removing heat from the reactor.and the effective operation L

b

~of other safety related-equipment during a design = basis. accident or transient. The OPERABILITY of each independent subsystem of-the SSW system is based on having an OPERABLE. UHS,-an OPERABLE-pump in the subsystem, and an OPERABLE flow path capable' of :

taking suction from the' associated SSW cooling basin and ,

transferring the water to the appropriate plant equipment, as y required. Only those subsystems associated with the systems L and' components required to be OPERABLE by LCO 3.4.7, LCO 3.5.2,:

LCO 3.8.2, LCO 3.9.8 and LCO 3.9.9 are required to be OPERABLE "

D in MODES 4 and 5.-

The OPERABILITY of the UHS is based on having a minimum basin water level at or above elevation-130' 3" mean sea level :

-(which is equivalent to an indicated level of 1 7.25 feet) and ;

having two OPERABLE cooling tower fans for each subsystem > d required to be OPERABLE.

L APPLICABILITY The requirements for OPERABILITY of the SSW system in MODES.4 o

and 5 are governed by'the required OPERABILITY of the equipment ;

L serviced.by the SSW system in those MODES. SSW system '

requirements for MODES 1,2,=and 3 are.specified in LC0 3.7.1.:

ACTIONS L.1 With the required SSW subsystem inoperable, the capability of the affected systems to perform their intended functions cannot be assured. Therefore, the affected systems are required to be declared inoperable and the Required Actions specified in the appropriate LCOs followed.

i en (continued)

'i Grand Gulf - Unit 1 B 3.7-7 DRAFT B 2/13/90

SSW. System - Shutdown ~

B 3.7.2 BASES fcontinued)

SURVEILLANCE SR 3.7.2.1. SR 3.7.2.2. SR 3.7.2.3. SR 3.7.2.4 REQUIREMENTS-The Bases provided for SR 3.7.1.1 through SR 3. U 4 are applicable.

Surveillance Frecuencies In general, Surveillance Frequencies are based on industry accepted practice and engineering-judgement considering the unit conditions required to perform the-test, the ease of performing the test and a likelihood of a-change in.the system / component status, m

~ REFERENCES 1. -NEDO 31466, " Technical Specification Screening Criteria Application and. Risk Assessment," November- 1987.

Grand Gulf - Unit 1 B 3.7-8 DRAFT B 2/13/90

. . Grand Gulf Nuclear Station

Technical Specification Improvomont Program:

(y >

Revision Summary Sheet 2

Proposed LC0/Section: 3.7. 3 - Rev. _1_. HPCS Service Water 1113 l Chance Descriotion Cateaorv

< 1 .LCO 3.7.3'is reformatted from LIMITING CONDITIONS- 1 FOR OPERATION 3.7.1.2 and 3.7.1.3.

'2 . Details of' system-operability requirements are 2

. relocated to the Bases.

3- .The applicability is revised to be when HPCS is 3B required to be OPERABLE.

4- . CONDITION A is reformatted from the LCO 3.7.1.2 1-ACTION statement and LCO 3.7.1.3 ACTIONS'a and b.

5 CONDITION. A does not require the HPCS associated 3B diesel generator to be declared inoperable.

6' SR'3.7.3~.1 is reformatted from SR 4.7.1.3.a.' 1

7 OELETED 8 SR 3.7.3.2 is reformatted from SR 4.7.1.2.a.. 1 9 SR 3.7.3.3 is reformatted from SR 4.7.1.2.b and 1 SR 4.7.1.3.c.

10 CROSS REFERENCES are added. 1 11- The' references in the ACTION statement to.3.5.1 1 and 3.5.2~are deleted.

s l

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b ~ HPCS Service. Water System

3. 7.3 '

s 3.7- PLANT: SYSTEMS 3.7.3- HPCS Service Water' System-i LC0 3.7.3. ' The HPCS Service Water System shall be OPERABLE.

- APPLICABILITY: When the.HPCS system is required to be OPERABLE.

- ACTIONS CONDITION- REQUIRE 0L ACTION' COMPLETION TIME A. HPCS Service Water A.l ~ Declares HPCS inoperable. Immediately

' System inoperable. -

l SURVEILLANCE RE0VIREMENTS i

SURVEILLANCE FREQUENCY l

4 SR 3.7.3.1- -Verify the ultimate heat sink basin-water 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months '!

level is 2 7.25'.

i

u. SR 3.7.3.2 -Verify each manual, power operated.or 31 days- !

automatic valve in the HPCS service water i flow path servicing safety related. ,

l '. equipment not locked, sealed or.otherwise

secured in position is in its correct ,

L _ position. j

~

SR 3.7.3.3 Perform a system functional test 18 months 1 l including simulated automatic actuation >

l of the system. i a

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Grand Gulf - Unit 1 3.7-5 DRAFT B 5/15/90 i

gr. , j Ex ;

~

er . HPCS Service Water System 3.7.3' 5 pl j CROSS REFERENCES i

[ ,

TITLE NUMBER !

ECCS Operating' 3.5.1 ECCS . Shutdown 3.5.2 SSW -Operating! 3.7.1

-SSW.- Shutdowni 3.7.2 s..

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4hN HPCS Service Water System B 3.7.3

'B[3.7 PLANT SYSTEMS" B 3.7.3 HPCS Service Water System BASES 7 BACKGROUND' The High Pressure Core Spray (HPCS) Service Water System is designed to provide cooling water for the removal of heat-from essential support components of- the HPCS system.-

- For the- purpose of this technical specification the HPCS Service Water system consists of the ultimate heat sink (VHS),

one cooling water header (subsystem C-of_ the SSW system), the-U HPCS Service Water Pump and the associated piping'and valves.

' The UHS'for the HPCS Service Water System is described in the-Bases for LC0 3.7.1.

Cooling water is pumped from cooling tower, Basin "A" by the HPCS service- water pump to the essential _ Division 3 support components through the HPCS service water supply header (Subsystem C). After removing heat from the components, the

, water is. discharged to the cooling towers' where the heat is rejected through direct contact with ambient air.

The HPCS service water system specifically supplies cooling water to.the HPCS diesel generator jacket water coolers and-HPCS pump room cooler. . The HPCS service water system pump is :

sized such that it will provide adequate cooling water to the

~

o. Division 3 equipment -required for safe shutdown. Following a design basis accident or transient, the HPCS service water system will operate automatically.

APPLICABLE The ability of the HPCS service water system to support both~

SAFETY short term and long term cooling'of the reactor following a ANALYSIS DBA is evaluated in FSAR' Chapters ' 6 (Engineered -Safety Features), 9 (Auxiliary Systems) and 15 (Accident Analyses).

These analyses explicitly assume the HPCS Service Water System-will provide adequate cooling support to Division 3 equipment required for safe reactor shutdown.

HPCS ' service water system satisfies the requirements of-Selection Criterion 3 of the NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 2.

(continued)

Grand Gulf - Unit 1 B 3.7-9 DRAFT B 5/15/90 l

l m l

HPCS Service Water System-t B 3.7.3

BASE, (continued)

LCO The OPERABILITY of the HPCS service water system is based on-having an OPERABLE UHS, an OPERABLE pump and an OPERABLE flow-

_path capable of.taking suction from the associated SSW cooling basin and - transferring the water to -theEappropriate plant equipment, as required. Requiring the HPCS' service water system OPERABLE assures that HPCS service water' system will be available to provide adequate capability ~ to' meet cooling requirements of the Division 3 ESF equipment required for_ safe shutdown.

The OPERABILITY'of the UHS is based on having a minimum basin water level at or above evaluation 130'3" mean sea level (which -

is equivalent to an indicated level of 17.25 feet).

-APPLICABILITY The requirements for OPERABILITY of the HPCS- service water system including the cooling tower basins are governed by the required OPERABILITY of the Division 3 ESF equipment serviced-by the HPCS service water _ system.

,t ACTIONS a.J L'

When the' HPCS service water system is inoperable, the capability of the HPCS system to perform its intended function; cannot be assured. Therefore, the HPCS system-is required to be declared-inoperable and the Required Actions'specified in the appropriate LC0 followed, w SURVEILLANCE' SR 3.7.3.1

' REQUIREMENTS This surveillance verifies that the associated cooling tower basin has sufficient cooling water inventory (as measured by basin water -level) to satisfy the design basis cooling capability for Division 3 components during a Design Basis Accident. SR 3.7.1.1 addresses this requirement for SSW Division 1 and 2. With the ultimate heat sink inoperable,-the HPCS system must be declared inoperable.

SR 3.7.3.2 Verification of the correct alignment of all applicable valves is essential to ensure the proper flow paths servicing safety related systems or components for the HPCS system.

(continued)

Grand Gulf - Unit 1 B 3.7-10 DRAFT B 5/15/90

~

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m .

.HPCS Service Water System- -

, B 3.7.3 )

, BhSES-(continued) o, : SURVEILLANCE SR 3.7.3.3 I REQUIREMENTS- .

i

[ - (continued) This surveillance verifies that the automatic valve in the HPCS service water system will automatically move to its emergency -)

position and_ that- the HPCS- service water ' pump will '

automatically start to : provide flow: to its safety. related equipment during an~ accident.

Surveillance Freauencies.

In general, Surveillance Frequencies are based on-industry ,

_ accepted practice and engineering judgement considering the-unit conditions required to . perform the test, the ease of performing the' test - and a ' likelihood of a change. in the.

system / component status.

t REFERENCES 1. Regulatory Guide 1.27, " Ultimate Heat Sink for Nuclear Power Plants", Para. C.I. a

2. NED0-31466, " Technical Specification Screening Criteria Application-and Risk Assessment," November 1987..

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l Grand Gulf . Unit 1 B 3.7-11 DRAFT B 5/15/90 1

I 5

I I v _ _ - . _ _ _ _ _ _ __ m _ -__ -__ -_e

' Grand Gulf Nuclear' Station Technical Specification Improvement Program.

Revision Summary Sheet.

Proposed LCO/Section: 3.7.4 Rev. _L Control Room Fresh Air-System him -Chanae Descrintion Cateaorv

-11 LCO 3.7.4 is reformatted from LIMITING CONDITION 1

, . -FOR OPERATION 3.7.2.

2 Footnote '*' to page'3/4 7-5 is-deleted. The 1 applicability statement in LCO:3.7.4 directly includes this requirement.

3 'A NOTE is added to indicate CONDITIONS may be 1 concurrently applicable.

4 CONDITIONS A and B are reformatted from ACTION a. 1 5 CONDITIONS A and C are reformatted from ACTIONS. 1.

b.1 and b.2 except'as discussed:below.

6 The Specification 3.0.4 exception in ACTION b.1 and l' footnote '#' to page 3/4 7-5 are deleted as being no longer applicable.

7 CONDITION B' adds actions to be taken when both 3B+-

. subsystems are inoperable in MODES 1, 2 and 3.-

8 CONDITION ~C permits REQUIRED ACTIONS C.2.1, C.2.2 3B and C.2.3 to'be done instead of REQUIRED ACTION C.1.

.. ACTION b.1^did not provide.this portion.

9 The NOTE-added to REQUIRED ACTION C.2.2 is 1 reformatted from ACTION c.

'10 COMPLETION TIMES'are provided for CONDITION c. 3A Times were previously unstated.

,w 11 SR 3.7.4.1 is reformatted from SR 3.7.2.a except 1-as discussed below.

12 The STAGGERED TEST BASIS of SR 4.7.2.a is deleted. 3B The Compar.ison Document discussion for LCO 3.7.4 characterizes such testing as having negative benefits in justifying deleting the requirement.

NOTE: The Improved Tech Specs adds the STAGGERED TEST BASIS for some other LCOs.

13 ,SR.3.7.4.2 is developed from SR 4.7.2.b.2 and 3B SR 4.7.2.e.

14 SR 3.7.4.3 is developed from SR 4.7.2.b.2 and 3B j SR 4.7.2.f.

____- _ _ - - = _ - _ _ _ - _ _ - _ _

Grand Gulf Nuclear Station Technical Specification' Improvement Program .

, .c Revision Summary Sheet !

Proposed LC0/Section: 3.7.4 Rev. _1_ Control Room Fresh Air System 9

igg Chance Descriotica Cateoorv r

15 SR 3.7.4.4 is reformatted from.SR'4.7.2.b.3 and 1 l SR 4.7.2.c.except as discussed below.

16 The frequency for SR 3.7.4.4 is reduced to 1440 hours0.0167 days 0.4 hours 0.00238 weeks 5.4792e-4 months 3B of_ charcoal adsorber operation from 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months ,

based upon industry experience.. '

]

17 SR 3.7.4.5 is reformatted from SR 4.7.2.d.1. I 18- SR'3.7.4.6 is developed from SR 4.7.2.d 3. 3B 19 SR 3.7;4.7 is developed from SR 4.7.2.d.2. '3B

-l 20 Actuation instrumentation in SR 4.7.2.d.2 is moved 1 1 to LCO 3.3.7.1. ,

21 SR 4.7.2.b.4 is verified during performance of. 38 <

SR 3.7.4.2, SR 3.7.4.3 and SR 3.7.4.5. +

22 Details for system operability are removed from LCO 2 3.7.'2 and relocated to the Bases.

\

23: SR-3.7.4.1 deletes requirements to initiate from 4 1: tae control room.the CRFA subsystems currently in j

SR 4.7.2.a.

L L 24 SR 3.7.4.1 deletes requirements to have subsystem 4- -

run at least 10. hours continuously currently in CTS SR 4.7.2.a. '

25 Requirements to perform SR 4.7.2.b after any 4 I structural maintenance in the HEPA filter or charcoal adsorber housing is deleted. <

l 26 'The methods and acceptance criteria of SR 4.7.2.b.3 2

and 4.7.2.c are relocated.

27 The control room manual initiation test condition 4 i

is deleted from SR 4.7.2.d.2.

i lf1

CR Fresh Air. System- '

3.7.4

.v 3.7 PLANT SYSTEMS- ;

3.7.4 Control Room Fresh Air System Y

LC0 3.7.4 Two Control Room Fresh Air (CRFA) subsystems shall- be ;

OPERABLE.

3 ' APPLICABILITY: MODES 1, 2, 3, 4, and 5, :t When handling irradiated fuel er :::p nd:d '!;ht '::d: 0cer >

4 r:di:ted fu ' in the primary or secondary containment, i

.........................N0TE--------------------- --- .

. Conditions A, B and C may be concurrently applicable. "

ACTIONS 1

CONDITION COMPLETION TIME' i REQUIRED ACTION A. One CRFA subsystem A.1 -Restore th'e ' inoperable 7 days 'from inoperable, subsystem to 0PERABLE -discovery-of status, inoperable

subsystem l.

^4 B. : Required Action and B.1 Be in' MODE 3. 12. hours associated Completion i . Time of Condition A- AND l not~ met in MODE 1,-2, or 3.-

B.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Q8' E Both CRFA subsystems ?!

Linoperable in MODE ~1, 2, or 3.

,(continued)

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Grand Gulf - Unit 1 3.7-7 DRAFT B 11/30/89

CR Fresh Air System-3.7.4-ACTIONS (continued)

CONDITION REQUIRED ACTION- COMPLETION-TIME.

C.: Required Action and C.1 Place OPERABLE subsystem Immediately associated Completion in the isolation mode of Time of Condition A- . operation, not met in M00E-4, 5, or when handling QB irradiated fuel ee

:;:nded it;ht 1::d: C.2.1 Suspend CORE ALTERATIONS. :Immediately re:r 1 r:di:t:d ft:1 in the primary or AHQ secondary containment.

C.2.2 ---------NOTE-----------

Provisions of LCO 3.0.3 QB are not applicable.

< Both CRFA subsystems inoperable 11n: M00E 4, Suspend handling of Immediately 5, or when handling irradiated fuel in the

,~ irradiated fuel 4s' primary and secondary'

d

d light 1::d: containment.

r irr: dict:d ft:1 in the primary or- . &HQ >

secondary containment.

C.2.3 Suspend operations with As soon as a potential for draining practicable the reactor vessel.

B%

C.2.4 Susp;ad h;adlia; light As :::a ;;

-10:d: Over irr:dt:ted pr:: tic:ble ftrek Grand Gulf - Unit 1 3.7-8 DRAFT B 11/30/89 .

CR Fresh Air System 3.7.4 SURVEILLANCE REOUIREMENTS SURVEILLANCE -

FREQUENCY SR 3.7.4.1 -Demonstrate each subsystem operates with' 31 days flow through the HEPA filters and charcoal adsorbers for > 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months with the heaters operabTe, SR 3.7.4.2- Demonstrate < 0.05% penetration of the 18 months HEPA filters by a DOP test at a system flow rate of 3600 to 4400 cfm. M Once within 7 days-after painting, fire, or' chemical release in the area being

, serviced by the~

ot filter E ..

Prior to declaring subsystem OPERABLE after ea'ch complete or partial

- replacement of a filter (continued)

Grand -Gulf - Unit 1 3.7-9 DRAFT B 11/30/89 a

4 ,

CR FreshLAirLSystem? -,

y,g '3.7.4' .-

( SURVEILLANCE REOUIREMENTS (continued) i SURVEILLANCE FREQUENCY l SR. 3.7.4.3 Demonstrate < 0.05% bypass leakage. 18 months through the adsorber section by a .i halogenated hydrocarbon = test at a system : AND flow rate of 3600 to 4400 cfm. ,

-Once:within J 7 days =after painting, fire, '

or chemical release in-l :

the area being-serviced by the-e ,

filter a

bBk Prior.to declaring

~

subsystem:

OPERABLE:

after each complete or E -

partial replacement of an adsorbor. bank 1

(continued):

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Grand Gulf - Unit 1 3.7-10 DRAFT B 11/30/89 ic

CR Fresh Air System:

3.7.4'

SURVEILLANCE REGUIREMENTS (continued)-

SURVEILLANCE ' FREQUENCY:

SR 3.'7.4.4 l-----------------NOTE-----------------

-Analysis must be completed within

31. days of sampling.

Remove and perform a laboratory analysis 1440 hours0.0167 days 0.4 hours 0.00238 weeks 5.4792e-4 months

.of a representative charcoal adsorber of charcoal sample for methyl iodide / pe.ne.4raHon, adsorber operation M

18 months-Once within

. 7 days.after painting, fire, or chemical release in.the '

area being- '

serviced by<the-filter

'SR -3.7.4.5 -Demonstrate'< 7.2 inches water gauge '18monihs pressure drop across'the combined HEPA'.-

filters and charcoal ~ adsorber banks at a o system flow rate of 3600 to 4400 cfme.

SR. 3.7.4 6

. Demonstrate heaters-dissipate from 18 months 18.6 to 22.8 kw.

(continued) e Grand Gulf - Unit 1 3.7-11 DRAFT B 11/30/89

db 4

.CR Fresh Air. System. .

3.7.4

~

, SURVEILLANCE RE0VIREMENTS (continued)- ,

t SURVEILLANCE FREQUENCY SR 3.7.4.7 Demonstrate each subsystem automatically 18 months switches,to the isolation mode of operation on receipt of an actuation. !

signal, '

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CROSS-

REFERENCES:

~ None I 4

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L Grand Gulf - Unit 1 3.7-12 DRAFT B 11/30/89

.- .L. _ _ _ . _ - _ _ _ _ _ . - _

, u CR Fresh Air System; I

3. 7. 4-B3.7.'Pi. ANT' SYSTEMS ,

t B 3.7.4 Control Room Fresh Air System 4

, . - BASES

~

BACKGROUND The-Control Room Fresh Air (CRFA) system is designed to provide-a radiologically. controlled environment to ensure the control !

, . room will remain habitable for personnel during and following a 1 design basis; accident -(DBA) (Ref.1). To meet these 1

-requirements, this. system is designed in conjunction with control-room design provisions such that'the radiation exposure to personnel inside the control room-is less than 5 rem whole' .

body consistent with the requirements of General Design Criterion 19 of Appendix A to 10 CFR 50. !

The safety related function of the CRFA system used to control I radiation exposure consists of two independent and: redundant

4. high efficiency air filtration- subsystems. Each subsystem consists of a demister, an electric heater, prefilter, e high :-!

efficiency-particulate air (HEPA) filter, an activated charcoal adsorber section, a second HEPA. filter and a fan. -!

In addition to the safety related standby emergency filtration- ,

function, parts of the CRFA system are operated to maintain the 1 control room environment during normal operation. Upon receipt *

. of an. actuation signal (indicative of conditions that could

. result' in radiation exposure to control room personnel) the CRFA system automatically switches to.the isolation mode of operation' to prevent infiltration.of contaminatedf air into the control .

room. : A' system of dampers isolates the control- room and control

, room air flow.is recirculated and procassed through either of the two filter-subsystems. When con: n ions permit, fresh air can be manually brought into the control room through the.

charcoal filter system.

4 (continued) H i

v 1

Grand Gulf - Unit 1 8 3.7-12 DRAFT B 11/30/89

CR:FreshAihSystem. I

~3.7.4

-BASES (continuedI "

' APPLICABLE l The ability of the CRFA system to maintain the. habitability of 1

, SAFETY- - the. control room is an explicit assumption for' the safety

~ ANALYSES analyses. evaluated in FSAR Chapters 6, (Engineered Safety- J Features)-and:15,(AccidentAnalyses). The isolation mode ,

of the CRFA system is assumed to operate following a loss Of Coolant Accident (LOCA), main -steam:line break (MSLB), fuel i

..-The

^

n, handling radiological accident a,- :ontrol.

doses to= control' rod roomdrop accident personnel as (CRDA)lt a resu of the

, various design basis accidents are summarized in Reference 2. ,

In- all cases,. the' doses are within the limits of 10 CFR 50, t

- Appendix A, General: Design Criterion.19.

r Control Room' Fresh Air System satisfies the requirements of. .i Selection Criterion 3 of. the NRC Interim Policy Statement on

- Technical Specification Improvements as documented in Reference 5.

LCO Two. redundant subsystems of the CRFA system are required to i ensure at least one is available-assuming a single failure '

disables the other subsystem. The CRFA subsystem consists of-

two independent and redundant filtration trains. Should any component:in one subsystem-fail, filtration can be performed by * ,

the other' subsystem. The OPERABILITY of;each: independent subsystem is based on having adequate system flow and OPERABLE ~ -

HEPA filters. charcoal adsorbers= and heaters. ' A description of what'is required for CRFA to be considered OPERABLE.is provided ,

in'the Background Section. 3 A

t

' APPLICABILITY The standby emergency filtration portion of the CRFA system is required to be OPERABLE in MODES 1, 2,'3, 4,-5 and when handling

~

= irradiated fuel er r5:n 5::dling ':::p: d:d light 10:d: "cr .

irradiated f"el in the primary'or: secondary. containment to ensure 3 the. control room will be habitable for personne1Lduring and J following a design basis accident, y

+

(continued) ,

Grand Gulf - Unit 1 B 3.7-13 DRAFT B 11/30/89

c n

[; :CR Fresh Air System t

3.7.4 x

BASES ' (continued) p , -ACTIONS: AJ l With one CRFA' subsystem inoperable, the remaining OPERABLE -

-subsystem can maintain the habitability of. the control room

.during the postulated design basis accidents assuming no '

j additional failures in the OPERABLE subsystem. However, if a

single active component fails concurrent with the postulated design basis accident, depending on the specific' failure, the 1 CRFA system may not be able to perform itsuintended safety

. function. Therefore, system reliability is. reduced and operation is only allowed to continue for a-limited time. 1 B.1. B.2- H In MODES 1, 2 and 3, with an inoperable subsystem not restored to OPERABLE status and the associated-Completion Time not met i

.or.both subsystems inoperable, the CRFA system may not be < li L capable.of performing its intended safety function and the l reactor is required to be in MODE 3 and subsequently in MODE 4.

C.1. C.2.1. C.2.2. C.2.3 .

1 In MODES 4, 5 and when handling irradiated fuel-er : :pertd i

'!C ht le & ever "r:di:ted ft:1 in the primary or secondary ,

containment, if an inoperable subsystem cannot be-restored to 0PERABLE- status and the associated Completion Time is not met,

' the' remaining OPERABLE subsystem may be 'placed in the: isolation l- mode of operation. This ensures'the CRFA system is operating L

L prior to any potential-design' basis' accident that'could require the CRFA system to actuate.. This action provides a continuous.

check of the operation of the'CRFA system. . Alternatively, CORE-'

ALTERATIONS, handling' of' irradiated fuel er. :;cred li;ht 1:26 ever "radiatedLft:1 in the primary or secondary containment and 4 l- operations,that could drain the reactor vessel must be suspended.

This-eliminates the potential accidents under these conditions ;

that could require the-CRFA system to function. Suspension-of

these activities shall not preclude completion of the movement of a component to a safe,. conservative position.

Comoletion Times

+

t 1

All Completion Times are based on industry accepted practice and engineering judgement considering the number of available systems- and the time required to reasonably complete the :

Required Action. 4 (continued) 3 Grand Gulf - Unit 1 B 3.7-14 DRAFT B 11/30/89 H

o l,

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CR Fresh Air System 3.7.4 l- BASES (continued) ,

SURVEllt.ANCE General REQUIREMENTS

, In addition to the ANSI N510 test requirements, and normal preventive and post maintenance testing, there are a number of specific tests which must be performed to ensur6 proper functioning of the CRFA system.

SR 3.7.4.1 Standby systems should be checked will start and function pro >erly. periodically to ensure they This Surveillance Requirement ensures each sussystem will start on demand and continue to operate. Operation with the heaters on for greater than or equal to 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months every 31 days reduces the buildup of moisture on the adsorbers and HEPA filters.

SR 3.7.4.2. SR 3.7.4.3. SR 3.7.4.4 and SR 3.7.4.i These Surveillance Requirements demonstrate the designed

-filtration capability of the system is maintained by verifying the system flow rate, HEPA filtars, and charcoal adsorbers satisfy the in place testing acceptance criteria, the surveillance intervals and procedures of Reference 3 and ANSI N510 1975. The laboratory analysis of a representative carbon ,

sample .(SR 3.7.4.3) must be performed in accordance with the testing criteria of Regulatory Position C.6.a of Reference 3.

a The carbon sam >1e to be used in this test must be obtained in accordance witi Regulatory Position C.6.b of Reference 3. The in place acceptance criteria of SR 3.7.4.2 and SR 3.7.4.3 are defined in Regulatory Position C.5 of Reference 3. The system flow rate is verified during subsystem operation for SR 3.7.4.2, SR 3.7.4.3 or SR 3.7.4.5 when tested in accordance with ANSI N510 1975.

In addition to the 18 month Frequency, SR 3.7.4.2, SR 3.7.4.3 and SR 3.7.4.4 are also required whenever major modifications or events (e.g. painting, fire or chemical release) which may have affected the integrity of the HEPA filters or charcoal adsorbers have occurred. For the purpose of this specification,

" area being serviced by the filter" means the area where painting, fire or chemical release occurred and from which suction is being taken by an operating fan, (continued) i Grand Gulf - Unit 1 B 3.7-15 DRAFT B 11/30/89 i ~ '

a CR Fresh Air System 3.7.4 BASES (continued)

SURVEILLANCE SR 3.7.4.6

REQUIREMENTS (continued) This Surveillance Requirement verifies the duct heater performance of the CRFA system. The test is performed in accordance with Section 14 of ANSI N510 1975 with the exception of the 5% current phase balance of Section 14.2.3.- The offsite power system for the Grand Gulf Nuclear Station consists of a non transpositional 500 kV grid. The grid has an inherent unbalanced load distribution which results in unbalanced voltages in the plant. Voltage unbalances exceeding the 5%

criteria of ANSI N510 1975 are not atypical.

SR 3.7.4.7 This Surveillance Requirement verifies the CRFA system will automatically switch to the isolation mode of operation to maintain control room habitability on receipt of an actuation signal.

Surveillance Frecueltciel l In general, Survei' oce Frequeficies are based on industry accepted practice (ngineering judgement considering the unit conditions r- red to >erform the test, the ease of

performing the test ,nd a li(elihood of a change in the syste# component status. The Surveillance Frequencies for testing of the HEPA filters and charcoal absorber units are consistent with the requirements of Reference 4.

REFERENCES 1. Grand Gulf FSAR, Section 6.5.1.1.

2. Srand Gulf FSAR, Section 15,

3. Regulatory Guide 1-.52, " Design Testing and Maintenance 1 Criteria for Post Accident Engineered Safety Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Jater-Cooled Nuclear Power Plants",

Revision 2, March 1978.

4. Generic letter 83-13, " Clarification of Surveillance Requirements for HEPA Filters and Charcoal Absorber Units in Standard Technical Specifications on ESF Cleanup System ," Harch 2, 1903.

S. NEDO 31466, "Tectmical Specification Screening Criteria Application and Risk Assessmett,' Nomaber 1987.

i l

Grand Gulf Unit 1 S 3,7-16 DRAFT B 11/30/89 L

i:

Grand Gulf Nuclear Station

, Technical Specification Itprovement Program p Revision Summary Sheet Proposed LCO/Section: 3.7.5 Rev. _1_ Main Condenser Offaas 1115 Chance Deterietion Cateaorv 1 LCO 3.7.5 is reformatted from LIMITING CONDITION 1 FOR OPERATION 3.11.2.7.

- . 2 DELETED 3 CONDITION A is reformatted from the ACTION statement. 1 4 CONDITION B revises the shutdown requirement of the 3B ACTION statement to MSIV isolation.

5 SR 3.7.5.1 is reformatted from SR 4.11.2.7.2 1 except as discussed below.

6 The Specification 4.0.4 exception in SR 4.11 2.7.2 1 (footnote "**" to page 3/411-17) is deleted as it is no longer applicable.

7 SR 4.11.2.7.1 is deleted. (LCO 3.3.7.12 is relocated). 4

, See Note 1 below.

8 The words " primary coolant" in SR 4.11.2.7.2.b 4 is not included in SR 3.7.5.1.

/ NOTE

1. Per NRC letter on split document, the instrumentation is to be retained and also the GL on removal of RETs retains this requirement. SR 3.7.5.1 takes credit for the continuous monitoring of the offgas effluent in setting the surveillance frequency.

i- ,

Main Condenser Off as

3. 5 1 3.7 PLANT SYSTEMS E

3.7.5 Main condenser off m LCO 3.7.5 The gross gamma radioactivity rate of the noble gases ,

measured at offgas recombiner effluent shall be < 380 1

mil 11 curies /second, after 30 minutes decay.

APPLICABILITY: MODE 1, ;

MODES 2 and 3 with a steam jet air ejector in operation.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME i

A. Gross gamma A.1 Restore gross gamma 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months radioactivity rate radic. activity rate to

> 360 mil 11 curies / within limits.

second after 30 e minutes decay.

B. Required Action and B.1 Isolate all main steam 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion lines.

Time of Condition A not met.

7 Grand Gulf - Unit 1 3.7-13 DRAFT B 2/13/90

Main Condenser Offgas 3.7.5-

'SURVEll!,ANCE REOUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1 Perform an isotopic analysis of a 31 days-representative sample of gases taken at the offgas recombiner effluent. AND Once within.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after a

> 50%

Increase in the nominal steady state fission gas-release after factoring out increases due to changes in THERMAL POWER level 4

CROSS-

REFERENCES:

None i

i i

i.

Grand Gulf - Unit 1 3.7-14 DRAFT B 2/13/90 l

l

, Main Condenser B 3.7.5 B 3.7 PLANT SYSTEMS i B 3.7.5 Main condenser offeas Bal(S i BACKGROUND- During plant operation, steam from the low pressure turbine is :

exhausted directly into.the condenser. Air leakage and l noncondensible gases are collected in the condenser, then i exhausted through the steam jet air ejectors to the Offgas l System. The offgas from the main condenser normally includes !

radioactive gases, i

's The Main Condenser Offgas System has been incorporated into the ,

plant design to reduce the gaseous radwaste emission. This !

system uses a catalytic recombiner to recombine radiolytically 1 dissociated hydrogen and oxygen. The gaseous mixture is cooled <

by the offgas condenser and the water and condensibles are 4 stripped out by the offgas condenser and moisture separator. l The radioactivity of the remaining gaseous mixture (i.e., the j offgas recombiner effluent) is monitored downstream of the j offgas condenser. ,

1

~{

APPLICABLE The main condenser offgas gross gamma radioactivity rate is an SAFETY initial condition of the Offgas System Failure Event (Ref.1). l' ANALYSES The analysis assumes a gross failure in the Offgas System that results in the rupture of the Offgas System pressure boundary.

The gross gamma radioactivity rate is controlled to ensure that :

during the event the calculated offsite doses will be well within the limits of 10 CFR 100.

l Main Condenser Offgas satisfies the requirements of Selection 4 Criterion 2 of the NRC Interim Policy Statement on Technical ,

Specification Improvements as documented in Reference 2.

i LCO To ensure compliance with the assumptions of the Offgas System Failure Event (Ref. 1), the fission product release rate should be consistent with a noble gas release to the reactor coolant of 100 sci /MWt sec at 30 minutes decay. The LC0 is established consistentwiththisrequirement(3833MWtx100uti/MWt-sec= '

380 mil 11 curies /second).

(continued)

Grand Gulf - Unit.1 B 3.7-17 DRAFT B 2/13/90

c:

i i

Main Condenser B 3.7.5 BASES feontinued)

APPLICABILITY The LCO is applicable during MODES 1,2 and 3 when steam is being exhausted to the main condenser and steam jet. air ejectors are being used to maintain condenser vaccum. In MODES 4 and 5, steam is not being exhausti.d to the main condenser and the requirements are not applicable.

L ACTIONS Ad If the offgas radioactivity rate limit is exceeded, a limited l; time is permitted to restore the gross gamma radioactivity rate to within the limit because of the large margin to permissible dose and exposure limits.

L B.1. B.2 b If the gross gamma radioactivity rate is not restored to within l

the limits and the associated Completion Time is not met, all main steam lines must be isolated. This isolates the condenser -

I from the source of the radioactive steam. [

Comoletion Times All Completion Times are based on industry accepted practice and engineering judgement considering the number of available systems and the time required to reasonably complete the Required Action.

SURVEILLANCE SR 3.7.5.1 REQUIREMENTS This surveillance periodically analyzes a sample of the offgas to ensure that the required limits are satisfied. If the measured rate of radioactivity increases significantly (by 2 50% percent after correcting for expected increases due to changes in THERMAL POWER), an isotopic analysis is performed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , after the increase is noted, to ensure that the increase is not indicative of a sustained increase in the radioactivity rate. The Surveillance Frequencies are considered adequate based on the availability of ;

instrumentation to continuously monitor the offgas. '

REFERENCES 1. Grand Gulf Unit 1 FSAR, Section 15.7.1.

2. NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment," November 1987.

Grand Gulf - Unit 1 B 3.7-18 DRAFT B 2/13/90 ,

t

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