Text

Proposed Revs to Previously Submitted Tech Specs with LAR Related to RAI on Conversion to ITS
	ML20206H000
Person / Time
Site:	Farley
Issue date:	04/30/1999
From:	; SOUTHERN NUCLEAR OPERATING CO.
To:	;
Shared Package
ML20206G998	List: L-99-017, Forwards Responses to NRC RAI Questions for Chapters 3.1, 3.2,3.5,3.7 & 3.9 of Ts.Attached Pages Include Proposed Revs Previously Submitted LAR to Rais,Grouped by Chapters & RAI Numbers; ML20206H000;
References
	NUDOCS 9905110009
	Download: ML20206H000 (130)
	v • d • e

/ focR MTC[

B 3.1.4f B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1. Moderator Temperature Coefficient (MTC) T6 W %

I BASES BACKGROUND According to GDC 11 (Ref.1), the reactor core and its interaction with the Reactor Coolant System (RCS) must be designed for inherently stable power operation, even in the possible event of an accident. In particular, the net reactivity feedback in the system must compensate for any unintended reactivity increases.

The MTC relates a change in core reactivity to a change in~

reactor coolant temperature (a positive MTC means that reactivity increases with increasing moderator temperature; conversely, a negative MTC means that reactivity decreases !

with increasing moderator temperature). The reactor is I designed to operate with a negative MTC over the largest i possible range of fuel cycle operation. Therefore, a coolant temperature increase will cause a reactivity decrease, so that the coolant temperature tends to return j toward its initial value. Reactivity increases that cause a coolant temperature increase will thus be self limiting, and p.Tp stable power operation will result.

(( Q MTC values are predicted at selected burnup during e y safety evaluation analysis and are confi to be acceptable by measurements. Both initial dreloadres l are designed so that the beginning of cyc1 MTC is less than zero when THERMAL POWER is at RIP. e actual value of the MTC is dependent on core characteristics, such got.

as fuel loading and reactor coolant soluble boron concentration. ThecoredesignmayQecu ditional fixed distribu poisons to yield an MTC aT within the range l 60L z n the plant accident analysis. The end of cycle C is also limited by the requirements of the acc dent analysis. Fuel cycles that are designed to achieve high burnups or that have changes to other characteristics are evaluated to ensure that the MTC does not exceed the limit. O The limitations on MTC are provided to ensure that the value of this coefficient remains within the limiting conditions assumed in the FSAR accident and transiont analyses.

~

9905110009 990430 ADOCK 05 348

$DR y (continued)

WOG STS B 3.1-18 Rev 1, 04/07/95

r i

Associated Package Changes for RAI- 3.1.3-5 O

r u-

F i

FarleyNuclear Plant ITS conversion Subminal I l

Imoroved Technical Specifications Conversion Reviewers Guide l Changes to the CTS requirements categorized as Relocated are annotated with an "R" in the Enclosure 1 markup and Enclosure 2 DOC. Relocation changes are considered generic and each

! relocated TS will not be specifically discussed in this SHE. A specific DOC for each CTS i identified for relocation is provided in Enclosure 2. This evaluation will be applicable to each of the changes identified with an "R" in the Enclosure 1 markup of the CTS and the associated Enclosure 2 DOC.

l MORE RESTRICTIVE ("M" labeled chanoen)

This generic category consists of changes that modify the CTS to add new requirements or revise existing requirements to be more stringent. These changes are typically made to conform to j applicable requirements in the STS, correct discrepancies, or remove ambiguities from the CTS.

i More restrictive changes are proposed only when such changes are consistent with:

~

1)The FNP currentlicensing basis;

2) The applicable FNP safety analyses; and

3) Good engineering practice such that the availability and reliability of the affected equipment is not reduced.

l C%mnoaa to the CTS requirements categorized as More Restrictive are annotated with an "M" in the Enclosure 1 markup and Enclosure 2 DOC. These changes are considered generic and each change identified as More Restrictive will not be specifically discussed in this SHE. This evaluation will be applicable to each of the changes identified with an "M" in the Enclosure 1 markup of the CTS i and the associated Enclosure 2 DOC.

REMOVAL OF REOUIREMENTS FROM RETAINED TS ("LA"lahaled chanaan)

This generic category applies to changes that modify the CTS to achieve consistency with the style and format of the STS by removing some information and requirements from the CTS. This category of change is considered to be less restrictive (no longer controlled by TS) and is referred to as Less Restrictive - Generic. The "partiabelocation" ofinformation can not be addressed by the

%[9 application of the criteria in the NRC Final Policy Statement on Technical Spe Improvement for Nuclear Power Reactors. The NRC Policy Statement criteria is only intended to

[ apply to entire TS. This generic category is stended to apply to changes that remove some information that is typically descriptive in nature regarding the equipment, system (s), actions or

[

T surveillances from a CTS that is retained in the TS. The information that is removed from the CTS is proposed for inclusion in the TS Bases, FSAR, or other licensee controlled documents as appropriate. Altemately, the information may already be included in one of the above documents or an applicable CFR or other regulatory document. The removal of this information from the CTS to licensee controlled documents is acceptable because the documents will be controlled through a process which utilizes 10 CFR 50.59 or other applicable CFR. Reliance on the requirements in the regulations is acceptable as the regulations arc controlled by the NRC. Therefore, the information that has been moved, or is already contained in another document, continues to be maintained in an appropriately controlled manr.er. The proposed changes will not impact the actual information moved out of the CTS but, in the case ofinformation moved to licensee controlled documents, will

[ l reduce the level of regulatory control over this information.

Al-4 l-

Associated Package Changes for Addendum RAI-3.1.21 4 O

o FNP TS Conversion Enclosure 2 - Discussion of Changes to CTS ,

! Chapter 3.1 - Reactivity Control Systems i

l l

CTS - NA FNP ITS 3.1.2 CORE REACTIVITY DOC NQ SHE DISCUSSION 1 M The STS LCO 3.1.2, Core Reactivity is added to the proposed FNP ITS.

l This new LCO was derived from the CTS 4.1.1.1.2 surveillance for core !

g reactivity balance consistent with the development of this LCO in the STS.

g The surveillance for core reactivity balance was previously included in the b g g7 SDM LCO and was required to be performed in Mode 1. However, in g developing the STS, the SDM Mode 1 requirements were removed from the r TS. In addition, the SDM LCO Actions did not specifically address the-core reactivity not within limit. As such, a new LCO for Core Reactivity was included in the STS. The new LCO is applicable in Modes 1 and 2 and

[

provides specific Actions for core reactivity not within limit. The basis for including this new LCO in the STS is also applicable to FNP and as such I

this LCO is also included in the FNP ITS. However, as the Core Reactivity l LCO contains new Action requirements, which include placing the plant in l Mode 3 if the Actions are not met, the addition of this new LCO is considered a more restrictive change.

I l

l' Chapter 3.1 E2-1-C April,1999

ll3 Core React 3.1 REACTIVITY CONTROL SYSTEMS 3.1. Core Reactivity 1

LCO 3.1. The measured core reactivity shall be within

1% ok/k of predicted values.

APPLICABILITY: MODES I and 2.

ACTIONS

~

CONDITION REQUIRED ACTION COMPLETION TIME.

A. Measured core A.1 Re-evaluate core Q2 hour reactivity not within design and safety limit. analysis, and ,,7 M6 detemine that the reactor core is

g acceptable for ~T6TF-NL.

,g7 continued operation.-

U A.2 Establish appropriato Q { hour [

operating restrictions and SRs.

l B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion .

Time not met.

WOG STS 3.1-3 Rev 1, 04/07/95

/69 Core Reactivity i 8 3.1. l BASES I2;TF-l%

LCO that predicted is larger' than expected for normal operation (continued) and should therefore be evaluated.

When measured core reactivity is within 1% ak/k of the predicted value at steady state thermal conditions, the core is considered to be operating within acceptable design limits. Since deviations from the limit are normally y detected by comparing predicted and measured steady state

/ RCS critical boron concentrations, the difference between measured and predicted values would be approximately 100 ppe (depending on the boron worth) before the limit is reached.

These values are well within the uncertainty limits for :

analysis of boron concentration samples, so that spurious violations of the limit due to uncertainty in measuring the RCS boron concentration are unlikely.

' APPLICABILITY The limits on core reactivity must be maintained during

. MODES I and 2 because a reactivity balance must exist when the reactor is critical or producing THERMAL POWER. As the

fuel depletes, core conditions are changing,-and confirmation of the reactivity balance ensures the core is

@A operating as designed.* This Specification does not apply in T MODES 3, 4, and 5 because the reactor is shut down and the g reactivity balance is not changing.

In MODE 6, fuel loading results in a continually changing

, core reactivity. Boron concentration requirements (LC0 3.9.1, " Boron Concentration") ensure that fuel movements are performed within the bounds of the safety analysis. An SDM demonstration is required during the first startup following operations that could have altered core reactivity control rod (e.g., fuel movement, control rod replacement,'

shuffling).

ACTIONS A.1 and A.2 Should an anomaly develop between measured and predicted core reactivity, an evaluation of the core design and safety analysis must be performed. Core conditions are evaluated g/ to determine their consistency with input to design calculations. Measured core and process parameters are evaluated to determine that they are within the bounds of (continued) .

WOG STS B 3.1-15 Rev 1, 04/07/95 4

4

CoreReactivityf B 3.1.gr BASES 7np_g ACTIONS A.1 and A.2 (continued) the safety analysis, and safety analysis calculational TFT1 - q t_ models are reviewed to verify that they are adequate for representation of the core ions. The required q 3 Completion Time or uz nour s based on the low probability of a DBA occurring during this period, and allows sufficient time to assess the physical condition of the reactor and complete the evaluation of the core design and safety analysis.

Following evaluations of the core design and safety analysis, the cause of the reactivity anomaly may be resolved. If the cause of the reactivity anomaly is a -

mismatch in core conditions at the time of RCS boron concentration sampling, then a recalculation of the RCS boron concentration requirements may be performed to demonstrate that core reactivity is behaving as expected.

If an unexpected physical change in the condition of the core has occurred, it must be evaluated and corrected, if )

possible. If the cause of the reactivity anomaly is in the calculation technique, then the calculational models must be revised to provide more accurate predictions. If any of these results are demonstrated, and it is concluded that the reactor core is acceptable for continued operation, then the boron letdown curve may be renormalized and power operation may continue. If operational restriction or additional SRs 1 are necessary to ensure the reactor core is acceptable for l TsTF-lW continued operation, then they must be defined. I 4

L #

7 The required Completion Time ofU2 hours is adequate for preparing whatever operating restrictions or Surveillances that may be required to allow continued reactor operation.

Ad b

y[+s.+A If the core reactivity cannot be restored to within the 1% ak/k limit, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3.within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . If the SDM for MODE 3 is not met, then the boration required by SR 3.1.1.1 would occur. 'The allowed Completion Time is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

(continu'ed)

WOG STS B 3.1-16 Rev 1, 04/07/95

)

I Associated Package Changes for Chapter 3.2 i

f I

1 I

i 4

l I

I l

l I 1

f I

l Associated Package Changes for RAI- 3.2.13 l

O l

m

- FNP TS Conversion Enclosure 2 - Discussion of Changes to CTS Chapter 3.2 - Power Distribution Limits CTS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - Fo (Z)

FNP ITS 3.2.1 HEAT FLUX HOT CHANNEL FACTOR (Fo(Z))

DOC RQ SHE DISCUSSION change is made ts conform with the presentation and format of the STS and is considered administrative in nature.

7 LA The CTS 3/4.2.2 earveillances are revised consistent with the STS. The CTS surveillances contain details describing the surveillance requirements that are no't included in the corresponding STS surveillances. The STS surveillances are described in detail in the associated TS bases. As such, the CTS surveillances are revised to remove information that is already ~

present in the STS bases or that may be placed in the STS bases. The removal of this information from the TS and placement in the bases is consistent with the format and presentation of the STS. Reliance on the information contained in the STS bases for guidance in performing the associated surveillances is acceptable since changes to the information in the bases is controlled by the Bases Control Program specified in the administrative controls section of the TS. ,

l 4

8 LA The CTS 3/4.2.2 Fo(Z) relationships and specific values provided in CTS surveillance 4.2.2.2.d are removed from the TS and placed in the Core Operating Limits Report (COLR) consistent with the STS. The COLR contains the values and formulas used to determine the limits associated with those values for specific core parameters. The information contained in the COLR is developed using NRC approved methodologies. The methodologies used to determine the specific values e.nd limits contained in the COLR are specified in the report section of the TS Administrative Controls. As such, the removal from TS and inclusion in the COLR of these values is acceptable since changes to the values of these parameters are controlled by the use of NRC approved methodologies specified within the FNP TS. Additionally, the requirement to provide the COLR (including any revisions) to the NRC is also specified in the report section of the TS administrative controls.

I The

footnote to CTS surveillance requirement 4.2.2.2.e.1 is revised d 8a M consistent with the STS. The CTS

footnote is revised to require that of Fn(Z) be measured once after each refueling prior to THERMAL POWER exceeding 75% RTP consistent with the frequency for performing ITS SRs 3.2.1.1 and 3.2.1.2. As this places an upper limit on power for the initial ,

performance of these surveillance requirements after each fuel loadmg that l doesn't currently exist in the CTS, it is seen as a more restrictive change.

Chapter 3.2 E2-3-B April,1999 4

N F,(Z) {F Methodology) ~1 J.Z. 5 SURVEILLANCE REQUIREMENTS

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

During power escalation at the beginning of each cycle, THERMAL POWER may be increased until an equilibrium power level has been achieved, at which a power distribution map is obtained.

SURVEILLANCE (1) FREQUENCY

M sW8 SR 3.2.1.1 Verify , Z) is within mit. Once after each

refueling prior to THERMAL POWER exceeding

. 75% RTP 3 'S.g>3 i

g b

Once ithin' hours

after achieving C.T 5 4.1.2.1.E.1 equilibrium conditions 2, a after 20/* , g, by tRTP, t TH L at which )

was last verified M

31 EFP0 ,

thereafter (continued) l WOG STS 3.2-6 Rev 1, 04/07/95

.: - 1 1") #.

Fa(Z) ((F. Methodoloav) l 3.2.%

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY n m-3, )

SR 3.2.1.2 -------------------NOTE---- g ------------

If $%(Z) is within limits ind measurements indicate Z - [3 max ver K(Z) , g.ypA7,k.I hs creased since th revious evaluation j 3ef _[ PM C Se C.c LR.

a. ctorfo~f fl.027 nerease FRZ) b -

i and reverify , s within imits; l or O j &. b. Repeat SR 3.2.1.2 once per 7 EFPD Mye until two successive flux maps n cate

,, , ;,,4 e;- z ,f) -

6%.< maximum over K(Z) ,

has not increased. 1 Verify is within init. Once after each refueling prior to THERMAL

" POWER exceeding pf ,

($

, .75% RTP m

(continued)

WOG STS 3.2-7 Rev1,04/07/95

FNP TS Conversion l Enclosure 5 - JD from STS l Chapter 3.2 - Power Distribution Limits l STS 3.2.1 HEAT FLUX HOT CHANNEL FACTOR (Fn(z))

! FNP ITS 3.2.1 HEAT FLUX HOT CHANNEL FACTOR (Fo(z))

i JD l

! NUMBER JUSTIFICATION AFD must be controlled within the reduced limits. f SRs 3.2.1.1 and 3.2.1.2 are revised to reference the steady state and transient limits for Fn(Z), respectively.

ji 2 The Frequencies of the STS surveillances SR 3.2.1.1 and SR 3.2.1.2 are revised consistent with the FNP licensing basis as specified in the current FNP TS. The j

,n-current FNP TS is NRC approved and based on the currently accepted methodology p9

~

for verifying Fn. The proposed revisions consist of the following:

l The STS restriction requiring performance of these surveillances'within 12 l

l hours after achieving equilibrium conditions after a specified power increase i is deleted. The current FNP TS do not contain this 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months restriction.

This is reasonable considering that as power increases Fq generally decreases and during power escalation the plant is required to be controlled within the TS limits for rod alignment, rod insertion, AFD, and QPTR.

These additional TS limits provide reasonable assurance that the core is being operated within the design limits. Therefore, specifying a 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months time limit for the performance of this surveillance is an unnecessary restriction. l The STS specifies that these surveillances must be performed after a power increase of 210% RTP since the last time Fn(Z) was verified. The specified power level is revised to 20% consistent with the requirement of the current FNP TS. Although it is prudent to verify Fo(Z) within the required limit after a power increase, the exact magnitude of the power change at which l

Fn(Z) should be re-verified is not a specific assumption of any analyses.

The current practice is to perform the re-verification after a 20% power increase. This provides more operating flexibility than the STS 10% j requirement and has proven sufficient to ensure Fn(Z) remains within the

=

required limits. The STS requirement to re-measure Fn(Z) after each 10 %

l power increase may result in additional and unnecessary power distribution measurements.

3 The STS SR 3.2.1.2 Note is revised to incorporate FNP specific terminology and location for the penalty factor in the COLR consistent with the current FNP TS.

The Note to SR 3.2.1.2 is revised as follows:

The words "Fo*(Z) is within limits and" are deleted on the basis that Fo(Z)

Chapter 3.2 ES-3-A April,1999 m

I2p F.(Z) (F. Methodolnav)

B 3.2 3 BASES ACTIONS LJ (continued)

If Required Actions A.1 through A.4 or 8.1 are not met within their associated Completion Times, the plant must be placed in a mode or condition in which the LCO requirements are not applicable. This is done by placing the plant in at least MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

This allowed Completion Time is reasonable based on operating experience regarding the amount of time it takes to reach MODE 2 from full power operation in an orderly manner and without challenging plant systems. :

~

SURVEILLANCE SR3.271.1andSR3.2.1.2aremodifiedbyaNote. The

' REQUIREMENTS Note applies during the first power ascension after a d. i

. refueling. It stai.es that THERMAL POWER may be iner \SV until an equilibrium power level has been ach at which ils

,' a power distribution map can obtained. isJ 11owance is

. c. ,

modified, however, by one o e Fruuenc cogfitions_tha

.:. requires verification that Z) Canc c m ars7within k *; specified limits after a power rixof mor tha TP _ 20'4

'e tne inuamL POWER at whicirthey war ast itcFto be within specified limits. Because E(Z) and' Z) could

. not have previously been measured in this reload core, there f$ g/

is a second Frequency. condition, applicable only,for reload (gd before exceeding 7J5 %TP.wres, that requi 15GL

'P Th 3 ,

~

determination'of,IWZ)(ind F;(H) arWaade at a lower power level at which adequate margin is available before going to s10 E R_TP. Also, this Frequency condition, together witlythe OSU J F quency condition requiring verification of(FMZ) anmP"

. Z) followiryn power increase of more than (,05',rensuC 2O o

. tha hay aryverified as soon as RTP (or any oyer level SC C or xtended operation) is rehiived. In the absence of these Frequency conditions, it is possible to increase p r j to RTP anybperate for 31 days without verification of Z)

,L (nd F;(ZY.' The Frequency candition is not intended to goN r utre verification of these parameters after every increase in power level above the last verification. It y requi s verification after power level is achieved for exte ed operation that i higher than that power at which as last measured. s 20 %

4% &\d b (continued)

WOG STS B 3.2-16 Rev 1, 04/07/95 0

l

Q 12i 2.0

))w a 4)((h ,

n F (Z)(F. Meth dol y

+ke, Q le orwareeI ec+ees ha o cL peuaer d.ssfrebuf,'on me.P a.k, Tl4ERMAL PDu)ER. I BASES Leve,ls, deader th 5*/, RTP, p ~

SURVEILLAN SR 3 2.1.1 ,

REQUIREMENTS -- 2>

Verification Q.at(7 4Y F. Z) is within itscanectrieflimits [g 5 % in Ives inctLeasing F;(Z) o allow for manufacturing SP8' '

ance andvm easurement uncertainties in ord to obtain M 9M } ). Specifically, F; ) is the measured v l

CT5 4.2.2.2.h 1 i tained rom incore sap results and Z) = F;(Z)ue of F,(Z)[ g?

N*

d./1d.,t $

L(

M l1.081 (Ref. 4). ) is then compared to itsiipectrieg lxshle .m cor e (l'he limit with which Fj(Z) is compared varies inversely with l er loMS, .I power above 50% RTP and directly with aJaction .r.alled K(Z) '

, , , trovided_intheCOLR.f , ,

sne.Io, dive.

Performing this Surveillan in MODE 1' prior to exceeding;

~

I ensures that the Z) limit is met when RTP is 75%

eun eveRT)d, because peaking factors generally se as I power level is increased.

5 If THERMAL POWER has be inc eased by a: RTP since the N-@

Sy last determination of Z) _nother eval ion of this factor is requiredElz. nours after achieving equilibri I conditions at this mgiier power level (to ensure that Z) '

crs 4.7.2.2.c values are being reduced sufficiently with power increase to

= .--- stay within the LC0 limits).

%5 SM '8 The Frequency of 31 EFPD is adequate to monitor the change performecL use? of power distribution with core burnup because such' changes ne move.kle i re are slow and well controlled when the plant is operated in defeciers to e6tm.in accordance with the Technical Specifications (TS).

G pmerelesfrebufM Me.p 's.t TA/K4/M L / 1 Power t e.veJs /mSR 3.2.1.2 ,

TN" ' ffhe nuclear design process includes calculations performe to determine that the core can be operated within the 5'/.RTP. F,(Z) limits. Because flux maps are taken in steady state conditions, the variations in power distribution resulting from normal operational maneuvers are not present in the flux map data. These variations are, however,

3 g .4. g ' conservatively calculated by considering a wide range of 9 unit maneuvers in normal operation. The maximum peaking f factor increase over steady state values, calculated as a 5 , function of core elevation, Z, is called W(Z). Multiplying pe measured total peaking factor, Fj(Z), by W(Z) gives the)

(continued)

WOG STS -

B 3.2-17 Rev 1, 04/07/95

4'gi .,

F (Z) (F. Methodology)b

$i -

8 3.Z.g l BASES -

SURVEILLANCE SR 3.2.1.2 (continued)

REQUIREMENTS . , _ . ..

Performing the Surveillance in MODE 1/ prior to exceeding 75% RT} ensures that the F (Z) limit is met when RTP is hgv#3p,;8d m.meved, because peaking factors are generally decreased as

, power level is increased.'.- 2.0%

F.(Z) is verified at power levels a P above e THERMAL POWER of its last verification,(112] hourl after achieving equilibrium conditions to ensure that F.(Z) is within its limi higher power levels.

The Surveillance trequency of 31 EFPD is adequate to monitorl the change of power distribution with core burnup. The ..

Surveillance may be done more frequently if required by the 6

dsults of F.(Z) evaluations.f a yaoi\\n+d The requency of 31 EFPD is adequate to monitor the change '

, of _ power distribution,.because such a change is sufficiently slow, when the plant is operated in accordance with the TS, gAkV4 to preclude adverse peaking factors between 31 day surveillanc y 1

REFERENCE 1. 10 CFR 50.46,

2. disgulatory Guide 1.77, Rev. O, May 1974 F5AR,, 15.% 4.

$ 3. 10 CFR 50, Appendix A, GDC 26.

4. WCAP-7308-L-P-A, " Evaluation of Nuclear Hot Channel

((

\ /7 Factor Uncertainties," June 1988. ,

s. AP -lout,-P-A hn sA "Rel nbn af- @M i i

p \ c M 5et c d i g s m .,h c.e 7tekoa, g '

p c. h 4,iml % v q te+,

b; e osvit of 4%s Sv n eill e c t c. m e s w H i n l S oce 4 %w m & s w e h ce d Fq ( O 'S nu. m q , l WOG STS B 3.2-19 Rev 1, 04/07/95 1

i 1

L i

I Associated Package Changes for RAI- 3.2.4-1 .

9 O

i I

i l

l

'1 i

i j

l t

I l

l FNP TS Conversion Enclosure 2 - Discussion of Changes to CTS I Chapter 3.2 - Power Distribution Limits ;

l CTS 3/4.2.4 QUADRANT POWER TILT RATIO l FNP ITS 3.2.4 QUADRANT POWER TILT RATIO (QPTR)

DOC N_Q SHE DISCUSSION l repeated aRet the determination of QPTR every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months as necessary. At l reduced power levels, additional changes in QPTR would be slow. The l proposed STS actions address the peaking factors which are the safety ~I analysis parameters of prime concern. The limit on power required by A.1

provides additional margin below the fuel design limits to help ensure those limits are not challenged by local power peaking during conditions when a radial power tilt is indicated. The new STS Action A.3 provides additional assurance that power peaking is within design limits by specifying the performance of SRs 3.2.1.1 and 3.2.2.1 which verify the pamWg factors are j within the required limits within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after achieving equilibrium conditions with thermal power limited by required action A.I. Therefore, the proposed FNP ITS action requirements provide sufficient operating restrictions when the QPTR limit specified in the LCO is exceeded. In addition, the CTS requirement to calculate QPTR once per hour is I excessive and would potentially divert the attention of the control room staff from working to correct the cause of the out-of-limit condition.

4 A The CTS 3/4.2.4 action a.2.a is deleted consistent with the STS. In the STS, options to restore a system or parameter are not typically stated in the Actions. It is understood that restoration to meet the LCO requirements is always an option and need not be specifically stated as an altemative to l reducing power. As such, this change is considered an administrative change made to conform with the format and presentation of the STS. I 4a A The CTS 3/4.2.4 action a.2.b is revised by replacing the word "Reddce" with " Limit." Upon startup, QPTR may be in excess of 1.02 because of transient core conditions. These transient conditions are usually self-correcting as the direct result of power ascension. Changing the language of action a.2.b to require Limiting thennal power rather than reducing Q thermal power ensures that power is limited below the appropriate level and g

,V helps to avoid the confusion which could be introduced by the word reduce ,

y with respect to this action. As this change continues to limit power below RTP based on QPTR, there is no technical change to the requirements of this LCO, and the change merely clarifies the intent, this is seen as an administrative change.

5 L The CTS 3/4.2.4 action a.2.b is revised consistent with the STS. The requirement to reduce the power range neutron flux high trip setpoint within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months of reducing power 3% for each 1% QPTR exceeds 1.00 is deleted.

Chapter 3.2 E2-2-D April,1999

FNP TS Conversion Enclosure 5 - JD from STS Chapter 3.2 - Power Distribution Limits STS 3.2.4 QPTR FNP ITS 3.2.4 QPTR JD N11MBER JUSTIFICATION 1 . The wording of FNP ITS LCO 3.2.4 Required Action A.1 is revised to " Limit THERMAL POWER to 2 3% below RTP for each 1 % of QPTR > 1.00."

Changing the language of Required Action A.1 to require limiting thermal power p'[ rather than reducing thermal power clarifies the inte g minimizing the potential confusion related to the actions required to be taken when C this condition is entered.

la The STS 3.2.4 QPTR Mode of Applicability is revised to be consistent with the Mode of Applicability for the Axial Flux Difference LCO. The QPTR Applicability is revised from "> 50% RTP" to "2 50% RTP". Although this change introduces a slightly more conservative Applicability for the QPTR LCO, it maintains the CTS consistency between the AFD and QPTR Applicabilities. This change has a minimal impact on plant operation and eliminates the introduction of subtle differences between two LCO Applicabilities that were previously the same.

2 Required Actions A.1 and A.2 are revised to clearly identify the determination of QPTR and the requirement to limit power after the determination has been completed. The second portion of the existing Action A.2 and the existing Action A.1 are the same. Since the existing Action A.2 could require a power reduction below the limit required by existing Action A.1, references to the reduced power for Required Actions A.4, A.5 and A.6 would not reflect the additional power reductions required by existing Action A.2. Revising Required Actions A.1 and A.2 ensure that subsequent Actions are based on the most limiting power based on QPTR determinations. In revision 0 of NUREG-1431 (STS), actions A.1 and A.2 appeared one action (A.1) with a compound Completion Time. In revision 1 of the STS, two actions were used (A.1 and A.2) to express the requirements for limiting power relative to the amount of core tilt. However, in revision 1 the Completion Times of actions A.4, A.5, and A.6 which referenced the limitation on power established by action A.1 were not revised to reflect the addition of action A.2.

This was an oversight. A revision to the STS was transmitted to the NEI/ Industry Technical Specification working group on 5/15/96 by the Westinghouse Owners Group as a recommended editorial change to the STS (no TSTF number assigned for editorial changes). This change agrees with proposed WOG-95.

N 3 Required Action A.3 is revised to require the peaking factors Fn(Z) and F ,s (SRs 3.2.1.1 and 3.2.2.1) to be verified within limits 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after achieving equilibrium conditionsivith thermal power limited by Required Action A.1 and once per 7 days thereafter. Required Action A.5 is revised to require the peaking factors Fo(Z) and Chapter 3.2 E5-1-B April,1999

r Associated Package Changes for Chapter 3.5 I

f I

I e

l

.1 l

j l

l l

. I l

l r

E_. ..

4 i

Associated Package Changes for RAI- 3.5.1-1 i

i 1

i I

\

i i

t l

i 1

I i

i i

i l

l l

1 I

)

FNP TS Conversion Enclosure 2 - Discussion of Changes to CTS Chapter 3.5 - ECCS CTS 3/4.5.1 ACCUMULATORS FNP ITS 3.5.1 ACCUMULATORS DOC NQ SHE DISCUSSION removed. Therefore, the TS requirement to have the automatic actuation of these valves operable during thc Mode of applicability (when power is also required to be removed) can not be met. Based on the CTS and FNP ITS surveillance requirements to verify these valves open with power removed, the safety function of these valves above the P-11 interlock is assured. In addition, the SI actuation of these valves is verified during other required surveillances which perform SI signal component actuation testing (master and slave relay testing required in the ESFAS LCO for the SI function as'

. well as integrated ESF actuation testing required .in TS section 3.8) and will continue to verify the capability of these valves to respond to an SI signal every 18 months. As such, the verification of the automatic actuation of these valves is not required in the accumulator TS to preserve the assumptions of the safety analyses and have been deleted. This change is consistent with the Accumulator requirements in the STS.

12 M In the conversion to ffS, a note is added which does not exist in CTS 3.5.1.

This note is necessitated by the adoption of the 1000 psig Applicability limit for the accumulator LCO consistent with the STS. The 1000 psig applicability limit has also been adopted to confonn to the assumptions of f ,Y[

g WCAP-12476, " Evaluation of LOCA during Mode 3 and Mode 4 Operation for Westinghouse NSSS," which is applicable to FNP. The new note, which allows testing of the accumulator isolation valves in Mode 3, with RCS pressure above 1000 psig, is necessary to allow the performance of the required RCS PIV testing consistent with current FNP practice. Thi.snote is unnecessary in the CTS since the testing is performed prior to the current accumulator LCO applicability limit of P-11 (or 2000 psig). Although this note is based on current FNP practice, the addition of this note adds a restriction to the testing of the RCS PIVs that does not currently exist and therefore is. considered a more restrictive change. .

13 A In the STS, SR 3.5.1.4 directs the verification of boron concentration in an accumulator based on a change in indicated level. . CTS surveillance 4.5.1.1.b uses percent of tank volume. To conform with the STS and to enhance the operators ability to easily determine when to verify accumulator boron concentration, a calcalation was performed to determine what percent ofindicated level corresponded to a 1% change in tank volume. A change of 1% tank volume conesponds to a 13.3% indicated level change. Since the MCB indicator is divided into 2 % increments,12%

Chapter 3.5 E2-5-A April,1999

hh ,

Accumulators 3.5.1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) ,

' Three 3.5.1 Accumulators A

LCO 3.5.1 ((Four] ECCS accumulators shall be OPERABLE.

[ 4,d e 3 -

c,5 3

g f}t 0_00 p_

~ t w c c w m u m m ,,x a .c e -- t ACTIONS ( F_b__P"5T YT2 "9'_DY_"2'*_'3'E _ .

%wv --

CONDITION REQUIRED ACTION COMPLETION TINE A. One' accumulator A.1 Restore boron 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable due to concentration to boron concentration within limits.

not within limits.

B. One accumulator B.1 Restore accumulator I hour

inoperable for reasons to OPERABLE status.

other than '

' Condition A.

Required Action and 3

C. C.1 Bs in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A Afgl ,

8,CS or B not met. ~

C.2 Reds erassuririe [ 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months pre sure/to .

. s 1000F9519 D. Two or more D.1 Enter LCO 3.0.3. Immediately .

accumulators inoperable.

WOG STS 3. 5-1_ Rev 1, 04/07/95

FNP TS Conversion Enclosure 5 - JD from STS _ !

Chapter 3.5 - ECCS STS 3.5.1 ACCUMULATORS FNP ITS 3.5.1 ACCUMULATORS JD NUMBER JUSTIFICATION 1 An FNP specific note is added which is necessitated by the adoption of the 1000 psig applicability limit. The 1000 psig applicability limit has been adopted to conform to the assumptions of WCAP-12476," Evaluation of LOCA during Mode 3 and Mode 4 Operation for Westinghouse NSSS," which is applicable to FNP. The CTS Applicability of the Accumulator LCO in Mode 3 was P-11 or 2,000 psig.

Under the CTS requi:ement, routine RCS PIV leakage testing was performed on the gf g accumulator check valves which required the accumulator isolation valves to be h5.F closed between 1,000 and 2,000 psig. The proposed FNP specific note modifying the Applicability of the Accumulators allows testing of the accumulator isolati6n l valves in Mode 3, with RCS pressure above 1000 psig, to continue to be performed in the same manner as under the CTS. This note, including the time limitation, is consistent with the current RCS PIV leakage testing performed at FNP but represents a restriction on the perfonnance of this testing that was not in the CTS.

2 In the STS, SR 3.5.1.4 directs the verification of boron concentration in an accumulator based on a change in indicated level. CTS surveillance 4.5.1.1.b uses percent of tank volume. To conform with the STS and to enhance the operators ability to easily determine when to verify accumulator boron concentration, a calculation was performed to determine what percent ofindicated level corresponded to a 1% change in tank volume. A change of 1% tank volume corresponds to a 13.3% indicated level change. Since the MCB indicator is divided into 2 % increments,12% indicated level change is used as the level at which to verify accumulator boron concentration. This number is slightly more conservative than the current 1% of tank volume. Since the CTS is based on a change in tank volume and the number in the ITS markup is based on that change, the wording was revised to read as follows: "Once per 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after each solution volume increase of > or = 12 % level, indicated, that is the result of addition from the refueling water storage tank." to clarify that it is speaking of an indicated level change of 12% vs. a change of 12% of the previously indicated level.

3 The references to pressurizer pressure in LCO 3.5.1 are revised to indicate RCS pressure. This deviation to the STS is required because the FNP pressurizer -

pressure narrow range instrumentation range is 1700 to 2500 psig. The RCS wide-range 0 to 3000 psig pressure instruments are used for monitoring system pressure below approximately 1700 psig. This change makes the FNP ITS terminology consistent with the plant design.

Chapter 3.5 - E5-1-A April,1999

1G CHAPTER 3.5 INSERT K TO STS BASES PAGE B 3.5-5 FNP SPECIFIC BASES FOR THE NOTE MODIFYING THE AtCUMULATORLCO APPLICABILITY g4F \

f The Accumulator Applicability is modified by a note which takes exception to the LCO

[ requirements for the Accumulators to be OPERABLE in MODE 3, with RCS pressure above 1,000 psig for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during the performance ofisolation valve testing required by SR 3.4.14.1. The applicability of the Note is restricted solely to the isolation valve testing required by SR 3.4.14.1. In order to perform the required isolation valve testing, the Accumulators must i l be isolated. The exception provided by this Note allows operation in MODE 3 with RCS pressure above 1,000 psig for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,with Accumulators isolated such that the. Actions

~

for an inoperable Accumulator are not applichble. !

-. l 1

l l

l Chapter 3.5 Insert Page

m Accumulators m wm u , can u . ,pu

' M *-

BASES

~

ACTIONS M - (continued) 0, reduced. Jhe ron ip the cumulato s con butes t the' f) $@ rassump ion th t the obabine ECCS wa r in e partia ly pvA* gh recov red co duri ithe rly ref1 oding ase of large "

brea LOCA i suffi ent t keep th t nortion of th cores subc itical J One accumulator below the minimum boron

.M W concentration limit, however,,jwill have nq effect on available

'ECCS water and an_ insignificant effect on rcore a suberiticality 4, d r':1 r:eed. fbo 11ng o tccs wa er in s core o.srtn Erfr1 nc trat s .o ...in the. tur_ ate. 1.1. uid iat sac.-sw2m-c.) /e ns n t e cb / n ddition curren ana ysi tec iq s de nst ate that~ he acc ulator do not ischarg fol wi g'a 1 as ain stea line b eak for he mai ity of ant . ve '

kttheydodischarge.) heir impact is minor and not a desigii

1miting event. Inus, 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is allowed to return the boron concentration to withi limits.

%s by mA sb.m m W Je

.~ .y &b w + 4 4h 6 gel

U $40 e5' b(pD* #*"*M Ati. If one accumula r is inoperable for a reason other than boron C,r# concentration, he accumulator must be returned to OPERABLE ,

status within hour. In this condition, the required contentsofqir accumulators cannot be assumed to reach the l core during a A. Due to the severity of the consequences should a LOCA occur in these conditions, the I hour Completion Time to open the valve, remove power to the valve, or restore the proper water volume or nitrogen cover pressure ensures that prompt action will be taken to return the inoperable accumulator to OPERA 8LE status. The Completion Time minimizes the potential for exposure of the plant to a LOCA under these conditions. ._

C.1 and C.2 r If the accumulator cannot be returned to OPERABLE status

. within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant apt be brought to MODE 3 within .I 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and eressurizerrpressure reduced to 3h (continued)

WOG STS 8 3.5-6 Rev1,04/07/95

Associated Package Changes for RAI ~ 3.5.12 I

1 i

1

la O Accumulators y 8 3.5.1 BASES

[5 ^

SURVEILLANCE REQUIREMENTS SR 3.5.1.4 aN~Mb D*

W4 ~

(continued) The boron concentration should be verified to be wititin required limits for each accumulator every 31 days sin the static design of the accumulators limits the ways in ich the concentration can be changed. The 31 day Frequency is adequate to identify changes that could occur from methanisms p, 1 I such as stratification or inleakage. Samp1 he af'ected Me >l accumulator within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after acIE voi crease (vill #

identify whether inleakage has caused a reduc ton in baron

~

p tow concentration to below the required limit. It is not necessary to verify boron concentration if the-added water inventory is from the refueling water storage tank (RWST),

h heeewse the water contained in the RWST is within the _

accumulator boron concentration requirements. This is consistent with the recommendation of NUREG-1366 (Ref. .

SR 3.5.1.5 Verification every 31 days that power is removed from each

@ accumulator isolation valve operator when the pressurizer pressure is a: 2000 psig ensures that an active failure could on g' gheuh notoperated result isol,afion in the undetected closure valve. If this were of an to occur, only tor accumulator wo e

Isog h ,r o

accumulatorfwould be available or injection given a s ng e

.failurecoincidentwithaLOCAgSincepowerisremovedunder

d. %ah fa'diiii'nistrative control, the 31 day Frequency will provide o.

l 60*!adeq;uateassurancethatpowerisremoved.

pg.onnect ?

This SR allows power to be supplied to the motor operated isolation valves when pressurizer pressure is < 2000 psig, thus allowing operational flexibility by avoiding unnecessary delays to manipulate the breakers durino plant startuos of shutdowns. IEven with power suppiled to tse vaives, 7 1nauvertent closure is prevented by the RCS pressure interlock)

% associated with the valvesd P Should closure of a valve occurEn spite of the interlocl the SI signal provided to the valvervould open a closed valve in the event of a LOCA. -

belous zooof*I r

w ,

2 . ..

.. ., , ;. 9 (continued)

WOG STS B 3.5-8 Rev 1, 04/07/95 W.

Associated Package Changes for RAI- 3.5.2-1 e

a i

1 I

l i

L .. .

m

~. - '

. NerfE - -

_ _ _ _ _ b6 ll

'[ ', 33 (

8132A ECCS - Operat ng

{> ,,.lY'{2g bT ,,

D f A i8 p SbsfRdidE;01=h - -- -- -A t,.I g [ SURVEII. LANCE FREQUENCY R 3.5.2.1 Verify the following valves are in the 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months I listed position with power to the valve operator removed. --

Number Eggigia Function Nh 8884,8886 Closed Charging Pumpto RCS Hot Leg 8132A,8132B Open 4 Charging Pump , I hp isolatio'n ~

l 8889 Closed RHRto RCS Hat 1

I .

14gleiection ,,

l SR 3.' 5. 2. 2 Verify each ECCS manual, power operated, 31 days and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.

S .5.2.3 Verif ECCS p ping is ull of ater. 31 ys 3

SR 3.5.2./ Verify each ECCS pump's developed head at In accordance the test flow point is greater than or with the ,

equal to the required developed head. Inservice . i Testing Progr,am i l

SR 3.5.2 ~ Verify each ECCS automatic valve in the months !

flow path that is not locked, sealed, or i otherwise secured in position, actuates to the correct position on an actual or !

simulated actuation signal.

~

(continued) i l

I WOG STS 3.5-5 Rev 1, 04/07/95 i

l l

l I

i l

l i

Associated Package Changes for RAI- 3.5.2-2 l

l I

l 1

I l

l l

l 1

i

FNP TS Conversion Enclosure 2 - Discussion of Changes to CTS Chapter 3.5 - ECCS CTS 3/4.5.2 ECCS SUBSYSTEMS - Tavg > 350* F FNP ITS 3.5.2 ECCS - OPERATING DOC NQ SHE DISCUSSION performance requirements. The changes in the statement of the surveillance and the placement of details conceming specific test methods and data in the Bases is consistent with the level of detail in the STS surveillances and is acceptable due to the adequate level of change control provided by the

(( bases control program in the administrative controls section of the STS and the 10 CFR 50.59 process for information located in the IST program.

~

I 14a M The FNP specific CTS 4.5.2.h surveillance is combined with CTS 4.5.2.e into one ITS surveillance (SR 3.5.2.6) for ECCS valve position stop verification. 'Ihe CTS requires that the mechanical stops on the low head safetyinjection valves RHR-HV 603 A/B be verified to be intact. In conversion to the ITS, the STS requirement to verify that the stops for these valves are in the correct position is adopted. ' As this adds a new requirement that doesn't currently exist in the CTS, it is seen as a more restrictive change.

15 LA The FNP specific CTS 4.5.2.h surveillance is combined with CTS 4.5.2.e into one ITS surveillance (SR 3.5.2.6) for ECCS valve position stop

. verification. In the CTS, the verification of the mechanical stops installed l on the RHR valves was covered under a separate surveillance due to the differences between the RHR valves and stops and the other ECCS valves with mechanical stops installed. In converting to the STS, the existing details of the CTS surveillances have been moved into the TRM (see .

discussion for CTS 4.5.2.e.l.above). In addition, the requirement in.4.5.2.h l for the RHR valve mechanical stop verification " Prior to entry into Mode 3 from Mode 4" (which verifies the mechanical stop is intact after the RHR valves have been utilized during a shutdown for decay heat removal) is moved into the TRM along with the other post stroke operation position stop verification requirements for ECCS valves in 4.5.2.e.l. The requirement for post valve stroke mechanical stop verifications is similar to post maintenance test requirements which are no longer included in the STS and are considered inherent in the requirements necessary to maintain l equipment operability as defined in the TS. The necessity to specify two surveillances for ECCS valve mechanical stop verification is eliminated as discussed in DOC 14a-M and CTS 4.5.2.e.1 and CTS 4.5.2.h are combined with one 18 month frequency required in the TS (SR 3.5.2.6) and post maintenance and valve stroke related requirements maintained outside of the TS (in the TRM). As such, the specific requirements for the RHR Chapter 3.5 E2-6-B April,1999

F ,

FNP TS Conversion Enclosure 2 - Discussion of Changes to CTS Chapter 3.5 - ECCS E CTS 3/4.5.2 ECCS SUBSYSTEMS - Tavg > 350 F I$ FNP ITS 3.5.2 ECCS - OPERATING DOC N_,Q SHE DISCUSSION valves and other ECCS valves are addressed in the TRM. The placement of this information in the TRM is consistent with the level of detail in the STS surveillances and is acceptable due to the adequate level of change control l provided by the 10 CFR 50.59 process for the TRM.

16 LA CTS 4.5.2.i requires an ECCS flow balance test to be performed after any modifications that alter the subsystem flow characteristics. This CTS is~ a post maintenance test reqmrement. In the STS, post maintenance test requirements are not specifically included in the TS. The bases for removing post maintenance test requirements are the general rules of TS and the TS definition of operability. Since the capability of the subsystems to inject specific flows is an assumption of the safety analysis and an inherent operability requirement, this injection capability must be met whenever equipment is retumed to service and declared operable to meet the associated LCO. Therefore, specific TS requirements requiring surveillances to be performed or met after maintenance are unnecessary as all applicable surveillances (as well as any other operability requirements of

~

4 the system / component) must be met in order to meet the associated LCO. l Therefore, the post maintenance test requirement is moved into the TRM along with the specific flows required for the ECCS subsystems for this 4 surveillance. The retention of these requirements in the TRM is acceptable i as an adequate level of change control will be provided by the 10CFR 50.59 process for the TRM.

i 1

i

. chaear 3.5 E2-7-B April,1999 1

j

IFA -v w ECCS - Operating (Csn t'l a)Clnt '

BASES 4 S' SURVEILLANCE SR 3.5.2.5 and SR 3.5.2.I (continued)

REQUIREMENTS simulated SI signal and that each ECCS pum[ starts on receipt of an actual or simulated SI signal. This Surveillance is not required for valves that are. locked, sealed, or otherwise secured in the required position under administrative controls. The 18 month Frequency is based on the need to perform these Surveillances under the conditions that apply during a plant outage and the potential for unplanned plant transients if the Surveillances were performed with the reactor ,at power. The 18 month Frequency is also acceptable based orr consideration of the design reliability (and confirming operating experience) of the l equipment. The actuation logic is tested as part of ESF ..

l Actuation System testing, and equipment performance is monitored as part of the Inservice Testing Program.

(, h __ - - . - - --

g. ;

SR 3.5.2./ [ c .

.} f* i,

& Realig of vai m m me r iow pasn on an,SI signal is 4/ necess y for proper ECCS performanceJ These v v s have g.4 stops to allow proper positioning fortestric low to a ruptured cold leg, ensuring that the other cold legs race.ive at least the required minimum flowt t/m1f survy11arp rnou reguir.an or piants wizn fler y shf une orificeso The OPsc T8 montn trequency s based on the saan reasons as ; hose stated in SR 3.5.2.J and SR 3.5.2. ws = - -g ,

i The. /efuered,ver'i6ca.kio n 7 PSE fev%e:RHR.wJvessfec3 (l 3 i SR 3.5.2.f I a5w<is tuk 4W MS*c

( tafjM od be evook.

Periodic inspections of the containsienE suirp suction Inlet-ensure that it is unrestricted and stays in proper operaiting ;

condition. The 18 month Frequency is based on the need to \

perform this Surve111anceumder the conditions that ap>1y i during a plant outace,6iLthe need to have accans to tie location efing becausf o* tie /potentAal for an/unplanrfe< !

prangient if/the Suseilla fe were/perfonnedArith t%)

@ treacjor at dower s This Froquency nas oeen rouna to De surricient to detect abnornal degradation and is confirmed by operating experience. --

For eSer Ecd 5 Oud. [ees the /=h i

l devsCe IS VSfi j sn +ke.cerreekpaseto.as.

(continued)

WOG STS B 3.5-19 Rev 1, 04/07/95 i J

Associated Package Changes for RAI- 3.53-3 i I

1 a

l i

l l

1 I

l l

4 l

l i

i i

FNP TS Conversion Enclosure 2 - Discussion of Changes to CTS Chapter 3.5 - ECCS CTS 3/4.5.2 ECCS SUBSYSTEMS - Tavg > 350 F FNP ITS 3.5.2 ECCS - OPERATING DOC HQ SHE DISCUSSION valves and other ECCS valves are addressed in the TRM. The placement of this information in the TRM is consistent with the level of detail in the STS surveillances and is acceptable due to the adequate level of change control provided by the 10 CFR 50.59 process for the TRM.

16 LA CTS 4.5.2.i requires an ECCS flow balance test to be performed after any modifications that alter the subsystem flow characteristics. This CTS is, a post maintenance test requirement. In the STS, post maintenance test requirements are not specifically included in the TS. The bases for removing post maintenance test requirements are the general rules of TS and the TS definition of operability. Since the espability of the subsystems to inject specific flows is an assumption of the safety analysis and an inherent operability requirement, this injection capability must be met whenever equipment is returned to service and declared operable to meet 4,4[ the associated LCO. Therefore, specific TS requirements requiring surveillances to be performed or met after maintenance are unnecessary as all applicable surveillances (as well as any other operability requirements of the system / component) must be met in order to meet the associated LCO.

Therefore, the post maintenance test requirement is moved into the TRM i along with the specific flows required for the ECCS subsystems for this surveillance. The retention of these requirements in the TRM is acceptable as an adequate level of change control will be provided by the 10CFR 50.59 process for the TRM.

l l

l Chapter 3.5 E2-7-B April,1999 j

I96 ECCS - Operating B 3.5.2 BASES APPLICABLE The effects on containment mass and energy releases are SAFETY ANALYSES accounted for in appropriate analyses (Refs. 3 and 4). The (continued) LCO ensures that an ECCS train will deliver sufficient water to match boiloff rates soon enough to minimize the consequences of the core beine uncovered following a larae LOCA Iso sures that t con ifugay charging d gp / del er s ficie wate and h6ron durin a all 9 gb lg i

\i t main in_c e su

~

itica' tv.I For saaer t.0CAs r t:1e centrifugal charging pump de' fvers sufficient fluid to ep/g maintain RCS inventory. For a small break LOCA, the steam

, generators continue to serve as the heat sink, providing part of the required c' ore cooling.

The ECCS trains satisfy criterion 3 of the NRC Policy j Statement.

M LCO N00ES 1, 2, and 3, two independent (and redundant) ECCS K trains are required to ensure that sufficient ECCS flow is gp mp available, assuming a single failure affecting either train. 4 3

must i F Mf.t,

. j Additionally, individual components within the ECCS trains gem an.r h un g may be called upon to mitigate the consequences of other i pump mustin ed g 39o2 3p These, i tr Islefits artd accidelits. h M,2-3

-V W3 m p g ,

d[' l i In nuuES 1, f,7 nd T, an ECCS train consists of f;' .

)% widthe sm R een ST o w.h o n centrifugal charging subsystem,41SL subsystem 7 and an RHR subsystem. Each train includes t te piping, instruments, and

.Jp i mmi m veron .-

1 CM controls to ensure an OPERABLE flow path capable of takin

$$$gmde , .,, q suction from the RWST upon an SI signal and(automatica11yf -

transferring' suction to the rontainment sump.

re'it h us e.

%3-AdGec.5 -

L+brch izmee criferw, During an event requiring ctcs ctuation, a flow path is, o^a. *s Fede, required to provide an abunda supply of water from the RWST to the RCS via the ECCS umps.a_nd their respective supply headers to each of theuou cold leg injection

'@SE .

nozzles.

In the long ters, this flow path may ba switched to take its supply from the containment sump and to supply its flow to the RCS hot and cold legs.

/k,[ The flow pathtofor independence each ensure thattrain must no single maintain failure its designed can disable jP both ECCS trains. ,

@NSER.T TEXT trkLr p - -

~

T5TF-IS3 D '

k m t. 5, .)G' .

(continued)

WOG STS B 3.5-14 Rev 1, 04/07/95 a !

,1, . .

r 1

i i

l Associated Package Changes for RAI-3.5.31 I

l

FNP TS Conversion Enclosure 2 - Discussion of Changes to CTS Chapter 3.5 - ECCS ,

4 CTS 3/4.5.3 ECCS SUBSYSTEMS - Tavg < 350* F i

FNP ITS 3.5.3 ECCS - SHUTDOWN DOC HQ M DISCUSSION )

of the ECCS flow, equivalent to a single operable ECCS subsystem. At FNP, the requirement to ensure that only one centrifugal charging pump is operable for overpressurization concerns is not applicable until the temperature of one or more of the RCS cold legs is less than or equal to 180 degrees F. Therefore, in MODE 4, two or more centrifugal charging pumps may be available. ' As stated in the Bases for LCO 3.5.2 in the STS, an '

ECCS train is inoperable ifit is not capable of delivering design flow to.the RCS. Individual components are inoperable if they are not capable of performing their design function or supporting systems are not available. 1 I

The intent of this Condition is to maintain a combination of equipment such that 100% of the ECCS flow equivalent to a single OPERABLE ECCS train remains available, so that it can perform its design function. Since the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is acceptable when the unit is in MODES 1,2, and 3 (Based on NRC Memorandum to V. Stello, Jr., from R. L. Baer,

" Recommended Interim Revisions to LCOs for ECCS Components,"

December 1,1995) and MODE 4 represents less severe conditions for the initiation of a LOCA, the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is also acceptable for MODE 4 provided that the remaining capacity of the operable components is equivalent to one operable ECCS subsystem or 100% of the required flow.

5 L The CTS action statement "a" time allowed to be in Mode 5 is revised from 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months consistent with the STS. The additiornal 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is reasonable considering the reduced pressure and temperature conditions of the RCS in Mode 4, the time it takes to shift RCS cooling to the RHR system, and the reduced rate of heat transfer and resulting slower cooldown at lower RCS temperatures. The additional time allows a more controlled plant tmnsient to Mode 5. In addition, a completion time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is a s sufficient restriction, considering the small likelihood of a severe transient g

g.

occurring during this time, to avoid an undue risk to public health and safety.

f6 A The statement in CTS action statement "a" that modifies the requirement to place the unit in Mode 5 is revised consistent with similar information presented in the STS and included as a second requirement in the statement of the FNP ITS Condition D. The CTS statement "if at least one RHR loop is operable" ensures that the required RHR system cooling capacity exists to support safely placing the unit in Mode 5 and to support continued safe Chapter 3.5 E2-2-C April,1999

FNP TS Conversion Enclosure 2 - Discussion of Changes to CTS Chapter 3.5 - ECCS CTS 3/4.5.3 ECCS SUBSYSTEMS - Tavg < 350 F FNP ITS 3.5.3 ECCS - SHUTDOWN DOC HQ SHE DISCUSSION operation in Mode 5. Since the STS does not have a corresponding provision this FNP specific statement represents a deviation from the STS based on the existing FNP TS and the need to ensure safe and reliable unit operation transiting to and continuing operation in Mode 5. As this change effectively preserves the existing provisions of the FNP TS in a format compatible with the STS, it is considered an administrative change.

7 LA The CTS action statement "b" is revised to be consistent with the STS. 'The action statement becomes Condition A in the FNT ITS and the description of the RHR subsystem (pump and flow path) is mplaced by the term subsystem. The subsystems that comp:ise an ECCS train are described in detail in the bases of LCO 3.5.2. Placement of this information in the bases is acceptable due to the bases control program in the administrative controls section of the STS. This program provides an appropriate level of control for changes to the TS bases.

8 LA The CTS action "b" is revised to be consistent with the STS. The action in the CTS for maintaining the RCS Tavg < 350*F by the use of alternate heat removal methods is moved to the bases discussion for the corresponding STS action A.I. As the primary focus of the STS condition is the restoration of the RHR subsystem, it is appropriate that the CTS action for maintaimng the RCS temperature be moved to the bases consistent with the STS action and bases. Placement of this information in the bases is acceptable due to the bases control program in the administrative controls section of the STS. This program provides an appropriate level of control for changes to the TS bases.

9 M The CTS action statement "b" is revised by additional change:; made to be consistent with the STS. The CTS statement to restore one subsystem to operable status is modified by the STS phrase to " initiate action immediately to restore ...". The STS wording of this action highlights the urgency appropriate for this condition and focuses all action on restoration of the RHR subsystem to opere.ble status. The CTS allowance to use attemate decay heat removat methods is clearly restricted to the time it takes to restore the RHR subsystem to operable status and is no longer an option in the action statement. As the CTS action for the use of alternate decay heat removal methods is removed as an option and the action is revised to begin restoration of the RHR immediately, this change is Chantor 3.5 E2-3-C March,1998 IN

m:

87 ECCs - Snuteo.n 3.5.3 l

3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.3 ECCS - Shutdo.n LCO 3.5.3 One ECCS train shall be OPERABLE.

Q T5W-io,#4v,l l

PPLICA8ILITY: M00E 4. - __-

, xasem- 1 F l ACTIONS I CONDITION p REQUIRED ACTION COMPLETION TIME :

. Required ECCS residual A.1 Initiate action to Immediately heat removal (RNR) restore required ECCS subs st_em inoperable. RHR subsystem to OPERA 8tE status.

_ m 5-ECCS

- fL h1 p

N Restore / equired ECCS 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months v

C' syst [ hign nead)subsyst noperable. ;0 Urt.XABLE status. '

^

e

..,g ,

i Requi Action and 1 8e in MODE 5. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 3* assoc edCompletionJ '

Time ofConditionBy l

~

l g ...,, .

u ..- .

~ .

At \e.44 one MR '

r.

5vb5pb CPERABLIL. r6

l 3 '

h3'9 ,Y \

WOG STS 3.5-7 Rev1,04/07/95

FNP TS Conversion Enclosure 5 - JD from STS Chapter 3.5 - ECCS STS 3.5.3 ECCS - SHUTDOWN FNP ITS 3.5.3 ECCS - SHUTDOWN JD NUMBER JUSTIFICATION 1 A new Note 2 based on the FNP specific

footnote applicable to part d. of the LCO statement is added to the LCO statement in the FNP ITS TS 3.5.3 consistent with the presentation of similar notes in the STS. Addition of this note is based on the FNP specific design and the requirement to address a change in the required state of the power supplies for the RHR discharge to charging pump suction valves 8706A and 8706B when transitioning from Mode 3 into Mode 4. This revision to the STS is consistent with the current FNP TS.

~

2 STS LCO 3.5.3 is revised by the addition of an action which provides an allowed outage time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for the required ECCS centrifugal charging subsystem to be inoperable provided the remaining operable ECCS components are capable of providing 100% of the ECCS flow, equivalent to a single operable ECCS subsystem consistent with a similar allowance provided in STS LCO 3.5.2. At FNP, the requirement to ensure that only one centrifugal charging pump is operable for overpressurization concems is not appliceble until the temperature of one or i more of the RCS cold legs is less than or equal to 180 degrees F. Therefore, in Mode 4, two or more centrifugal charging pumps may be avai!able. As stated in the l Bases for LCO 3.5.2 in the STS, an ECCS train is inoperable ifit is not capable of delivering design flow to the RCS. Individual components are inoperable if they axe not capable of performing their design function or supporting systems are not available. The intent of this Condition is to maintain a combination of equipment such that 100% of the ECCS flow equivslent to a single operable ECCS train remains available, so that it can perform its design function. Since the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is acceptable when the unit is in Modes 1, 2, and 3 (Based on NRC Memorandum to V. Stello, Jr., from R. L. Baer, " Recommended Interim Revisions to LCOs for ECCS Components," December 1,1995) and Mode 4 represents less severe conditions for the initiation of a LOCA, the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is also acceptable for Mode 4 provided that the remaining j capacity of the operable components is equivalent to one operable ECCS subsystem ;

or 100% of the required flow. This change to the STS is made to provide this allowance based on the FNP-specific restriction for centrifugal charging pump

@ operability being applicable below I80 degrees F, and the FNP design allowing for similar flexibility as afforded by STS 3.5.2.

3 The statement of STS Condition D is revised with the addition of a second required 1 Condition consistem with the ccrresponding FNP CTS action statement "a". The FNP CTS action statement "a" contains a statement that modifies the requirement to place the unit in Mode S. The WP CTS statement "if at least one RHR loop is operable" ensures that prior to oeginning a plant transient to Mode 5 the required Chapter 3.5 ES-1-C April,1999

p j

(

l FNP TS Conversion l

! Enclosure 5 - JD from STS Chapter 3.5 - ECCS STS 3.5.3 ECCS - SHUTDOWN FNP ITS 3.5.3 ECCS - SHUTDOWN l' JD i NUMBER JUSTIFICATION RHR system cooling capacity exists to support safely placing the unit in Mode 5 and to support continued safe operation in Mode 5. Since the STS does not have a corresponding provision, this FNP specific statement represents a deviation from l the STS based on the existing FNP TS and the need to ensure safe and reliable unit l operation transiting to and continuing operation in Mode 5. This change to the STS preserves the existing provisions of the FNP TS in a format compatible with the STS. r l

l 4 The STS 3.5.3, ECCS - Shutdown, is revised by the addition of a FNP specific.

l surveillance requirement (proposed SR 3.5.3.2). This SR contains FNP specific Mode 4 valve position verifications consistent with the FNP CTS (4.5.3.2). This l CTS requirement is revised to conform with the presentation of similar information in the STS. Since the STS does not have a corresponding specific Mode 4 surveillance, this FNP specific surveillance represents a deviation from the STS based on the existing FNP TS and FNP design which requires these valves to be verified in the correct position in Mode 4 to ensure the operability of the ECCS subsystems.

l I

l 1

s

. Chapter 3.5 ES-2-C March,1998

i I37 ECCS - S u BASES ACTIONS 1L1.(continued) continue until the inoperable RHR loop components can be restored to operation so that decay heat removal is continuous.

With both RHR pumps and heat exchangers inoperable, it would y be unwise to require the plant to go to MODE 5, where the only available heat removal system is the RHR. Therefore, the appropriate action is to initiate measures to restore p5 N one ECCS RHR subsystem and to continue the actions until the

$ subsystem is restored to OPERA 8LE status.

A With no ECCS %Iah haas subsystem OPERA 8tE, due to the g inoperability of the centrifugal charging pump or flow path from the RWST, the plant is not prepared to provide high l

pressure response to Design Basis Events requiring SI. The gA '

1 ompletion Time to restore at least one ECCS 1 i 0[ ,

ubsystem to OPERABLE status ensures that prompt action i aken to provide the required cooling capacity or to initiate actions to place the plant in MODE $, where an ECCS l

train is not required.

2, .g ,..p ..ie

..~.,.'. , e rC When the Required Actions of.. Condition cannot be completed prodleA M ithin the required Completion Time, a controlled shutdown

. wenty-four hours is a reasonable i p edec[WR contig

,36 te should ime, be based initiat on op at ng experience, to reach MODE 5 in an i

' N. in rderly manner d wi out challenging plant systems or '

l au o erators. - -- 7 l o- tOb .

.rMsERS M M cDe.S en nss4fcLv . +

12 SURVEILLANCE SR 3.5.3.1 REQUIREMENTS The applicable Surveillance descriptions from Bases 3.5.2 apply.T his su is moa m ma Dy a Note t.nat allows an RH (tratn to be considered OPERA 8LE during alignment and i operation for decay heat removal, if capable of being I

Qually realigned (remote or local) to the ECCS mode of

) $T %, far. \

(continued)

WOG STS 8 3.5-23 Rev 1, 04/07/95

l

\h CHAPTER 3.5 l A

b INSERT M l

[3' TO STS BASES PAGE B 3.5-23 ACTION D.1 l

With both RHR subsystems inoperable, it would be unwise to require the plant to go to MODE 5, I where the only available heat removal system is the RHR. Therefore, the appropriate action is to initiate measures to restore at least one ECCS RHR subsystem and to continue the actions until the subsystem is restored to OPERABLE status. Only then would it be safe to go to MODE 5. l l

l l

I 1

l l

. Chapter 3.5 Insert Page

[

l l

r l l

ATTACHMENTIII SNC Identified Editorir.1 Changes :

Associated Package Changes l

l l

f' I

)

i The following change has been made to the Chapter 3.1 submittal to address editorial changes,

! omissions, and inconsistencies in the package:

l l

l 1. STS Bases page B 3.1 14 for ITS LCO 3.1.2, Core Reactivity, is revised to refer to the l "beginning of cycle life (BOL)" consistent with the response to RAI 3.1.3-2.

l

'Ibe following change has been made to the Chapter 3.5 submittal to address editorial changes, l omissions, and inconsistencies in the package:

1. STS Bases Inserts C and D for ITS LCO 3.5.5, Seal Injection Flow, are revised to refer to

" pressurizer pressure" instead of"RCS pressure" since the graph is based on the differential pressure between the chwging header pressure and the pressurizer pressure. ,

l

'Ihe following changes have been made to the Chapter 3.7 submittal to address editorial changes, omissions, and inconsistencies in the package:

1. The heading for the DOCS related to CTS 3/4.7.1.4, Activity, incorrectly referred to ITS 3.7.14. The heading for these DOCS has been corrected to refer to ITS 3.7.16,

2. JD-5 related to ITS 3.7.2, Main Steam Isolation Valves (MSIVs), incorrectly referenced a closure time of 5 seconds for surveillance testing and 7 seconds for ESF response time.

These numbers are changed to 7 and 9 respectively. In addition, JD-5 refers to FSAR Table 7.3.6. 'Ihe corrected reference is 7.3.16.

1

?'

e 1

i I

1 l

l l

1 l

l I

1 l

l l

1 Chapter 3.1 S

I l

t l

w

l58 Core Reactivity B 3.1 BASES m rF- %

APPLICA8LE behavior and the RCS boron concentration requirements for SAFETY ANALYSES

g (continued) reactivity control during fuel depl The comparison between measur a la qfe,V'(80 M-M . itial core h

reactivity provides a normal zation for the calculational models used to predict core activity. f the measured and predicted RCS boron concentrat n dentical core conditions at beginning of c'c1y do not agree, then I the assumptions used .in the relo cycle design analysis or I

L. the calculational models used to nredict soluble boron

% requ nts may not be accurate. If reasonable agreement I n measured and ' predicted cope- reactivity exists at then the prediction may be normalized to the measured wron concentration. Thereafter, any significant deviations in the measured boron concentration from the predicted boron

@/ letdown curve that develop during fuel depletion may be an indication that the ulational model is not adequate for core burnups beyond or that an unexpected change in i core conditions has The normalization of p RC! ' boron concentration to the measured value is typically trformed after reaching RTP following startup from a refuell tage, with the control l rods in their normal position r operation. The ;

normalization is performed at l conditions, so that core reactivity relative to predict alues can be continually monitored and evaluated as core conditions change during the cycle.

Core reactivity satisfies Criterion 2 of the NRC Policy Statement.

LCO Long term core reactivity behavior is a result of the core physics design and cannot be easily controlled once the core design is fixed. During operation, therefore, the LCO can only be ensured through measurement and tracking, and

.3 appropriate actions taken as necessary. Large differences '

Wf between actual and predicted core reactivity may indicate that the assumptions of the DBA and transient analyses are no longer valid, or that the uncertainties in the Nuclear Design Methodology are larger than expected. A limit on the reactivity balance of

15 ok/k has been established based on engineering judgment. A 1% deviation in reactivity from (continued) .

WOG STS B 3.1-14 Rev 1, 04/07/95

6 Chapter 3.5

(' I CHAPTER 3.5 /hj INSERT C i TO LCO 3.5.5 BASES established by adjusting the reactor coolant pump seal injection needle valves to provide a total seal injection flow in the Acceptable Region of Figure B 3.5.5-1 at a given pressur- differential

\ between the charging header pressure and the pressurizer pressure.

Eok&*.\

1 I

i I

Chapter 3.5 Insert Page

/ 'ft--

CHAPTER 3.5 INSERT D TO SR 3.5.5.1 BASES A differential pressure that is above the reference minimum value is established between the l charging header (PT-121, charging header pressure) and the pressurizer, and the total seal l

injection flow is verified to be within the limits determined in accordance with the ECCS safety analysis. )

Edb;$l 4

- l Chapter 3.5 Insert Page

f;

(

l 1

1 1

i f

Chapter 3.7 i

! ~

r I

FNP TS Conversion Enclosure 5 - JD from STS Chapter 3.7 - Plant Systems l STS 3.7.2 MAIN STEAM ISOLATION VALVES 1

l FNP ITS 3.7.2 MAIN STEAM ISOLATION VALVES (MSIVs)

JD NUMBER JUSTIFICATION Therefore, plants designed with a single MSIV per steam line only have a " separate Condition entry" allowance for STS 3.7.2 Condition C (Modes 2 and 3). The FNP dual MSIV design allows the availability of the MSIV isolation function to be teured for each steam line in Mode 1 with a single inoperable MSIV as well as Modes 2 and 3. The application of the " Separate Condition entry" note is further justified for FNP due to the short Completion Time (4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ) provided for the ,

proposed FNP ITS Conditions B and E which address two inoperable MSIVs in a single steam line. FNP ITS Conditions B and E address a loss ofisolation function in a steam line and effectively correspond to STS 3.7.2 Conditions A and C for a plant with a single isolation valve. However, the proposed Completion Time of FNP ITS Conditions B and E is more restrictive than the STS Condition A and C Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . 'Ibe FNP ITS 4-hour Completion Time provides an adequate limitation on plant operation to assure appropriate action is taken to either restore the MSIV isolation function in the affected steam line, reduce power, or isolate the affected steam line.

4 The STS surveillance SR 3.7.2.1 is revised consistent with the corresponding CTS surveillance. CTS surveillance 4.7.1.5 is performed pursuant to Specification 4.0.5.

CTS Specification 4.0.5 references the Inservice Testing and Inspection Programs.

The STS does not have a corresponding specification to the CTS 4.0.5. The STS surveillance SR 3.7.2.1 provides the option to reference the Inservice Testing Program directly in the frequency column. As the CTS referenced the Inservice Testing Program (via 4.0.5) the STS surveillance option referencing the same program is selected to be consistent with the CTS.

5 STS SR 3.7.2.1 is revised consistent with the corresponding CTS surveillance requirement 4.7.1.5. This surveillance verifies the closure time of the MSIVs. The CTS surveillance requirement is performed in accordance with the Inservice Testing s Program (Specification 4.0.5). The Inservice Testing Program does not require that valve closure time be measured using a simulated or actual (ESP) actuation signal.

M The Inservice Testing program allows the associated manual handswitch to be used when verifying valve closure time. The required closure time in this FNP (T surveillance (7 seconds) is not the ESF Response Time (9 seconds) which is specified in FSAR Table 7.3-16 and required to be tested by ESFAS LCO 3.3.2, ESF Response Time surveillance. The ESF Response Time testing requires the channel response from the sensor to and including the actuated component be measured. The CTS MSIV surveillance 4.7.1.5 is intended to meet the requirements of the Inservice Testing Program to detect valve degradation over time by verifying valve closure within 7 seconds. The time required to be met by the

[

Chapter 3.7 ES-4-B April,1999

l l

I FNP TS Conversion I l

Enclosure 5 - JD from STS Chapter 3.7 - Plant Systems STS 3.7.2 MAIN STEAM ISOLATION VALVES FNP ITS 3.7.2 MAIN STEAM ISOLATION VALVES (MSIVs)

.A JD NUMBER JUSTIFICATION CTS surveillance does not include the two additional seconds of the ESFAS 4pl Response Time requirement (9 seconds specified in FSAR Table 7.3-16) for the main steam isolation function which accounts for the associated ESF actuation electronics and sensor. Therefore, deletion of the STS requirement to use a simulated or ac.ca! 'ESF) signal in SR 3.7.2.1 to measure the MSIV stroke time is acceptable and . 4 stent with the intent of the corresponding CTS surveillance and ;

the Inservice Tuting Program which require the valve closure time be measured but do not specify an actuation signal be present or simulated in order to accomplish the j measurement. i l

1 l

i i

Chapter 3.7 ES-5-B April,1999

FNP TS Conversion Enclosure 2 - Discussion of Changes to CTS Chapter 3.7 - Plant Systems g CTS 3/4.7.1.4 ACTIVITY l FNP ITS 3.7.16 SECONDARY SPECIFIC ACTIVITY DOC EQ SHE- DISCUSSION 1 A The CTS 3/4.7.1.4 Action statement is revised consistent with the STS.

The specific activity limit contained in the CTS Action statement repeats 1 the LCO requirement and is deleted. Since this limit is specified in the LCO, the Action is revised to simply reference the " limit". This revision

does not introduce a technical change and is made to conform with the format and presentation of the STS. Therefore, this change is considered l administrative. -

l 2 M The CTS surveillance 4.7.1.4 is revised consistent with the STS. The CTS surveillance is simplified to verify the LCO limit (0.10 ci/gm Dose Equivalent I-131) is met every 31 days. In addition, the CTS Table 4.7-2 and references to it are deleted from the CTS. The CTS Table 4.7-2 contained conditional frequencies for the performance of surveillance

' 4.7.1.4 based on the results of a more frequent gross activity determination.

These CTS conditional frequencies are deleted consistent with the STS.

The frequency of surveillance 4.7.1.4 is revised to a straight 31 days without conditions based on the gross activity level. The requirement to perform this surveillance at the more restrictive frequency of every 31 days provides additional assumnce the LCO limit is met. However, this change results in a requirement to always perform this surveillance at 31 day intervals instead of the CTS allowance for 6 month intervals (depending on the results of the gross reactivity determination). Therefore, this change is considered more restrictive. ,

3 L He CTS surveillance 4.7.1.4 is revised consistent with the STS. The CTS surveillance requirement on Table 4.7-2 to perfonn a gross activity determination every 72 days is deleted. The results of the gross activity surveillances are used in the CTS to detennine the frequency at which the Dose Equivalent I-131 surveillance must be performed (31 days or 6 months). In the applicable dose analyses, the radiciodines and the resulting thyroid dose are limiting not the noble gases and the whole-body dose. As such, the Specific Activity LCO requirement is based on Dose Equivalent I-131 not gross activity. Therefore, the deletion of the CTS gross activity requirement is acceptable based on the more restrictive requirement imposed on the Dose Equivalent I-131 surveillance which must always be performed every 31 days regardless of the gross activity. The Dose Equivalent I-131 surveillance performed every 31 days provides adequate Chapter 3.7 ' E2-1-D April,1999

FNP TS Conversion Enclosure 2 - Discussion of Changes to CTS Chapter 3.7 - Plant Systems N

./

-I y CTS 3/4.7.1.4 ACTIVITY 4l FNP ITS 3.7.16 SECONDARY SPECIFIC ACTIVITY DOC N_Q Sjig DISCUSSION ,

assurance that the LCO limit is being met and that the assumptions of the applicable dose analyses regarding radioiodines and thyroid dose remain valid.

l l

l 1

j i

i i

I i

Chapter 3.7 E2-2-D April,1999 I

i l

1 1

l Associated Package Changes for Chapter 3.7 l

l 1

i l

1 1

1 Associated Package Changes for RAI- 3.7.1.1 I l

I

' I

FNP TS Conversion Enclosure 2 - Discussion of Changes to CTS Chapter 3.7 - Plant Systems )

CTS 3/4.7.1.1 SAFETY VALVES N- FNP ITS 3.7.1 MAIN STEAM SAFETY VALVES (MSSVs)

Y N DOC T EQ SHE DISCUSSION 4 L CTS 3/4.7.1.1 Action a is revised consistent with the STS Actions format.

The CTS action requirement is divided into three separate Action statements. The STS intmduces Action A.1 to reduce THERMAL POWER to s 87% when one or more Steam Generators (SGs) have one MSSV inoperable and the Moderator Temperature Coefficient (MTC) is zero or negative. This corresponds to the actions taken in CTS action a, with the MTC negative or zero based on the

footnote on CTS Table 3.7-1. -

However, for this case the reduction of the Power Range Neutron Flux-High trip setpoint is not required in the STS. This proposed change is !

acceptable based on the fact that a reactor power reduction alone is I sufficient to limit primary side heat generation such that overpressurization of the secondary side is precluded for any RCS heatup event. Fmthennore, for this case there is sufficient total steam flow capacity provided by the turbine and the remaining OPERABLE MSSVs to preclude overpressurization in the event of an increased reactor power due to reactivity insertion, such as in the event of an uncontrolled RCCA bank withdrawal at power. The STS introduces a separate Action statement (B.1) to reduce power to below the value specified in Table 3.7.1-1 within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months when one or more SGs have one MSSV inoperable and the MTC is positive or when one or more SGs have two or more MSSVs inoperable.

The existing CTS Action to reduce the Power Range Neutron Flux-High trip setpoint becomes Action B.2. In addition, the 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months allowed to reduce the Power Range Neutron Flux-High trip setpoint is revised to 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months in the proposed Action B.2. The proposed increase in the time allowed to adjust the power range instrumentation trip setpoint is based on allowing the technicians a more reasonable time in which to perform this l action. A total of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months from Condition entry or 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months following the required power reduction is a more reasonable time for the technicians to prepare for and perform the required setpoint adjustments. The proposed time would allow the adjustments to be made in a more orderly manner and reduce the potential for an unplanned plant transient due to a reactor trip resulting from a personnel error in adjusting the trip setpoints. The proposed increase in the time to adjust the setpoints is acceptable based on the fact that the TS Actions still require power be reduced according to Table 3.7.1-1 within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the likelihood of an event occurring within the l 36 hour4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months period that would require the trip setpoint reduction is small, and the potential for an inadvertent reactor trip resulting from an error in the Chapter 3.7 E2-2-A April,1999 i

FNP TS Conversion Enclosure 2 - Discussion of Changes to CTS Chapter 3.7 - Plant Systems CTS 3/4.7.1.1 SAFETY VALVES FNP ITS 3.7.1 MAIN STEAM SAFETY VALVES (MSSVr)

DOC NLQ SHE DISCUSSION setpoint adjustment is reduced.

. ~4a L The CTS 3/4.7.1 Action Statement "a" requirement to reduce the power range neutron flux-high trip setpoint per Table 3.7-1 is revised by the addition of a Note which specifies that this Action is only required in Mode

1. CTS 3/4.7.1 is applicable in Mode 1-3. The proposed Note will provide an exception to the requirement to perform this Action in Modes 2 and 3.

This change is consistent with the FNP safety analyses which rely on the MSSVs for accident mitigation (confinned by Westinghouse). During operation in Modes 2 or 3, the power range neutron flux-lpw trip and the source range neutron flux-high trip functions (required operable in the Reactor Trip System LCO,3.3.1) provide adequate protection consistent with the assumptions of the applicable safety analyses. From Modes 2 or 3, a rod withdrawal event is the only transient described in the FSAR which would result in a significant power rise and potentially challenge the relief capacity of 2 operable MSSVs (minimum allowed on a single SO by Table 3.7.1-1). In Mode 1, Table 3.7.1-1 would require that the power range neutron flux-high trip setpoint be reduced to 24% RTP with only 2 operable MSSVs on any SG. The 24% trip setpoint limit is based on a near steady-state calculation which demonstrates that 2 MSSVs per loop are capable of maintaining the main steam system pressure below the required value durmg an event initiated with measured reactor power s 24% RTP. For a transient initiated from less than 5% RTP (Mode 2), the reactor would experience a significant power rise prior to approaching protection system actuation by the power range neutron flux-low trip setpoint (25% RTP) or challenging the available MSSV capacity. For a rod withdrawal event initiated at less than 5% power there is sufficient lag between the core power generation (driven by rod withdrawal) and the corresponding steam j demand such that the capacity of 2 MSSVs is sufficient to maintain the !

main steam system pressure below the safety analysis limit. Therefore, although the power range neutron flux-low trip setpoint (25% RTP) is slightly higher than the required Mode 1 power range high flux setpoint reduction (24% RTP), the power range neutron flux-low setpoint will sufficiently limit the peak core power level and, thereby, the main steam system pressure. A rod withdrawal transient initiated from a suberitical condition (Mode 3) would be terminated by the source range neutron flux-5 high trip function (at 10 cps) before any significant power level is attained and, therefore, no challenge to the MSSV relief capacity or main steam Chapter 3.7 E2 3-A March,1998

FNP TS Conversion Enclosure 2 - Discussion of Changes to CTS Chapter 3.7 - Plant Systems CTS 3/4.7.1.1 SAFETY VALVES FNP ITS 3.7.1 MAIN STEAM SAFETY VALVES (MSSVs)

DOC MQ SHE DISCUSSION system pressure would occur. Given the protection provided by the power range neutron flux-low reactor trip (in Mode 2) and the soun:e range neutron flux-high reactor trip (in Mode 3), the reduction of the power range neutron flux-high setpoint is overly conservative and not required in these Modes to ensure the acceptance criteria of the safety analyses are met. l Therefore, this change is acceptable for FNP. l l

M 5 CTS 3/4.7.1.1 Action e is revised consistent with the STS. CTS Action c specifies that the provisions of Specification 3.0.4 are not applicable. This CTS Action allows the Mode of Applicability (1-3) to be entered in order to '

perform the required surveillance testing on the MSSVs at normal temperatures and pressures. In the STS, this allowance is accomplished by the use a note that modifies the applicable surveillance. The standard STS i note used is explained in Section 1.4," Frequency" of the STS. In this case I the note specifies that the surveillance is only required to be performed in Modes 1 and 2. Therefore, entry into Mode 3 is permitted in order to perform testing of the MSSVs. The note, as used in the STS and explained ;

in Section 1.4, requires testing be performed prior to entering Mode 2 or 1. I The allowance to test the valves in Mode 3 is sufficient to assure the MSSVs are tested at normal operating temperatures and pressures. The STS note provides a similar but more restricted allowance for testing than N

y Action c of the CTS. Action c of the CTS provided a blanket exception to the requirement to meet the MSSV LCO in any Mode. Therefore, this

^/ change provides an adequate allowance to ensure the operability of the 9 MSSVs but is considered a more restrictive change.

Y 6 L The CTS 3/4.7.1.1 Actions are revised consistent with the STS. A new g Actions Condition (C) is added. Condition C contains the shutdown Actions of CTS action statement a for the condition of failing to reduce power or setpoints within the required time (Required Action and associated Completion Time not met) as well as the new STS condition for less than the minimum (2) MSSVs operable. The new STS Condition addresses a condition (less than the minimum MSSVs operable) that was not previously addressed by the CTS However, the Actions and Completion Times associated with this new Condition are effectively the same as the actions and times associated with LCO 3.0.3 which would otherwise be applicable without this new Condition. The Completion Time l associated with the new Condition C requires that the unit be placed in Chapter 3.7 E2-4-A April,1999

FNP TS Conversion Enclosure 2 - Discussion of Changes to CTS Chapter 3.7 - Plant Systems N

\) CTS 3/4.7.1.1 SAFETY VALVES T FNP ITS 3.7.1 MAIN STEAM SAFETY VALVES (MSSVs)

DOC N_Q SBE DISCUSSION Mode 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in Mode 4 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ).

l Adopting the STS Completion Time for Condition C introduces a technical change to the CTS. The corresponding CTS Completion Time requires that the unit be placed in Mode 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and Mode 5 in the following 30

hours. Consistent with the normal application of Actions and Completion l Times throughout the TS, the STS Action and Completion Time address l removing the unit from the Mode of applicability (1-3) by requiring that the i unit be placed in Mode 4. In Mode 4, the MSSV LCO is no longer l applicable and the MSSVs are no longer required operable to mitigate transient events. The CTS Action to place the unit in Mode 5 is overly conservative and is technically no longer required once the unit reaches

' Mode 4 where the LCO is no longer applicable. Therefore, the CTS Actions are revised consistent with the normal application of Actions and ;

l Completion Times in the TS and with the STS.

7 A 'Ibe CTS surveillance 4.7.1.1 is revised consistent with the STS. The CTS surveillance references Specification 4.0.5 which does not exist in the STS.

CTS Specification 4.0.5 contains requirements for the Inservice Inspection and Test Programs. The Inservice Testing Program which is applicable to this surveillance is specified in the Program section of the STS Administrative Controls Section. Therefore, the STS MSSV surveillance specifies that the lift settings of each MSSV must be verified and directly references the Inservice Testing Program. In addition, the STS surveillance contains the il% as left tolerance previously contained in the " footnote to MSSV CTS Table 3.7-3. This change to the CTS MSSV surveillance requirement is made to conform with the presentation and format of the information in the STS. No technical change is intended. Therefore, this change is considered administrative.

8 A CTS Table 3.71 is revised consistent with the STS. This CTS Table contains the limiting power when MSSVs are inoperable. The unit must operate below and adjust power range high flux trip setpoints to the power level specified in this Table when MSSVs are inoperable. The CTS Table is revised to state the number of MSSVs in terms of those operable instead of maximum inoperable. The actual power level requirements are not changed. Therefore, no technical change is introduced by the editig and reformatting of this Table. The changes made to the CTS Table are editorial und are required to conform with the format and presentation of Chapter 3.7 E2-5-A April,1999 )

i

l l'

)

FNP TS Conversion Enclosure 2 - Discussion of Changes to CTS Chapter 3.7 - Plant Systems CTS 3/4.7.1.1 SAFETY VALVES FNP ITS 3.7.1 MAIN STEAM SAFETY VALVES (MSSVs) l DOC N_Q SHli DISCUSSION this information in the STS. As such all changes to this CTS Table are considered administrative. ,

l 9 LA The CTS Table 3.7-3 is revised consistent with the STS. The MSSV orifice size listed in this CTS Table is moved to the bases of the corresponding ,

STS MSSV LCO 3.7.1. This level of detail is typically not included in the STS. Additionally, the MSSV orifices are design features of the system,and i any changes to the design of safety-related systems would be adequately controlled by the provisions of existing FNP QA/QC programs. The l placement of this information in the associated TS bases is consistent with 1

. the location of similar system description details and operability requirements in the STS bases. Reliance on the information contained in the STS bases for system descriptions and design information is acceptable l since changes to the information in the bases is controlled by the Bases l

Control Program specified in the administrative controls section of the TS.

10 LA The

footnote to CTS Table 3.7-3 is revised consistent with the STS. This footnote provides guidance in performing the required surveillance testing of the MSSVs. The CTS note requires that lift setting pressure correspond l to nominal operating temperature and pressure. The STS bases for the ;

I required MSSV surveillance test contains similar guidance regarding the lift setting pressure being corrected to operating temperature and pressure.

Therefore, the existing CTS

footnote is considered to be effectively incorporated into the STS bases. Reliance on the information contained in

! the STS bases for guidance in performing surveillance testing is acceptable l since changes to the information in the bases is controlled by the Bases Control Program specified in the administrative controls section of the TS.

11 A- The " footnote to CTS Table 3.7-3 is revised consistent with the STS.

l ,

This CTS footnote provides guidance in performing the required l surveillance testing of the MSSVs. The footnote specifies that after testing l the setpoint must be within i 1% of the required lift setting. The l cwsponding STS surveillance SR 3.7.1.1 contains the same guidance within the surveillance requirement stated in the TS. The information contained within this footnote is moved from the CTS Table to within the applicable STS surveillance requirement consistent with the presentation of this information in the STS. This change is made to conform with the presentation and format of this information in the STS and does not t h ee 3.7 E2-6-A March,1998

Q l'

FNP TS Conversion Enclosure 2 - Discussion of Changes to CTS l

Chapter 3.7 - Plant Systems l

CTS 3/4.7.1.1 SAFETY VALVES l l l l

FNP ITS 3.7.1 MAIN STEAM SAFETY VALVES (MSSVs) l Doc HQ SE DISCUSSION l l introduce any technical change. Therefore, it is considered administrative. l 12 LA The CTS Table 3.7-1 *" footnote is revised consistent with the format of the corresponding STS 3.7.1. In the STS, a new Condition (A) is added to address the case where,one MSSV is inoperable on one or more SGs and the MTC is zero or negative. This footnote provides guidance on the maximum power level which may be maintained under these conditions.

The maximum power level is incorporated into Action A.1 and the cont'ents of the note are moved to the Bases for Action A.I. Reliance on the i information contained in the STS bases for details related to the associated

/.ctions is acceptable since changes to the information in the bases is controlled by the Bases Control Program specified in the administrative controls section of the TS.

D' i

i l

l Chapter 3.7 E2-7-A April,1999

I FNP TS Conversion Enclosure 3 - Significant Hazards Evaluations Chapter 3.7 - Plant Systems III. SPECIFIC SIGNIFICANT HAZARDS EVALUATIONS CTS 3/4.7.1.1 SAFETY VALVES y 4l- FNP ITS 3.7.1 MAIN STEAM SAFETY VALVES (MSSVs)

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

~

The proposed changes extend the time allowed to adjust the Power Range Neutron Flux-High trip setpoints for the case of two or more inoperable MSSVs per SG and/or positive Moderator Temperature Coefficient (MTC) and removes the requirement to adjust the Power Range Neutron Flux-High trip setpoints only one MSSV is inoperable and the MTC is zero or negative and do not result in any hardware or operating procedure changes. The affected trip setpoints, the requirement to reduce them or the time allowed to adjust them are not assumed to be an initiator of any analyzed event. In addition, the affected trip setpoints, the requirement to reduce them and the time allowed to adjust them are not a precursor to any accident analyses. Therefore, the proposed changes do not increase the probability of an accident previously evaluated. The Power Range Neutron Flux High trip functions to mitigate the consequences of an analyzed event by shutting down the reactor. The proposed l changes continue to provide assurance that the setpoints will be properly adjusted to ensure the system functions as assumed in the applicable safety analyses. Therefore, the consequences of an accident are not significantly increased.

2. Does the change create the possibility of a new or different kind of accident from any accident previouslyevaluated?

l The proposed changes do not necessitate a physical aheration of the plant (no new or different type of equipment will be installed) or changes in parameters goveming normal 1 plant operation. The proposed changes still ensure the operability of the trip function at the correct setpoint and will facilitate the adjustment of the setpoints such that the probability of l error is minimized. 'Ihus, these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

The time allowed to adjust the setpoints of the affected instrumentation is not a specific assumption of any safety analysis. For the case of a single inoperable MSSV with a zero or negative MTC, a reactor power reduction alone is sufficient to limit primary side heat generation such that overpressurization of the secondary side is precluded for any RCS heatup event. Furthermore, for this case there is sufficient total steam flow capacity provided by the turbine and the remaining OPERABLE MSSVs to preclude overpressurization in the event of an increased reactor power due to reactivity insertion, such as in the event of an Chapter 3.7 E3-1-A April,1999

FNP TS Conversion Enclosure 3 - Significant Hazxds Evaluations Chapter 3.7 - Plant Systems g,\ III. SPECIFIC SIGNIFICANT HAZARDS EVALUATIONS

'3 ' CTS 3/4.7.1.1 SAFETY VALVES FNP ITS 3.7.1 MAIN STEAM SAFETY VALVES (MSSVs) uncontrolled RCCA bank withdrawal at power. The proposed changes still ensure the setpoints are reduced consistent with the assumptions of the safety analysis for the case of two or more inoperable MSSVs or a positive MTC. The proposed changes also reduce the potential for an inadvertent reactor trip that could result from adjusting the trip setpoints too quickly. As such, any reduction in a margin of safety will be insignificant and will likely be offset by the benefit gained from the reduced potential for an inadvertent plant trip. :

Chapter 3.7 E3-2-A April,1999 l

1%

MSSVs 3.7.1 3.7 PLANT SYSTEMS 3.7.1 Main Steam Safety Valves _(MSSVs)_ T6TF 2 3 % Rev. ) l LCO 3.7.1 h

[sh had MSSVstshail be OPERA 8LE ss specified in Table 3.7.1-1[

s Qno laDie J./.1- [

~

APPLICA8ILITY: MODES 1, 2, and 3.

Y .

n. A.CTIONS .

4 ---------------------------------NOTE-----------------------------

Separate Condition entry is allowed for each MSSV. ~

CONDITION .

REQUIRED ACTION COMPLETION.T A. One or required A.1 Reduce power to less 4 rs MSSVs inope 1e. than or equal to the

( applicable % RTP listed in Table 3.7.1 .

B. Required Action and 8.1 MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

associated Completion .

Time not met. I QB 8.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months One or so saa generat with less that wo] MSSVs LE.

e x IeJ569-T LL

's,,

WOG STS 3.7-1 Rev 1, 04/07/95

1 L

CHAPTER 3.7 INSERT LL ~ 1 3 5, 8cv I TO STS LCO 3.7.1 PAGE 3.7-1 CONDITION REQUIRED ACTION COMPLETION TIME A. One or more steam generators A.1 Reduce THERMAL 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months with one MSSV inoperable POWER to s 87 % RTP.

and the Moderator Temperature Coefficient (MTC) zero or negative at all power levels.

B. One or more steam generators B.1 Reduce THEPJMAL 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months with two or more MSSVs POWER to less than or inoperable. equal to the Maximum Allowable % RTP specified -

QE in Table 3.7.1-1 for the -

number ofOPERABLE One or more steam generators MSSVs.

with one MSSV inoperable !

and the MTC positive at any 3 l power level.

--.--NOTE----

Only required in MODE 1 l 9

Q .

B.2 Reduce the Power Range 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Neutron Flux-High reactor trip setpoint to less than or equal to the Maximum Allowable % RTP specified in Table 3.7.1-1 for the number of OPERABLE MSSVs.

C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AN_D gg C.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months One or more steam generators with 2 4 MSSVs OPERABLE Chapter 3.7 Insert Page

IST l

T3TF-2.3% Ru, ( j Table 3.7.1-1 (page 1 of 1) $$(( f t

OPERA 8LE Main Steam Safety Valves versus Applicable Power in Forcent of RATED THERMAL POWER MINIMUM NUMBER OF MSSVs PER STEAM GENERATOR APPLICABLE POWER REQUIRED OPERA 8LE (% RTP) 5 s 100 4 s 80 3 :

s 60 -

2 s 40 p ,\

g i

i l .

i s.

W0G STS .

3.7-3 Rev 1, 04/07/95 4

V; \

178 %

~

T '235 Ov I CHAPTER 3.7 INSERT MM TO STS PAGE 3.7-3 3'

q.\

Table 3.7.1-1 (page 1 of 1)

OPERABLE Main Steam Safety Valves versus Maximum Allowable Power NUMBER OF OPERABLE MAXIMUM ALLOWABLE j MSSVs PER STEAM GENERATOR POWER (% RTP) l l .

4 60 3 43 2 24 l

l i

l Chapter 3.7 Insert Page

FNP TS Conversion Enclosure 5 - JD from STS Chapter 3.7 - Plant Systems STS 3.7.1 MAIN STEAM SAFETY VALVES (MSSVs)

FNP ITS 3.7.1 MAIN STEAM SAFETY VALVES (MSSVs)

JD NUMBER JUSTIFICATION 1 Not used.

2 Not used.

3 Not used.

4 The FNP specific Unit I and 2 information has been added to STS Table 3.7.1.-2. !

s The lift setpoints and 3 steam generator MSSV identification numbers are provided.

y The addition of this information is consistent with the CTS.

T 5 Not used.

6 Not used.

Chapter 3.7 E5-1-A April,1999

963 -

MSSVs 8 3.7.1 83.7 PLANT SYSTEMS B 3.7.1 Main Steam Safety Valves (MSSVs) i BASES TM-UT b I

! BACKGROUND The primary purpose of the MSSVs is to provide overpressure rotection for the secondary system. The MSSVs also provide '

fmust h M,fave sum. .eD N

rotection against overpressurizing the i YT M_- [_pressure boundary (RCPB) by providing a heat sink for theremova l

the preferred heat sink, provided by the fondenser and 16! Circulating Water System, is not avai ~.

e i 1 Five MSSVs are located on each main header, outside ~

g

, containment, upstream of the main s l b described in the FSAR, Sectioni )l0.3. Molatihvalves, (Ref. I as P 9'

IEN~I Lcapacity criteria is lluz or ra The MSSV

ea sneaa r ow/au llot " "

1 (I* g4 >

steam generator design pressu e requirements of the ASME Code g1pthe on ad go.r.2g).' ThencmnerQ@

%ch M86V , MSSV design includes staggered setpoints, according to ho4 a.,u,M.m. '

Table 3.7.1-2 in the accompanying LC0, so that only the orific.e., to CimW needed valves will actuate. Staggered setpoints reduce the I g%pw, potential for valve chattering that is due to steam pressure l insufficient to fully open all valves following a turbine l reactor trip.

APPLICABLE The design basis for the MSSVs comes from Reference 2 and SAFETY ANALYSES its purpose is to limit the_ secondary system pressure to A is s 110% of design pressiirer.n o p. .ing ivv. or assign steam

~~

h r (a~

This desian hat < t it sufficient in cana inith Tow. operational occurrence (A00) or accident nticipated y

considered in the Design Basis Accident (DBA) and transient g,\.\ analysis.

g The events that challTWfe the relievin apa yo the 9,1 MSSVs, and thus RCS essure, are the characterized as decreasedheatjeno events, whic are presented in the

,,g FSAR,Section$15. (Ref. 3). Of hese, the full power cg TA turbine trip without steam dump is he limiting A00. This PM 0' event also terminates normal feedwater flow to the_ steam eg.y . o35 generators-QgarnIpsdawanetahsM%e3 W.

The nsient response for turbine tria.without a direct reactor trip presents no hazard to the[ Integrity of the RCS (continued)

WOG STS B 3.7-1 Rev 1, 04/07/95

86V ,

MSSVs

'GTF 235 Av. ) B 3.7.1 l

BASES .TNSEA.T F y

APPLICABLE or the Main Steam systemMJIf a minimum reactivity feedback

+

SAFETY ANALYSES ris assumed, the reactor is tripped on high pressurizer (continued) pressure. In this case, the pressurizer safety valves open, and RCS pressure remains below 110% of the desi 6'2 '7 (The MSSVs also open to limit the saraadary =+===gn value.oressure.)

! 0 1.7.13 ]f maximum reactivity feedback is assumed, the reactor is CN,p cg-oG g# tripped on overtemperature AT. The departure from nucleate by , e i-boilingrattoincreasesthroughoutthetransient,andnever) drops below its initial value. Pressurizer relief valves CH-%'.96.e3'I A o -

.and MSSVs are activated and prevent nyer>ressurization in/

Lthe pri_ mary and secondary systems.J ne Insus are assumed to nave two active and one passive failure modes. The active failure modes are spurious opening, and failure to reclose once opened. The passive failure mode is failure to open

, p9 q,\.g upon demand.

. g. .

e MSSVs s sfy Cri+=* ion 3 of_1he NRC Pol Statement. l 3

/M NE5&3LE. k 'd e' .T_

N I )** b I iv D S_

LC0 *p The accideit analysis requ res our MSSVs per steam generato o provide overp ssu rotection for design 1 V

CN.yA'g*l 4,eg- basis transients occurr nel t 102% RTP. fan MSSV will '

p (consiaeres Tno mracle if < t vaiis i.e o wn on dennnd a fk i i.w reautrae t iat nye resus*De ort.nas.t in comp' f ance with 68 maference2)pensnougnsnisisnotarequiremento th i D8A analysis J This is because operation with less than e full number of MSSVs requires limitations on allowable g THERMAL POWER (to meet ASME Code requirements). These limitations are according to Table 3.7.1-1 in the accompanying LCO, and Reauired Action A.2. r No -

The OPERABILITY of the MSSVs is defin as the ability to E, n ithin the setpoint tolerances, elleve steam generator overpressure, and ressat when pressure has been reduced.

The OPERABILITY of the MSSVs is detemined by periodic 00 surveillance testing in accordance with the Inservice 09 Testing Program.

The lift settings, according to Table 3.7.1-2 in the accompanying LCO, correspond to ambient conditions of the valve at nominal operating temperature and pressure.

4 @asm 4o a4 /

M M w SR 3 7.h ,p. 8.fa -5 (continued)

W0G STS B 3.7-2 Rev 1, 04/07/95

c 45 CHAPTER 3.7

~T?tF-1-K 4. I INSERT F )

6 Q TO STS BASES PAGE B 3.7-2 FNP SPECIFIC SAFETY ANALYSIS INFORMATION Y (From Westinghouse) b f One turbine trip analysis is performed assuming primary system pressure control via operation of the pressurizer relief valves and spray. This analysis demonstrates that the DNB design basis is met.

l l Another analysis is performed assuming no primary system pressure control, but crediting reactor trip on I

high pressurizer pressure and operation of the pressurizer safety valves. This analysis demonstrates that l RCS integrity is maintained by showing that the maximum RCS pressure does not exceed i 10% of the design pressure. All cases analyzed demonstrate that the MSSVs maintain Main Steam System integrity l by limiting the maximum steam pressure to less than I 10% of the steam generator design pressure.

In addition to the decreased beat removal events, reactivity insertion events may also challenge the relieving capacity of the MSSVs. The uncontrolled rod cluster control assembly (RCCA) bank withdrawal at power event is characterized by an increase in core power and steam generation rate until l reactor trip occurs when either the Overtemperature AT or Power Range Neutron Flux-High setpoint is reached. Steam flow to the turbine will not increase fkom its initial value for this event. The increased heat transfer to the secondary side causes an increase in steam pressure and may result in opening of the MSSVs prior to reactor trip, assummg no credit for operation of the atmospheric or condenser steam dump valves. The FSAR Section 15.2.2 safety analysis of the RCCA bank withdrawal at power event for a range ofinitial core power levels demonstrates that the MSSVs are capable of preventing secondary l side overpressurization for this AOO.

l The FSAR safety analyses discussed above assume that all of the MSSVs for each steam generator are OPERABLE. If there are inoperable MSSV(s), it is necessary to limit the primary system power during steady state operation and AOOs to a value that does not result in exceeding the combined steam flow capacity of the turbine (if available) and the remaining OPERABLE MSSVs. The required limitation on primary system power necessary to prevent secondary system overpressurization may be determined by system transient analyses or conservatively arrived at by simple heat balance calculation. In some circumstances it is necessary to limit the primary side heat generation that can be achieved during an l AOO by reducing the setpoint of the Power Range Neutron Flux-High reactor trip function. For example, if more than one MSSV on a single steam generator is inoperable, an uncontrolled RCCA bank withdrawal at power event occurnng from a partial power level may result in an increase in reactor power that exceeds the combined steam flow capacity of the turbine and the remaining OPERABLE MSSVs. Thus, for multiple inoperable MSSVs on the same steam generator it is necessary to prevent this power increase by lowenng the Power Range Neutron Flu-High setpoint to an appropriate value.

When the Moderator Temperature Coefficient (MTC).is positive, the reactor power may increase above the initial value during an RCS heatup event (e.g., turbine trip). Thus, for any number ofinoperable MSSVs it is necessary to reduce the trip setpoint if a positive MTC may exist at partial power conditions, unless it is demonstrated by analysis that a specified reactor power reduction alone is sufficient to prevent overpressurization of the steam system.

The maximum allewable power levels specified in Table 3.7.1-1 are overly conservative at middle and 1-end-of-life conditions, when the MTC is not positive. Therefore, a rpecific analysis which credits a middle-of-life MTC was psfwsd to relax the power reduction associated with one inoperable MSSV h

l per steam generator. In addition, for the above case, no reduction in the Power Range Neutron Flux-High l trip setpoint is required. The middle-of-life analysis assumes a -10 pcm/ degree F MTC and demonstrates that the maximum allowable power level associated with one inoperable MSSV per steam generator can be relaxed to 87% RTP when core average bumup is 214,000 MWD /MTU. The MTC value at 14,000 l l MWD /MTU is verified to be more negative than -10 pcm/ degree F for each reload cycle.

Chapter 3.7 Insert Page

hhh -

MSSVs

, 8 3.7.1

_w BASES e

5b Ye N'h LCO This LCO provides assurance that the MSSVs will perform I (continued) their designed safety functions to miti ate the consequencesj of accidents that could result in a cha lenge to the RCP APPLICABILITY I E I above 40% RTP, the number of MSSVs per steaa l enerator required to be OP LE must be according to abla 1.7 1.1 4a A **e ' vino LCO. Aslaw 40E RTP in; g penf AW MODES 1, 2, and 3, eniv tuspSSVs per steam generator are SW km requiredtobeOPERA8Ly ,

v f"*"' In MODES 4 and 5, there are no credible transients requiring ths MSSVs. The steam generators are not normally used for ~ .

heat removal in MODES 5 and 6, and thus cannot be l overpressurized; there is no requirement for the MSSVs to be OPERABLE in these MODES.

ACTIONS The ACTIONS table is modified by a Note indicating that separate Condition entry is allowed for each _MSSV.

fionNS bd f> 9 WithoneormoreMSSVsinoperable)<.educepowersothatthe available MSSV relieving capacity me . ce 2 requirements.for the applicanie IMtunAL POWER W Operation with less than all five MSSVs OPERA 8LE for each Tsteau qenerator is permissible, if THERMAL POWER is ca.TA. W o21 gropor :1onally) limited to the relief capacity of the -

88V * # remain'ng nuvs. This is accomplished by restricting CNorn Waf THERMAL POWER so that the energy transfer to the mort

"*** limiting steam generator is not greater than the available reliefcapacityinthatsteamgenerator.JForex le, it L

w 1Eie MSSV is 1 parable in a steam ge retor, th relief capacity of t at steam ge rator is r uced by a roximate 205. To off t this redu tion in rol f capaci , energy transfer to ; hat steam g erator must be simil ly reduc by at less 205. This 1 accomplish by redu ing THE L POWER by a least 205, ich conserv tively 1 its the energy tr sfer to all team genera rs to a roximat y 80 of total apacity, co 1 stent with he reli capaciyofj%

the nos limiting ste generator.c "

T95ERT&

g (continued)

WOG STS 8 3.7-3 Rev1,04/07/95

2li7 CHAPTER 3.7 q3 3 INSERT G T TO STS BASES PAGE B 3.7-3 REQUIRED ACTIONS A.1, B.1, and B.2 Ad, '

In the case of only a single inoperable MSSV on one or more steam generators, when the Moderator Temperature Coefficient is not positive, a reactor power reduction alone is sufficient to limit primary side host generation such that overpressurization of the secondary side is precluded for any RCS heatup event. Furthermore, for this case there is sufficient total steam flow capacity provided by the turbine and the remaining OPERABLE MSSVs to preclude overpressurization in the event of an increased reactor power due to reactivity insertion, such as in the event of an uncontrolled RCCA bank withdrawal at power. Therefore, Required Action A.} requires an eyyivyriate reduction in power within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

The maximum THERMAL POWER corresponding to the heat removal capacity of the remaining OPERABLE MSSVs is determined via a conservative heat balance calculation as described in the attachment to Reference 6, with an appropriate allowance for calorimetric power uncertainty.

B.1 and B.2 In the case of multiple inoperable MSSVs on one or more steam generators, with a reactor power reduction alone there may be insufficient total steam flow capacity provided by the turbine and the remaining OPERABLE MSSVs to preclude overpressurization in the event of an increased reactor power due to reactivity insertion, such as in the event of an uncontrolled RCCA bank withdrawal at power. Furthermore, for a single inoperable MSSV on one or more steam generators when the Moderator Temperature Coefficient is positive the reactor power may increase as a result of an RCS heatup event such that the flow capacity of the remaining OPERABLE MSSVs is insufficient. Therefore, in addition to reducing THERMAL POWER within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months as required by Required Action B.1, the Power Range Neutron Flux-High trip setpoint must be reduced to less than or equal to the applicable value correspondir- to the number of OPERABLE MSSVs specified in Table 3.7.1-1 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months as recjir d by Required Action B.2 (applicable in MODE 1 only). The safety analysis of the loss of -

[

load / turbine trip event analyzed from part power conditions to specifically support the requirements of Table 3.7.1-1, explicitly credits the Power Range Neutron Flux-High trip function to ensure that tre peak power does not exceed an acceptable level. With two or more MSSVs inoperable on one or more steam generstors, the reduced Power Range Neutron Flux-High trip setpoints will also limit the peak power to an acceptable level for an RCCA withdrawal at power transient occurring from similar conditions.

The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time for Required Action B.1 is consistent with A.I. An additional 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months is allowed in Required Action B.2 to reduce the setpoints. The Completion Time of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months is based on a reasonable time to correct the MSSV inoperability, the time required to perform the power reduction, operating experience in resetting all channels of a protective function, and on the low probability of the occurrence of a transient that could result in steam I generator overpressure during this period.

Chapter 3.7 Insert Page

c -

}

257. .

CHAPTER 3.7 l

TM-2.?fi", #cv I q,t \ ' INSERT G (con't) t TO STS BASES PAGE B 3.7-3 REQUIRED ACTIONS A.1, B.1, and B.2 The maximum THERMAL POWER corresponding to the heat removal capacity of the remaining -

OPERABLE MSSVs is determined via a conservative heat balance calculation as described in j the attachment to Reference 6, with an appropriate allowance for Nuclear Instrumentation System trip channel uncertainties.

Required Action B.2 is modified by a Note, indicating that the Power Range Neutron Flux-High reactor trip setpoint reduction is only required in MODE 1. In MODES 2 and 3, the reactor l protection system trips specified in LCO 3.3.1, " Reactor Trip System Instrumentation," provide sufficient protection.

The allowed Completion Times are reasonable based on operating experience to accomplish the Required Actions in an orderly manner without challenging unit systems. -

l 1

l l

l Chapter 3.7 Insert Page

861 MSSVs f 4 g.1'\ \ B 3.7.1 BASES r3g_z3g g,)

ACTIONS A d (continued)

. or each steam generator, at a specified pressure, th etional relief capacity (FRC) of each MSSV is det ined

. . as allows:

FRC =

8 where:

A = the lief capacity of the MSS , and -

B = the tot relief capacity o all the MSSVs of the steam gen ator.

The FRC is the relief pacity ne ssary to address operation with reduced RMAL ER.

The reduced THERMAL POWER 1 s in the LCO prevent operation at power levels gr ter than the relief capacity of the remaining MSSVs. T r uced THERMAL POWER is j determined as follows: .

RP = ,1 - (W, x FRC 3 + x FRC, + - + x FRC,), x 100%

where:

RP = Red THERMAL POWER for the most limiting steam ge rator expressed as a percent of TP; ,

N , N.,

i .. N represent the status of the MSSV 2' ,

..., 5, spectively, 0 if the MSSV is OPERABLE, I if the MSSV is inoperable; F 3, FRC , ... FRC, = the relief capacity of the MSSV 1,

.., 5, respectively, as defined above. l 1

l

'l (continued) l WOG STS g 3,7 4 Rev 1, 04/07/95 l I

I l

$09 S l 'I MSSVs Si T.6Tft *L3$ A6N./ B 3.7.1 Q ve d ht.:hevw m no b c. a p k b k BASES g

ACTIONS a (continued) _

A If the@SSVs cannot be restored to OPERABLE statuBwithin the associated Eompletion Tie f or if one orJmore steam

) 4 inop6M venerators have ue<< rama 'WSVs ggg8BJr, the unit must be placed in a MODE in which the LC0 does not apply. To achieve' this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full

~

power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.1.1 REQUIREMENTS This SR verifies the OPERASILITY of the MSSVs by the verification of each MSSV lift setpoint in accordance with f the Inservice Testina Program. (The coce, s n on ai IRef. 4), equires t (at safat and rol f valve ests be performed in accord nce with SI/ 0M-1-198 (Ref. 5).

Lecordi to Refer ce 5, t followi tests e requir :

a. V ual exami ation;

[ b. at tight ss date ination;

c. Setpoint ressure d terminat n (lift etting);

+*fi*p M S d. Complia e with er's sea tightnes criter ; and l A.Paika mah

e. Verifi tion of t a balanc ng device integri y on -)

balanc ,

valves. -

The ANSI /ASME Standard requires that all valves be tested !

Og'- every 5 years, and a minimum of 20% of the valves be tested j every 24 months._ T ~ pECodespecifiestheactivitiesand :

frequencies necessa .o satisfy the requirements. Table Eh pr # N t

3.7.1-2 allows a i setpoint tolerance for OPERASILITY; i

9wm Lco M 5 however, the valves are reset to i 1% during the t g 8 3.")-2. Surveillancetoallowfordrift.y This SR is modified by a Note that allows entry into and l operation in MODE 3 prior to performing the SR._ The MSSVs j may be either bench tested or tested in situ at hot ;

(continued) j l

WOG STS B 3.7-5 Rev 1, 04/07/95 ,

l l

e 9 71 MSSVs B 3.7.1 !

BASES SURVEILLANCE SR 3.7.1.1 (continued)

REQUIREMENTS conditions using an assist device to simulate lift pressure.

If the MSSVs are not tested at hot conditions, the lift setting pressure shall be corrected to ambient conditions of the valve at operating temperature and pressure.

REFERENCES 1. FSAR, Section 10.3

2. ASME, Boiler and Pressure Vessel Code,Section III, ,

Article NC-7000, Class 2 Components) 1971 ed'stsn .

I

3. FSAR, Section 5.2

4. ASME, Boiler and Pressure Vessel Code,Section XI 1%i MA y

5. GusI/ASME OM-1-1987 mg [d g(go Q',

~

le. @C. 2::4cccJim NoWa *N - l overpr*SS"iwW @ M Nw M(oo 5pkm, " Pa%%.I

" T A9sk. v2.j t%4. )

I T5tF- 2 56, Rev. } l 9# q.\ \

l i

l l

.. WOG STS 8 3.7-6 Rev 1, 04/07/95 i

l

I;

i l

l l

l Associated Package Changes for RAI-3.7.5-1 l

l i

l l

i l

l I

l l

1

FNP TS Conversion Enclosure 2 - Discussion of Chang'es to CTS l Chapter 3.7 - Plant Systems CTS 3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM l FNP ITS 3.7.5 AUXILIARY FEEDWATER (AFW) SYSTEM i Doc l NQ SHE DISCUSSION l STS Action in this condition (no AFW operable) is to prevent placing the ;

plant in a Mode where the AFW system would be required operable for heat !

removal. The addition of the STS note to this Action provides a clarification of the intent of this Action and assists in avoiding potential conflicts within the TS that may cause the plant to be placed in a Mode where the AFW was required. The addition of this STS note is consistent with the intent of the CTS Action and conforms with the presentation and format of this information in the STS. Therefore, this change is considered administrative.

6 M The CTS surveillance 4.7.1.2.1 is revised consistent with the STS. This CTS surveillance requires that each AFW pump be demonstrated operable pursuant to Specification 4.0.5. The CTS surveillance is revised to replace the general reference to Specification 4.0.5 with a specific STS test acceptance criterion; " verifying the developed head of each AFW pump at l the flow test point is 2 the required developed head". The CTS surveillance l

does not contain a direct reference to a specific acceptance criterion but relies on Specification 4.0.5 which in tum contains requirements for an Inservice Testing Program. The proposed STS acceptance criterion is consistent with standard industry pump testing methods and therefore is applicable to FNP. However, the addition of a specific test criterion in lieu of the existing reference to a testing program is considered a more restrictive change. ,

l 7 L CTS surveillance 4.7.1.2.1 is revised consistent with the format of the corresponding STS surveillance. This surveillance requires that each AFW l

! pump be demonstrated operable and contains an exception to Specification 4.0.4 for the turbine-driven pump to allow entry into Mode 3 for testing.

A

n.4 This exception is required to allow for adequate steam pressure to test the o' turbine-driven pump. The CTS exception to the provisions of 4.0.4 is i replaced with a more specific note in the ITS surveillance SR 3.7.5.2 consistent with the STS. Consistent with the STS, the proposed FNP ITS l

note does not restrict performance of the turbine-driven AFW pump to Mode 3 as does the CTS exception to Specification 4.0.4. As such, the adoption of the STS note format introduces a less restrictive change to the CTS requirements. However, the proposed FNP ITS note restricts the time allowed before the surveillance must be perfonned consistent with the STS.

l The time restriction provided by the note ensures the surveillance is Chapter 3.7 E2-3-B April,1999

FNP TS Conversion Enclosure 2 - Discussion of Changes to CTS Chapter 3.7 - Plant Systems CTS 3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM 1

FNP ITS 3.7.5 AUXILIARY FEEDWATER (AFW) SYSTEM l DOC l HQ SHE DISCUSSION l performed in a timely manner and provides the required allowance to reach the plant conditions necessary to perform the turbine-driven AFW pump

, testing. The steam generator pressure requirement of 21005 psig corresponds to the Reactor Coolant System no-load, hot standby, Mode 2

, y,6 A operating T.y, of 547 *F. The proposed change is made to conform with the 3- format of the STS surveillance notes and includes requirements that provide a sufficient limit on plant operation to ensure the required equipment is-maintained operable. Therefore, this change is acceptable.

8 A The CTS surveillance 4.7.1.2.1 is revised consistent with the STS by replacing the general reference to Specification 4.0.5. His CTS surveillance requires that each AFW pump be demonstrated operable pursuant to Specification 4.0.5. The CTS surveillance simply references Specification 4.0.5 for the acceptance criteria as well as the Frequency of the required testing. The STS provides a specific acceptance criterion (discussed previously) and a revised reference for the surveillance Frequency. As the STS no longer contains a Specification 4.0.5, the STS specifies a Frequency in accordance with the Inservice Testing Program.

The STS specifies a Inservice Testing Program directly in the Administrative Controls section of the TS and does not use a specification corresponding to CTS 4.0.5. Since CTS 4.0.5 contains requirements regarding the inservice Testing Program, which are applicable to the Frequency of the surveillance, the direct reference to the Inservice Testing Program in the Frequency of the STS surveillance is consistent witti the CTS surveillance 4.7.1.2.1 requirements. Therefore, this change is made to conform with the presentation and format of this information in the STS and is considered an administrative change.

9 LA- The CTS surveillances 4.7.1.2.2.a.1,2, and 3 are revised consistent with the STS. These 3 CTS surveillances verify valve positions every 31 days.

Each of the 3 CTS surveillances address different valves and contains specific requirements for those valves. The corresponding STS surveillance consists of a single surveillance that simply verifies all valves are in the correct position. Therefore, the CTS surveillances are revised to be a single surveillance which verifies all the AFW system valves are in the correct position. The specific valve position guidance contained in the CTS surveillance 4.7.1.2.2.a.3 is moved into the bases for the proposed new SR 3.7.5.1. The description of the correct position for the affected valves will

' Chapter 3.7 E2-4-B April,1999

y i

FNP TS Conversion Enclosure 2 - Discussion of Changes to CTS Chapter 3.7 - Plant Systems CTS 3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM FNP ITS 3.7.5 AUXILIARY FEEDWATER (AFW) SYSTEM DOC HQ SE DISCUSSION philosophy of the STS for this type of detail. Reliance on the information contained in the STS bases for guidance in perfonning the associated surveillance is acceptable since changes to the information in the bases is controlled by the Br.ses Control Program specified in the administrative j controls section of the TS. :

16 L The CTS surveillance 4.7.1.2.2.c.2 is revised consistent with the format of the corresponding STS surveillance. This CTS surveillance verifies the j automatic start of the turbine-driven AFW pump and is required to be performed every 18 months during shutdown. The CTS surveillance currently has no exception to Specifications 3.0.4 or 4.0.4 to allow entry in the applicable Modes for testing. The allowance to enter the applicable Modes to test the automatic actuation of the turbine driven AFW pump is provided by a note in the corresponding STS surveillance. The STS note allows the turbine-driven AFW pump automatic actuation testing to be performed at system pressures and temperatures close to normal operating p[ conditions which provides a better indication of system performance and operability. The steam generator pressure requirement of 21005 psig corresponds to the Reactor Coolant System no-load, hot standby, Mode 2 operating T.y of 547 F. The CTS surveillance is revised consistent with the allowance provided in the STS note. The incorporation of the STS note, which allows entry into the applicable Modes to test the automatic actuation of the turbine-driven AFW pump, provides the capability to perform this surveillance under conditions as close as practical to normal operating conditions and assures that an adequate assessment of the system performance and operability may be made in a timely manner.

17 LA The CTS surveillance 4.7.1.2.2.c.3 is revised consistent with the format of similar STS surveillances. This FNP specific surveillance verifies the turbine-driven AFW pump steam admission valves open when supplied by their associated air accumulators. The CTS surveillance contains the specific valve numbers of the steam admission valves being tested.

Consistent with the format and presentation of similar STS surveillance requirements, the specific valve numbers are removed from the CTS surveillance and placed in the proposed bases discussion for this FNP surveillance. The placement of descriptive information or guidance for performing a surveillance in the bases for the surveillance is consistent with the philosophy of the STS for this type of detail. Reliance on the Chapter 3.7 E2-8-B April,1999

FNP TS Conversion Enclosure 3 - Significant Hazards Evaluations Chapter 3.7 - Plant Systems j III. SPECIFIC SIGNIFICANT HAZARDS EVALUATIONS CTS 3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM FNP ITS 3.7.5 AUXILIARY FEEDWATER (AFW) SYSTEM 1:L

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

The proposed chinge eliminates the CTS requirement that the AFW turbine-driven AFW pump be tested in Mode 3 only. This change does not result in any hardware changes or change the manner in which the plant is normally operated. The turbine driven AFW pump is not assumed to be an initiator of any analyzed event. As a result, the revision of the testing requirements associated with the AFW turbine-driven pump does not affect the probability of an accident. The AFW pump is still required to be tested within a reasonable time and continued operation is not permitted without testing the pump. Therefore, the proposed change does not allow continuous operation such that a single failure could result in failure of the AFW System to perform its safety function. As such, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated.

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

The proposed change does not necessitate a physical alteration of the plant (no new or different type of equipment will be installed) or changes in parameters governing normal plant operation. The proposed change provides a reasonable time in which to test the turbine-driven AFW pump during a plant startup. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

,\

The proposed change involves the time allowed and plant conditions for testing a turbine-driven AFW pump during a plant startup. The applicable safety analyses do not contain specific assumptions regarding the time or plant conditions for performing surveillance testing of the turbine-driven AFW pumps. As such, the

[ proposed change does not introduce a new or less restrictive operating condition than the current technical specification requirements. Therefore, the proposed change does not involve a significant reduction in a margin of safety and is necessary to allow proper testing of the required systems.

Chapter 3.7 E3-1-B . April,1999

/ft C _ - - - - - - upe -. - - - - -

Not reqviced.to be perfocM6r ffe ARu f/aw ArW sy.t.m centrol des when 6 fo%RTP erwhen ne AFw 3.7.s G5s+em o$ no+

- - - - - m- autemsk. centte(, ._ _ - -- Q

.v-SURVEII.IJuRTE 'aEvut-SURVEII.IJutCE FREQUENCY SR 3.7.5.'1 J Verify each AFW manual, power operated, and automatic valve in each water flow path, 31 days in both steam supply f1 the to the steam turbine driven pump, t t is not locked, sealed, or otherwise secured in gp ,

J position, is in the correct position.

SR 3.7.5.2 -NOTE- -

I Notrequiriktobeperformedf the -

WI turbine pump until 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 8'Skih M' after 2: 1000 ig,in the steam-generator.

p9g. 4 h Verify the developed head of each AFW pump

& 'k31]daysona 6'

at the flow test point is greater than or STAGGERED TEST equal to the required developed head. BASIS [

b SR 3.7.5.3 I ot N applicable in

-- NOTE- ]

30003 4 when steam generator is relied upon for heat removal.j e -........ ..- ,

verify each AFW automatic valve that is IS months not locked, sealed, or otherwise secured -

in position, actuates to the correct position on an actual or simulated actuation signal.

(continued)

WDG STS 3.7 13 Rev 1, 04/07/95

AFW System 3.7.5 SURVIII. LANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.7.5.4 ---- - - - - - - - - - - - --- - - - - - - - - -

requi to be formed for the p bine de von AFW un< ;Jri * ,

(24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after 2: 1000 psig"in the e, team generator. - -

L-fg,1h

,\ D -

{2. Not applicable in MODE 4 when steam T generator is relied upon for heat g removal. J Verify each AFW pump starts automatically 1 months on an actual or simulated actuation signal.

SR 3. . Verify proper alignment of the required paths by verifying flow from eat.ering condensate e tank to e MODE 2, y

l generator. whenever unit has been in

~

w k $ fdf D6 bb lh Wb 8

% ar-w p3mp 5%.m een valves o u u> hen %*

Sufjoue ho ff*M m'"*' N*' gen' r w~~(

a e -

]

3 was STS 3.7-14 Rev 1, 04/07/95

FNP TS Conversion Enclosure 5 - JD from STS Chapter 3.7 - Plant Systems STS 3.7.5 AUXILIARY FEEDWATER (AFW) SYSTEM FNP ITS 3.7.5 AUXILIARY FEEDWATER (AFW) SYSTEM JD NUMBER JUSTIFICATION 1 The STS 3.7.5 AFW Mode 4 applicability is not included in the proposed FNP ITS consistent with the current FNP licensing basis as documented in the current FNP TS. No FNP design basis accident analysis assumes the operability of the AFW system in Mode 4. In Mode 4, with the heat load reduced and the RHR system available, the AFW system is not necessary to ensure adequate decay heat removal capacity. All references to the requirement to have AFW operable in Mode 4 are removed from the pmposed FNP ITS 3.7.5 LCO, Actions, Surveillance ;

,2 Requirements, and bases consistent with the current FNP TS. -

T@ 2 The STS note in each AFW pump Surveillance Requirement which modifies the d

T performance of the Surveillance for the turbine-driven AFW pump is revised consistent with the initial conditions of the current FNP surveillance test procedures for performance testing of the turbine-driven AFW pump. The FNP-specific initial condition for testing the turbine-driven AFW pump is based on RCS temperature at no-load conditions. The steam generator pressure requirement of 2: 1005 psig corresponds to the Reactor Coolant System no-load, hot standby, Mode 2 operating T.,, of 547 'F. The FNP-specific initial test condition utilizing pressure equivalent to the current RCS temperature requirements for testing of the turbine-driven AFW pump is acceptable based on the direct relationship between RCS temperature and SG pressure and the fact that operating experience has proven the current FNP test procedures adequate to verify the turbine-driven AFW pump performance.

3 The proposed FNP ITS 3.7.5 for the AFW system includes the FNP specific Surveillance Requirement SR 3.7.5.5 consistent with the AFW CTS. This.FNP surveillance verifies the capability of the air accumulators associated with the turbine-driven AFW pump steam admission valves to perform their intended safety function (open the valves). The proposed ITS surveillance is consistent with the AFW CTS surveillance.

4 The STS AFW SR 3.7.5.1 is revised by the addition of an FNP specific note consistent with the corresponding provision of CTS surveillance 4.7.1.2.2.a.2. The corresponding CTS surveillance contains a provision that provides an exception for verifying the full open position of the AFW flow control valves when those valves are being used for controlling SG level during low power operation (less than or equal to 10% RTP) or when the AFW system is not in automatic control, such as when AFW pump testing is performed at power. In both cases addressed in the exception, the AFW System, including the flow control valves, may be in manual control. The exception provided for the full open position verification of the flow Chapter 3.7 ES-1-E April,1999

Associated Package Changes for Chapter 3.9 O

r-Associated Package Changes for Chapter 3.9 RAI- 1 0

FNP TS Conversion Enclosure 2 - Discussion of Changes to CTS Chapter 3.9 - Refueling Operations CTS 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS ,

i FNP ITS 3.9.3 CONTAINMENT PENETRATIONS l DOC j HQ SHE DISCUSSION change involves consolidating related TS requirements into a single TS, it is I considered an administrative change. The details of the changes made to CTS _4.9.9 are identified in the markup of that CTS.

9 L The surveillance frequencies of Parts a and b of CTS surveillance 4.9.4 are revised consistent with the STS as follows:

The CTS frequency requiring the surveillance to be performed 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to entenng the Mode of applicability is deleted. The general rules for the use and application of surveillance requirements in the technical G,\ specifications (i.e., SR 3.0.4) require the surveillance to be successfully performed and current prior to entering the Mode of applicability. The general rules require the surveillance to be performed within 7 days prior to entering the Mode of applicability. Since there is little difference between 7 days (168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months ) and 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months , the existing requirement to perform these surveillances within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to entering the Mode of applicability is essentially #adant to the general rules requiring that the surveillance be successfully performed and current prior to entering the Mode of applicability. In both cases, whether performed 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to the start of Core Alterations or movement ofirradiated fuel in the containment building or within 168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months of the start of such evolutions, the surveillance must be met prior to entry into the Mode of applicability; the LCO must be met at the time the evolutions occur. In addition, once the status of the penetrations is verified, they remain under admmistrative control. Any changes to their status will be tracked to ensure that the requirements of the LCO (and associated SR) are met prior to entering the Mode of applicability. Therefore, the deletion of the 100 hour0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months preliminary frequency requirement for both CTS 4.9.4.a and b is acceptable.

The CTS requirement 4.9.4.b (4.9.9) requires testing the automatic containment purge and exhaust system isolation function every 7 days.

Consistent with the STS, this 7-day frequency is revised to once every 18 months. This change is consistent with the standard requirements for actuation testing of other ESFAS components. This change is reasonable considering the other existing testing that is required for the containment purge and exhaust isolation function. The remaining surveillance requirements are adequate to ensure the operability of the containment purge and exhaust isolation function. The surveillance requirements that are still applicable to this function are consistent with the surveillances Chapter 3.9 E2-3-D April,1999 h.

7 FNP TS Conversion Enclosure 2 - Discussion of Changes to CTS Chapter 3.9 - Refueling Operations CTS 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS FNP ITS 3.9.3 CONTAINMENT PENETRATIONS DOC N_Q SHE DISCUSSION required for ESFAS type functions and are therefore appropriate for the containment purge and exhaust isolation function. The surveillance requirements for the actuating instrumentation (manual and high radiation) are contained in the STS LCO 3.3.6, Containment Purge and Exhaust Isolation Instrumentation. These instrumentation surveillance requirements include a 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Channel Check (radiation monitors), a 31 day Actuation Logic Test and Master Relay Test, a 92 day COT, an 18 month Slave Relay Test, an 18 month TADOT (manual initiation), and an 18 month Channel Calibration and Response Time Test. Since the applicability of the STS instrument LCO 3.3.6 includes movement ofirradiated fuel assemblies and ,

Core Alterations, these instrumentation surveillances apply during j refueling. The replacement of the 7 day surveillance frequency with an 18 month requirement for the actuation testing of the containment purge and exhaust isolation function is acceptable based on the remaining surveillance requirements which adequately verify the containment purge and exhaust system isolation function operable in a manner and frequency consistent ;

with other ESFAS functions. i l

i I

i Chapter 3.9 E2-4-D March,1998 i

l l

Associated Package Changes for Chapter 3.9 RAI-3 and RAI- 4 l

m t- !

j.g REFUELING OPERATION.1 s[4.9 sf afBUAL MEAT e maVAL AND 1. ANT CIRCUI.ATION - N Of hVOi (AIL WATat L WrEr g _ _ _ .

LIMITING CONDITION FOR OPERATION Papstiard.

1.f 0 I ele Y

, b ,

Q.9.8.1 J - Qt lesej one residual heat removal loop shall be n operation.

6C8 NOTE 3* g APPLICABILITY: M005 6 W$O Og, tua.kt.f" lgvej g 13 [ pg, f ,[ =

AeTIoN: rent.+er vesse.1 %e.

_ar Mfuelange t.on0. A le 9CfE6 J PSfE tidA1./4 With less than one residual heat removal loon in aa-eation, exc'ept

/s it I as provided in b. below, eue 1 ope atfions inho14Ldr dahr4 Gncrease An/ *A=Iretezer escat Ilekt ^^d aEn reduction in boron Af d.'IM M ts concentration of the R'eactor coolant Sy'stemeir 51ose al d46Yb g.A bontainment penetratione providing direct access from the Makf4, ,

containment a* - --' re en *' -^ 'd- ** -- ----'=ce within 4 houre

^^ % = ITE -

foc+Nif1 TW The residual heat removaz Acop mer be removed r operation for "7C ._

up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months pe_c 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months periodguring the performance of coma %g manaTIwo in ene vactnssy as @ reactor =- - - - -i hot T' laae. L__

q.g l

e .

TP)4 proyLeione pff spec /ficat)6n 3.0/ are no[appif dfablyf.J j LA

,A SURVEILLANCE REQUIREMENTS

$R S.t %/ _ k VGAtFY ens RMUeep P 9.8. G/readddal hdat sismodi foog/eNtil befetdimined taf tin]Ln opera on and circulating reactor coolasit at a riow rate of greater than or equal to 3000 gpo at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

60 ope 8 8 M $E g dlgggfGdur.;b0Of Nf"C+'f 20I8 8 heron cowuitan. g. ,

.9 9

FARLEY-UNIT 1 3/4 9 9 AMENDNENT No. 26

(A CHAPTER 3.9 LNSERT J TO CTS PAGE 3/4 9-9 ACTIONS A.4, A.5, A.6.1 and A.6.2 CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.4 Close equipment hatch and 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months secure with four bolts.

M l A.5 Close one doorin each air 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months l lock. -

l l

M A.6.1 Close each penetration 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months providing direct access from gS'$ the containment atmosphere U to the environment with a manual or automatic !

isolation valve, blind flange, or equivalent.

QE A.6.2 Verify each penetration is 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months capable ofbeing closed by an OPERABLE Containment Purge and ExhaustIsolation System.

Chapter 3.9 Insert Page 1

I d

),3 REFUELINC OPERATIONS l.Q Id2!f jfATER LEVEL LIMITING CONDIT FOR OPERATIOli P

3.i.[ @t Two GreeMdgnh_ Residual I at Removal (RHR) loops shall be OPERA.LE "-

g -

APPLICABILITY: -

MODE 6 when the water [p level '- ~above g ..the_* top o_f the reactor _ k pressure vessel flange is less than 23 feet.

^*'***'

es N sh2 ash, . Ng s.4erg go.bove gon.b +M

+*P *9 resL%r vsa6eI Flusfe.

CDnD A / With less than the required loops o RAsLE, immediately initiate corrective action t etur red RHR loops to o-A.u .tato.w, w av ,...un. 34 s r, a.y .ipn. or ,p 1f1..,1 ,i.., +e n.t.. ,u .

SURVEILLANCE RSQUIREMENTS 9 .2 The r tred idual at al loops hallbe[deterufed I

.a SR 3.431 Ye.ri

,. - i.atto .....

one oglee Qdor coola.n+ Lt. s. Row ra.+e. er z aceo p os yp in m exer *&bn swL e.iewdshi1hearN.

so.S.s.1 vevi$ corm.t br der aMamen+ end,a. fed per o.Mla.hle. to +he. re9uired RNRpumj3 iha +is nof tri operguHon a,very, 7&. r 8 -

creduc.teninsacfwc

,d', Hen 8 - He MA L*y in apeend,on. Sus &open:h

(.nt berm cancentra: hen wnecL.:tet.g, o.,ncL i a.,

"#Y'*" I '8 N f" *" { % **P ' *fe t qg gpg ,-

. ~ . .. . . .; .. . .

9 t .. ,_ ..

afhVnopinalfe egi,rgestnr ocd.or searce fay 94 infper@e fg egh IUtm igh

~

g.g

[One RNR loop may be inoperable for up to 2 noure for performance of RNR,-

-(surveillancetesting.

A ,

v_- - - - -

a.nd.no MR1 s " " * * * "'** 7* 9

s*i<sbe en N deq herd W

i.c o z.9. s

.l_)i

, 4dekn' FARLEY-UNIT 1 3/4 9-10 i-AMENDMENT NO. 26 f

[.

/6 CHAPTER 3.9 INSERT K TO CTS PAGE 3/4 9-10 ACTIONS B.3, B.4, B.S.1 and B.S.2 l

CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.3 Close equipment hatch and 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months l secure with four bolts.

bHQ l

B.4 Close one doorin each air 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months )

lock. l i

l

'B.S.1 Close each penetration 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months I g7 k providing direct access from F the containment atmosphere to the environment with a manual or automatic isolation valve, blind flange, or equivalent.

9.E B.S.2 Verify each penetration is 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months capable ofbeing closed by an OPERABLE Containment Purge and ;

Exhaust Isolation System.

)

i Chapter 3.9 Insert Page

l l

FNP TS Conversion l Enclosure 2 - Discussion of Changes to CTS Chapter 3.9 - Refueling Operations CTS 3/4.9.8.1 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION - ALL WATER LEVELS FNP ITS - 3.9.4 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION - HIGH WATER LEVEL l DOC

! MQ SHE DISCUSSION 7 A The CTS action statement "c" regarding the provisions of specification 3.0.3 being not applicable is deleted consistent with the STS. This statement l

is no longer required as the STS LCO 3.0.3 clearly states that it is only ;

applicable in Modes I through 4. This change is administrative made to conform with the STS format and presentation of this information.

8 M The CTS action statement "b"is revised consistent with the STS. An additional restriction is added to the CTS allowance to remove the required RHR loop from service for up to one hour per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period. The new restriction prohibits operations that would cause a reduction of the RCS boron concentration during the time the required RHR loop is removed from service. The circulation of the reactor coolant provided by the RHR assists in maintaining a uniform boron concentration in the RCS. During the period that the required RHR loop is removed from service and no forced circulation exists, a uniform boron concentration in the RCS can not be assured. Therefore, the restriction to prohibit operations that would cause a reduction of the RCS boron concentration during the time the required RHR loop is removed from service is appropriate and applicable to FNP.

Since this change represents a new requirement, it is considered more restrictive.

9 L The requirement in CTS 3.9.8.1 Action Statement "a" to "close all

containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months " is revised consistent with the STS (as revised by TSTF-197, Rev.1). The CTS Action is revised g O to require actions consistent with the existing ITS LCO 3.9.3 requirements for containment penetrations. LCO 3.9.3 specifies the required status of the containment penetrations and does not require all penetrations be closed.

LCO 3.9.3 requirements may be met by a penetration " capable of being closed by an OPERABLE Containment Purge and Exhaust Isolation System." An operable containment purge and exhaust isolation system monitors the containment radioactivity and is capable of closing the purge and exhaust penetrations to prevent the release of radioactivity to the environment. As such, LCO 3.9.3 allows the purge and exhaust valves to be open if the associated isolation system is operable. This is acceptable Chapter 3.9 E2-3-1 April,1999

FNP TS Conversion Enclosure 2 - Discussion of Changes to CTS Chapter 3.9 - Refueling Operations CTS 3/4.9.8.1 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION - ALL WATER LEVELS FNP ITS - 3.9.4 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION - HIGH WATER LEVEL DOC HQ SEE DISCUSSION based on the fact that the purge and exhaust isolation valves will be closed automatically to mitigate the consequences of an event which releases r radioactivity to the containment atmosphere. In addition, LCO 3.3.6, Containment Purge and Exhaust Isolation Instrumentation, provides appropriate remedial Actions for an inoperable purge and exhaust isolation gO system. The STS Action, by requiring actions consistent with the existing ITS LCO 3.9.3 requirements for containment penetrations, eliminates any confusion or conflict which may be caused by different requirements being specified in the same Mode for the same equipment. As such, the STS Action maintains the consistency of containment penetration requirements within the TS.

Chapter 3.9 E2-4-I April,1999 l

l FNP TS Conversion Enclosure 2 - Discussion of Changes to CTS Chapter 3.9 - Refueling Operations CTS 3/4.9.8.2 - RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION - LOW l WATER LEVEL

' FNP ITS - 3.9.5 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION - LOW 4 WATER LEVEL DOC !

MQ SHE DISCUSSION the functional requirements of Mode 6 operation.

7 A The CTS 3.9.8.2 actions are revised consistent with the STS. Additional -

actions (Condition B, Actions B.1 and B.2) are added to CTS 3.9.8.2 for the Y condition of no RHR loop in operation. The additional actions are not new, but are consistent with the actions of CTS 3.9.8.1 and 3.9.8.2 for no RHR loop in operation. CTS 3.9.8.1 was applicable for all refueling cavity water I levels in Mode 6. Since CTS 3.8.9.1 has been revised to apply only when the water level is 2 23 feet above the reactor vessel flange it is necessary to add the requirement and actions for the RHR loop in operation to CTS 3.9.8.2 (water level < 23 feet above vessel flange) to preserve the existing TS requirement and actions for the RHR loop required to be in operation in l Mode 6 at all water levels. Since this change is made to conform with the presentation and format of the STS (two separate water level TS) and has no technical impact, it is considered administrative.

7a L The CTS 3.9.8.2 actions are revised consistent with the STS (as revised by TSTF-197, Rev.1) by the addition of Actions B.3, B.4, B.5.1, and B.S.2.

The CTS is revised to require actions consistent with the existing ITS LCO 3.9.3 requirements for containment penetrations. LCO 3.9.3 specifies the required status of the containment penetrations and does not require all penetrations be closed. LCO 3.9.3 requirements may be met by a -

penetration " capable of being closed by an OPERABLE Containment Purge and Exhaust Isolation System." An operable containment purge end d exhaust isolation system monitors the containment radioactivity and is gfI capable of closing the purge and exhaust penetrations to prevent the release 4 r of radioactivity to the environment. As such, LCO 3.9.3 allows the purge and exhaust valves to be open if the associated isolation system is operable.

This is acceptable based on the fact that the purge and exhaust isolation valves will be closed automatically to mitigate the consequences of an event which releases radioactivity to the containment atmosphere. In addition, LCO 3.3.6, Containment Purge and Exhaust Isolation Instrumentation, provides appropriate remedial Actions for an inoperable purge and exhaust isolation system. The STS Action, by requiring actions consistent with the existing ITS LCO 3.9.3 requirements for containment penetrations, eliminates any confusion or conflict which may be caused by different

.. Chapter 3.9 E2 3-J Apdl,1999

FNP TS Conversion Enclosure 2 - Discussion of Changes to CTS Chapter 3.9 - Refueling Operations CTS 3/4.9.8.2 - RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION - LOW WATER LEVEL FNP ITS - 3.9.5 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION - LOW WATER LEVEL DOC HQ SHE DISCUSSION requirements being specified in the same Mode for the same equipment. As ,

such, the STS Action maintains the consistency of containment penetration

. requiremeiits within the TS. As this change adds an alternative to " closing all contamment penetrations ... within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months " as contained in CTS 3.9.8.1, it is considered less restrictive.

8' A The CTS

footnote regarding the status of the normal or emergency power source for the RHR is deleted consistent with the STS. The footnote specified that one of the power sources for the RHR may be inoperable.

Since in the STS, the definition of operability is revised to state that only one (normal gr emergency) power supply is necessary, the allowance provided by the CTS

footnote is no longer required. Since this change is made to confonn with the presentation and format of the STS and has no technical impact, it is considered administrative.

The FNP specific CTS

footnote to LCO 3.9.8.2 is revised to clearly state 9 L the required exception and conform with the changes made to the LCO by the STS. The current CTS footnote provides an exception to the LCO requirement for two operable RHR loops for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for RHR surveillance testing. This CTS note is revised to also include an exception to the additional STS LCO requirement for one RHR loop to be in service (decay heat removal mode of operation) for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for RHR testing.

In addition, the FNP specific CTS footnote is moved up to the LCO in the FNP ITS consistent with the STS presentation of LCO exception notes.

This FNP surveillance test exception is retained in the FNP ITS to ensure accurate and complete testing of the RHR system in the ECCS injection mode of operation while the unit is shutdown.

The Farley design minimum calculated flow rate for the Residual Heat Removal / Low Head Safety Injection (RHR/LHSI) System in the cold leg injection mode considers the worst case single failure of an entire ECCS

rain and therefore, requires a flow of 3981 gpm from one RHR pump. This single pump performance requirement is currently specified in CTS surveillance 4.5.2.i and must be performed during shutdown conditions. In order to verify the required flow for a single RHR pump, the pumps must be tested one at a time.

Chapter 3.9 E2-4-J April,1999

FNP TS Conversion Enclosure 3 - Significant Hazards Evaluations Chapter 3.9 - Refueling Operations III. SPECIFIC SIGNIFICANT HAZARDS EVALUATIONS CTS 3/4.9.8.2 - RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION - LOW WATER LEVEL FNP ITS - 3.9.5 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION - LOW WATER LEVEL 7a-L

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

The proposed change involves changing the CTS 3.9.8.2 action requirements for the RHR system during refueling to more closely agree with the STS requirements and does not result in any hardware changes. The proposed change provides an allowance to the CTS Action (of CTS 3.9.8.1 which is applicable with no RHR loop in operation) which requires all penetrations to the outside atmosphere to be closed. The STS Actions allows the Containment purge and exhaust valves open if they are capable of being closed by an

'q automatic isolation system which is consistent with LCO 3.9.3, " Containment Penetrations."

The revised Actions pertain to an inoperable RHR loop. The RHR system is not assumed to be an initiator of any analyzed event. The role of the RHR system during refueling is to provide decay heat removal and reactor coolant boron mixing. The proposed change does not impact the RHR system capability to provide the required cooling and mixing functions as described in the FSAR nor does it impact the results of the analyses in the FSAR which remain bounding. The proposed change assures the containment penetrations are maintained consistent with the requirements of LCO 3.9.3 which in turn ensures the capability of the purge and exhaust valves to be isolated to mitigate an event that releases radioactivity to the containment atmosphere. Additionally, the proposed change does not impose any new safety analyses limits or alter the plants ability to detect and mitigate events. Therefore, this. change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

The proposed change involves changing the TS requirements for the RHR system during refueling to more closely agree with the STS requirements and does not necessitate a physical alteration of the plant or changes in parameters governing normal plant operation. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

The proposed change, which revises the TS requirements for the RHR system during refueling to be consistent with the STS requirements does not involve a significant reduction N

r3.9 E3-5-D Apn!,1999

1

[

FNP TS Conversion Enclosure 3 - Significant Hazards Evaluations Chapter 3.9 - Refueling Operations III. SPECIFIC SIGNIFICANT HAZARDS EVALUATIONS j CTS 3/4.9.8.2 - RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION - LOW WATER LEVEL FNP ITS - 3.9.5 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION - LOW n' -

WATER LEVEL in a margin of safety. The proposed change provides an allowance to leave the containment purge and exhaust valves open if they are capable of being isolated by an operable automatic l

isolation system. The STS Action continues to ensure the containment purge and exhaust l

penetrations are capable of being closed automatically to mitigate the consequences of an i event which results in the release of radioactivity to the containment atmosphere. The .

proposed change does not impose any new safety analyses limits or alter the plants ability to l detect and mitigate accidents currently described in the FSAR. As such, any reduction in the margin of safetyis insignificant.

i l ,

i l

Chapter 3.9 E3-6-D April,1999

r Containment Penetrations 3.9.g 3

3.9 REFUELING OPERATIONS I 3

3.9.f Containment Penetrations 3

LC0 3.9./ The containment penetrations shall be in the following I status: ;

a. The equipment hatch closed and held in place by gfour bolts; _=

g' psiy '

4 b. UnTdoor'in each air lock closed; and - &n vsronment c.; Each penetration providing direc<; neces from the A I containment atmosphere to the glu1;s de atmospherc either: "

1. closed by a manual or automatic isolation valve, blind flange, or equivalent, or 1

2. capable of being closed by an OPERA 8LE Containment

' Purge and Exhaust Isolation System.

! I APPLICABILITY: During CORE ALTERATIONS, During movement of irradiated fuel assemblies within containment.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TINE A. One or more A.1 Suspend CORE Immediately containment ALTERATIONS.

penetrations not in required status. glQ A.2 Suspend movement of Immediately irradiated fuel assemblies within containment.

-worsts- 3.9-6 Rev 1, 04/07/95 FN P un A 1.~d. 2.

l l%O l RHR and Coolant Circulation - High Water Level I 3.9.g ACTIONS i CONDITION REQUIRED ACTION CONPLETION TINE T^

A. (continued) A.4 FClos all c tai nt 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months pe trati s Nev. I gO cce conta ment ro at phere o ou ide -- .

a spher . ji ,

N SURVEILLANCE REQUIRENENTS SURVEILLANCE FREQUENCY SR, 3.9. 1 Verify one RHR loop is in operation and 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months circula" ro reactor coolant at a flow rate of k Q Q )9pa.

/

b

. W0tr-44 3.9-9 Rev 1, 04/07/95 FNP unoh tQ 2-

f; 12 0 CHAPTER 3.9 l

INSERT L TO STS PAOE 3.9-9 ACTIONS A.4, A.5, A.6.1 and A.6.2 CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.4 Close equipment hatch and 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months secure with four bolts.

M A.5 Close one doorin each air 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months

lock.

1 M -

A.6.1 Close each penetration 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months providing direct access from ,

the containment atmosphere gT[$ to the environment with a P manual or automatic isolation valve, blind flange, or equivalent.

DE A.6.2 Verify each penetration is 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months >

capable ofbeing closed by an OPERABLE Containment Purge and Exhaust Isolation System.

4 1

Chapter 3.9 Insert Page

11.b RHR and Coolant Circulation - Low Water Level 3.9.g ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.2 Initiate action to Immediately restore one RHR loop to operation, m Y B.3 bloseallcontainment 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months penetrations

- providing direct .

access from

[9 . containment atmosphere to outside atmosphere. j i

3nsegy y T5TF R7 SURVEILLANCE REQUIREMENTS Rev. t SURVEILLANCE FREgJENCY 5"

SR 3.9./.1 Verify one RHR loop is in operation and 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months circulating reactor coolant at a flow rate ofa:

S Verify correct breaker alignment and SR 3.9.f.2 7 days indicated power available to the required RHR pump that is not in operation.

w 3.9-11 Rev 1, 04/07/95 FNP Omts idl

I21.s CHAPTER 3.9 INSERT M TO STS PAGE 3.911 ACTIONS B.3, B.4, B.5.1 and B.S.2 i l

1 CONDITION REQUIRED ACTION COMPLETION TIME l B. (continued) B.3 Close equipment hatch and 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months secure with four bolts.

1 M

B.4 Close one doorin each air 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months lock.

M 7' B.S.1 Close each penetration 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months providing direct access from the containment atmosphere to the environment with a manual or automatic isolation valve, blind flange, or equivalent.

DE B.5.2 Verify each penetration is 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months capable ofbeing closed by an OPERABLE Containment Purge and Exhaust Isolation System.

Chapter 3.9 Insert Page

1 lYla Containment Pen 3trations l 83.9./3 83.9 REFUELING OPERATIONS 3 1 i

83.9./ Containment Penetrations BASES BACKGROUND During CORE ALTERATIONS or movement of irradiated fuel assemblies within containment, a release of fission product i radioactivity within containment will be restricted from l escaping to the environment when the LCO requirements are

[pV S met. In MODES 1, 2, 3, and 4, this is accomplished by l

maintaining containment-0PERA8LE as described in LC0 3.6.1, T5QWh Rev l = Containment." In MODE 6, the potential for containment

=r

@mlli pressurization as a result of an accident is not likely; N y+fbutsiae atmosphere 3can beTheless LCO /'

stringent.the

'^

-l

" requirements are referred to as 'conuJammvocop rather l Than " containment 0PERABILITY.* demenW ,os5 ans h.t.j sde Lu . hat all potential escape paths are closed or capab e of rem % il wing closed. Since there is no potential for containment pressurization, the Appendix J 1eakage criteria and tests Wkt

'P6E h are not required. Q o g go} g TM containment serves to contain fission product rad oactivity that may be released from the reactor core following an accident, such that offsite radiation exposures are maintained well within the requirements of 10 CFR 100.

Additionally, the containment provides radiation shielding from the fission products that may be present in the containment atmosphere following accident conditions.

The containment equipment hatch, which is part of the containment pressure boundary, provides a means for moving large equipment and components into and out of containment.

During CORE ALTERATIONS or movement of irradiated fuel assemblies within containment, the equipment hatch must be held in place by at least four bolts. Good engineering practice dictates that the bolts required by this LC0 be approximately equally spaced.

The containment air locks, which are also part of the containment pressure boundary, provide a means for personnel access during MODES 1, 2, 3, and 4 unit operation in accordance with LCO 3.6.2, " Containment Air Locks." Each air lock has a door at both ends. The doors are normally interlocked to prevent simultaneous opening when containment OPERA 8ILITY is required. During periods of unit shutdown (continued)

WOG STS 8 3.9-11 Rev 1, 04/07/95

le Containment Penetrations .

B3.9./

BASES 7 A)

'. [ [

8ACKGROUND when(cphraMm6stA>ro9 erd is not required, the door interlock (continued) mechanism may be disabled, allowing both doors of an air lock to remain open for extended periods when frequent containment entry is necessary. During CORE ALTERATIONS or movement of irradiatpd fuel assemblies within containment,

-w -- w rumn s required; therefore, the door interl ock mechanism may remain disabled, but one air lock %

door must always remain closedu - .

The requirements for CUUNundantexwuc.rosure ensure f / '

c.

. that a release of fission product radioactivity within containment will be restri ped from escaping to the environment. ThemasurmPestrictions are sufficient to '

% h. restrict fission product radioactivity release from -

psc ontainment due to a fuel handlina ur h + during refueling.

sl8 M"/)gY@ GAM'

9. , ('C The Containment Pt ge and Exhaus'. y en in ude t subsystems. The e ormal subsystem I include a c purge in n ese M Es penetration and a W inch exhaust nenetr ion. The second Cin$ % $23 subsystem,la minipurge, system, include an 8 inch purge M

,.f s m ur= = oand an 8 . inch.exhausth en rations'lDuring

%g 6 'NS 1. 2. 3, and 4, the two valve in ~each of the normal purge and exhau rations ar secured in the closed ines position. The twonva n each of the two minipurge .

g

. - 2 m be\open_ . but are closed

~

'I

. ically by the Engineered Safety Features._Actu,at}.oo pA rostrumen'Ikk. n N h am (ESFAS)p Neith'er'of the subsystems is subject to a 4 D *,,Io

(!f

{M g sa m can on in MODE 5. -

R M l In MODE 6, large air exchang rgnecessary to conduct rs onInsth>werd'-l. refueling operations. The normal inch purge system is

"#*0a -. Mu or this purpose, and all fou)r(valves are closed by the n rusm. with 00 3. .Z. "E Ineer a saf ty j .wdtionfystem Act ESFA Inst nt on." b mi purge sys en resa etional 1 o ope

( four v ves are a soclos by ESFAS./ MODE 6, dd allb) .

p pc ' ,The miniource system is not sed in MODE 6. @ fgur/8 /ng5)

JaWes art sepGre9'In tpe Mose(oMitio _ _

The other containment penetrations that nrovida direct _

c_ cess from containment atmosphere to 6utside atmoso tert - hM Ho@ever, iF1}>e inimg, R. , T GJ e &mg N t .'hauo-

+h capo.bleoF n- --

3 ug v a - .

WOG'STS B 3.9-12 Rev 1, 04/07/95 f/CEi

I Wl Containment Penetrations B3.9./

1 BASES &

a.

dD i l

BACKGROUND aust be iso ed ori a ast one side. Isolation ma be

_ (continued) achieved by automatic isolation valve, a manual isola on va ve, blind flanaa or mouivalent -

Equivalent isolation methodhFdt ble adordvdd I M *i*b5 P i inciuae use or a materias tnat can provide a t -

may

)

of so e So.h=

atmospheric pressure, ventilation barrier forhorary, other E containment _penetratinos durinamuel/mevemenvu(Raf I b"N _NN m f E..x.. x _ N..1... m . ...

Y b,,..E. a. p m

APPLICA8tE During CORE ALTERATIONS or movemed of irradIat h" SAFETY ANALYSES assemblies within containment, the most s radiological consequences result from a fuel han accident. The fuel.

handling accident is a postulat vent that involves damage to irradiated fuel (Ref 2). uel handling accident @ A g, d i analyzed lfn Reference 371ncludeFdropping a single v g irradiated fuel assempivfanp nand1/ng Jooi or/a heavy ob e fonto orner irrad ates fuel /assembYieur ine requirements o -

LCO 3.9.7, " Refueling cavrty water Level," and the minimum decay time of 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to CORE ALTERATIONS ensure that the release of fission product radioactivity, subsequent to a fuel handling accident, results in doses that are well within the guideline values specified in 10 CFR 100 Standard Review Plan, Section 15.7.4, Rev. 1 (Ref. 3), defines "well within" 10 CFR 100 to be 25% or less of the 10 CFR 100 values. The acceptance limits for offsite fac$ofradiationexposurep$

ill be 25% of 10 CFR 100 value:

%3

)

Containment penetrations satisfy Criterion 3 of the NRC Policy Statement.

LCO This LC0 limits the consequences of a fuel handling acciden -#

in containment by limiting the potential escape paths for ennfeetred T$TF-897

.. fissionproductrad-ioactivityreleasedwithincontainment[.D 4ing direct access The from LC0 requires any the containment penetration atmosphere to th prov;Jtrutside atmospherB Re.i.

to CINSERT'_E

..-- be closed except for the OPERABLE containment purge and exhaust penetrationsJ For the OPERABLE containment purge F3 and exhaust penetrations, this LCO ensures that these IFY penetrations are isolable by the Containment Purge and i ust Isolation System. The OPERABILITY requirements for Tiis LC0jensure that the automatic purge and exhaust valve _

54,hnfSnmenfbrye &ZsoONZnsfrunen e __ __ -- v --- - m........,

WOG STS B 3.9-13 Rev 1, 04/07/95

n

]

19 Containment Penetrations B 3.9./

3 BASES LCO closure times specified in the FSAR can be achieved and, (continued) therefore, meet the assumptions used in the safety analysis to ensure that releases through the valves are terminated, such that radiological doses are within the acceptance limit.

APPLICABILITY The containment penetration requirements are applicable during CORE ALTERATIONS or movement of irradiated fuel assemblies within containment ttecause this is when there is a potential for a fuel handling accident. In MODES 1, 2, 3, and 4, contairusent penetration requirements are addressed by.

LC0 3.6.1. In MODES 5 and 6, when CORE ALTERATIONS or movement of irradiated fuel assemblies within containment are not being conducted, the potential for a fuel handling accident does not exist. Therefore, under these conditions

'no requirements are placed on containment penetration status.

T&TF ACTIONS A.1 and A.2 trkviroA M-- 8 d j N , ,

p.,g If the containment equipment hatch, a locks, or any containment penetration th t rov de tract access fr.on the p p 3 containment atmosphere.to t us is'n'ot in the required status, including tne t.ontainment Purge and M Exhaust Isolation System not capable of automatic actuation (

when the purge and exhaust valves are open, the unit must be ;

Placed in a condition where' the isolation function is not needed. This is accomplished by ismediately suspending CORE ALTERATIONS and movement of irradiated fuel assemblies within containment. Performance of these actions shall not preclude completion of movement of a component to a safe position.

SURVEILLANCE SR 3.9 '

.1 REQUIREMENTS This Surveillance demonstrates that each of the containment penetrations required to be in its closed position is in that position. The Surveillance on the open purge and exhaust valves will demonstrate that the valves are not blocked from closing. Also the Surveillance will (continued)

WOG STS B 3.9-14 Rev 1, 04/07/95

l%

RHR and Coolant Circulation - High Water Level l B3.9./ i

'l BASES ACTIONS AJ. (continued) water level 2: 23 ft above the top of the reactor vessel l flange, corrective actions shall be initiated immediately.

' .6, M .l' w A A 4 L.

75TF-197Al -

P If RHR loop requirements are not met, all containment MaLTF penetrations providing direct access from the containment atmosphere to the outside atmosphere must be closed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . With the P}iR loop requirements not met, the pr3 potential exists for the coolant to boil and release radioactive gas to the containment atmosphere. Closing containment penetrations that are open to the outside atmosphere ensures dose limits are not exceeded. l The Com:1stion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is reasonable, based on the Qow pro > ability of the coolant boiling in that time.

i SURVEILLANCE SR 3.9. '

1 REQUIREMENTS This Surveillance demonstrates that the RHR loop is in operation and circulating reactor coolant. The flow rate is j determined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thermal and -

boron stratification in the core. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months !

is sufficient, considering the flow, temperature, pump control, and alarm indications available to the operator in the control room for monitoring the RHR System.

j l

REFERENCES 1. FSAR, Section .5.7 i

WOG STS B 3.9-20 Rev 1, 04/07/95

/67 CHAPTER 3.9

INSERT F TO ITS BASES PAGES B 3.9-20 AND B 3.9-23 BASES FOR RHR HIGH AND LOW WATER LEVEL ACTIONS (TSTF-197, Rev.1)

.h If no RHR is in operation, the following actions must be taken:

a) . the equipment hatch must be closed and secured with four bolts; b) one door in each air lock must be closed; and i

c) each penetration providing direct access from the containment atmosphere to the environment j must be either closed by a manual or automatic isolation valve, blind flange, or equivalet', or verified to be capable of being closed by an OPERABLE Containment Purge and Exhaust Isolation System. :

With RHR loop requirements not met, the potential exists for the coolant to boil and release radioactive gas to the containment atmosphere. Performing the actions described above ensures that all containment penetrations are either closed or can be closed so that the dose limits are not exceeded.

)

The Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months allows fixing of most RHR problems and is reasonable, based on the low probability of the coolant boiling in that time.

l I

Chapter 3.9 Insert Page

n IM RHR and Coolant Circulation - Low Water Level B3.9./

S BASES u ee r ACTIONS (continued) \.in th* C0LR * -

~

on Pec$h '

If no loop is in operation, there will be no forced circul tion to provide mixing to establish uniform bor conce trations. Reduced boron concentrations ca o occur h refore, y tL addition of water with a lower boron concen ration M n'T.nat containea in tne xts,. because all of Ine unoorateyd h __ er ' sources are isolatedJ

/ n +icoud.

he imdd. Phon result Acs en e bomn

' Cerdrahn of Mer b$ tothe. If no RHR flloop is in operation, actions shall 'be initiated .

{

+6.n the rerdwed, >immediately, and continued, to restore one RHR loop to operation. Since the unit is in Conditions A and 0

! borencone.

err concurrently, the restoration of two OPERABLE RHR loops and b ,d m Me, one operating RHR loop should be accomplished expeditiously.

LR must be.

h* G Sf no RHR loop is in operation, all containment penetrations

~ '

providing direct access from the containment atmosphere to the outside atmosphere must be closed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . With IN6Egr F the RHR loop requirements not met, the potential exists for

~

the coolant to boil and release radioactive gas to the containment atmosphere. Closing containment penetrations TSTF-/97/ that are open to the outside atmosphere ensures that dose i Rey, l flimitsarenotexceeded.

%Td (The Completion low pro > ability ofTime of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the coolant is reasonable, boiling in that time. based on the S

SURVEILLANCE SR 3.9.t.1 REQUIREMENTS This Surveillance demonstrates that one RHR loop is in operation and circulating reactor coolant. The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thennal and boron stratification in the core. In addition, during operation of the RHR loop with the water level in the vicinity of the reactor vessel nozzles, the RHR pump suction requirements must be met. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient, considering the flow, temperature, pump control, (continued)

. . WOG STS B 3.9-23 Rev 1, 04/07/95

p j

/4t

. CHAPTER 3.9

! INSERT F TO ITS BASES PAGES B 3.9-20 AND B 3.9-23 l

BASES FOR RHR HIGH AND LOW WATER LEVEL ACTIONS (TSTF-197, Rev.1)

If no RHR is in operation, the following actions must be taken:

a) the equipment batch must be closed and secured with four bolts; b

k c))each onepenetration door inproviding each air directlock must access becontainment from the closed; atmosphere and to the environment must be either closed by a manual or automatic isolation valve, blind flange, or equivalent, or !

verified to be capable of being closed by an OPERABLE Containment Purge and Exhaust I

. Isolation System.

1 With RHR loop requirements not met, the potential exists for the coolant to boil and release l radioactive gas to the containment atmosphere. Perfonning the actions described above ensures that all containment penetrations are either closed or can be closed so that the dose limits are not exceeded.

The Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months allows fixing of most RHR problems and is reasonable, based I on the low probability of the coolant boiling in that time.

(

Chapter 3.9 Insert Page

m I

1

?

l j l l

l l

t l

l ATTACHMENTIII SNC Identified Editorial Changes :

Associated Package Changes I

p 1

1 I

i The following change has been made to the Chapter 3.1 submittal to address editorial changes, i omissions, and inconsistencies in the package- ,

! 1. STS Bases page B 3.1 14 for ITS LCO 3.1.2, Core Reactivity, is revised to refer to the

"beginning of cycle life (BOL)" consistent with the response to RAI 3.1.3-2. ,

l I l

h following change has been made to the Chapter 3.5 submittal to address editorial changes, l omissions, and inconsistencies in the package: l I

l 1. STS Bases Inserts C and D for ITS LCO 3.5.5, Seal Injection Flow, are revised to refer to l " pressurizer pressure"instead of"RCS pressure" since the graph is based on the differential pressure between the charging header pressure and the pressurizer pressure. .

The following changes have been made to the Chapter 3.7 submittal to address editorial changes, ,

omissions, and inconsistencies in the package: I

1. The heading for the DOCS related to CTS 3/4.7.1.4, Activity, incorrectly referred to ITS 3.7.14. 'Ihe heading for these DOCS has been corrected to refer to ITS 3.7.16.

2. JD-5 related to ITS 3.7.2, Main Steam Isolation Valves (MSIVs), incorrectly referenced a closure time of 5 seconds for surveillance testing and 7 seconds for ESF response time.

l

'Ibese numbers are changed to 7 and 9 respectively. In addition, JD-5 refers to FSAR Table 7.3.6. b corrected reference is 7.3.16.

l l

1_

y,,

Chapter 3.1 G

/ 5'8 Core Reactivity B 3.1 BASES nTF-r%

APPLICA8LE behavior and the RCS boron concentration requirements for SAFETY ANALYSES reactivity control during fuel de "

g (continued)

The comparison between measu da 1 fe..'( 8 0 m ed P+-f . itial core h

'r reactivity provides a nonnal zation for the calculational models used to predict core activity. Jf the measured and predicted RCS boron concentrat dentical core conditions at beginning of cyc1 do not agree, then I the assumptions used .in the relo cycle design analysis or I[L, the calculational models used to predict soluble boron N requ nts may not be accurate. If reasonable agreement n measured and predicted cope reactivity exists at then the prediction may be normalized to the measured ron concentration. Thereafter, any significant deviations

[/ in the measured boron concentration from the predicted boron letdown curve that develop durinq fuel depletion may be an indication that the ulationa' model is not adequate for core burnups beyond or that an unexpected change in core conditions has cc The normalization of p,dered RC boron concentration to the measured value is typically rformed after reaching RTP following startup from a refuels tage, with the control rods in their normal position <

ower operation. The normalization is performed at peonditions,sothatcore reactivity relative to predictec values can be continually monitored and evaluated as core conditions change during the cycle.

Core reactivity satisfies Criterion 2 of the NRC Policy Statement.

LCO Long term core reactivity behavior is a result of the core physics design and cannot be easily controlled once the core design is fixed. During operation, therefore, the LCO can only be ensured through measurement and tracking, and

..g appropriate actions taken as necessary. Large differences W, , between actual and predicted core reactivity may indicate that the assumptions of the DBA and transient analyses are no longer valid, or that the uncertainties in the Nuclear Design Methodology are larger than expected. A limit on the reactivity balance of

1% ok/k has been established based on engineering judgment. A 1% deviation in reactivity from (continued)

WOG STS B 3.1-14 Rev 1, 04/07/95 l

l

F 1

(

Chapter 3.5 9

l l

l

m 3 i

CHAPTER 3.5

/

INSERT C TO LCO 3.5.5 BASES ;

established by adjusting the reactor coolant pump seal injection needle valves to provide a total seal injection flow in the Acceptable Region of Figure B 3.5.5-1 at a given pressure differential

)- between the charging header pressure and the pressurizer pressure.

Eokb I i

i I

1 l

1 i

I t

I Chapter 3.5 Insert Page

f:

/ I t ._

CHAPTER 3.5

( INSERT D ;

TO SR 3.5.5.1 BASES l

l A differential pressure that is above the reference minimum value is established between the l l charging header (PT-121, charging header pressure) and the pressurizer, and the total seal l

i injection flow is verified to be within the limits determined in accordance with the ECCS safety analysis. j i E A 4 ;.l I

Chapter 3.5 Irdert Page p

r-l Chapter 3.7

I FNP TS Conversion l Enclosure 5 - JD from STS Chapter 3.7 - Plant Systems STS 3.7.2 MAIN STEAM ISOLATION VALVES l

FNP ITS 3.7.2 MAIN STEAM ISOLATION VALVES (MSIVs)

JD NUMBER JUSTIFICATION Therefore, plants designed with a single MSIV per steam line only have a " separate Condition entry" allowance for STS 3.7.2 Condition C (Modes 2 and 3). The FNP dual MSIV design allows the availability of the MSIV isolation function to be assured for each steam line in Mode I with a single inoperable MSIV as well as Modes 2 and 3. The application of the " Separate Condition entry" note is further l justified for FNP due to the short Completion Time (4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ) provided for the :

proposed FNP ITS Conditions B and E which address two inoperable MSIVs in a ,

single steam line. FNP ITS Conditions B and E address a loss ofisolation function in a steam line and effectively correspond to STS 3.7.2 Conditions A and C for a !

plant with a single isolation valve. However, the proposed Completion Time of FNP ITS Conditions B and E is more restrictive than the STS Condition A and C Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The FNP ITS 4-hour Completion Time provides an adequate limitation on plant operation to assure appropriate action is taken to either restore the MSIV isolation function in the affected steam line, reduce power, or isolate the affected steam line.

4 The STS surveillance SR 3.7.2.1 is revised consiwt with the corresponding CTS surveillance. CTS surveillance 4.7.1.5 is performed pursuant to Specification 4.0.5.

CTS Specification 4.0.5 references the Inservice Testing and Inspection Programs.

The STS does not have a corresponding specification to the CTS 4.0.5. The STS surveillance SR 3.7.2.1 provides the option to reference the Inservice Testing Program directly in the frequency column. As the CTS referenced the Inservice Testing Program (via 4.0.5) the STS surveillance option referencing the same program is selected to be consistent with the CTS. .

5 STS SR 3.7.2.1 is revised consistent with the corresponding CTS surveillance l requirement 4.7.1.5. This surveillance verifies the closure time of the MSIVs. The CTS surveillance requirement is performed in accordance with the Inservice Testing s Program (Specification 4.0.5). The Inservice Testing Program does not require that valve closme time be measured using a simulated or actual (ESF) actuation signal.

The Inservice Testing program allows the associated manual handswitch to be used when verifying valve closure time. The required closure time in this FNP !

surveillance (7 seconds) is not the ESF Response Time (9 seconds) which is l specified in FSAR Table 7.3-16 and required to be tested by ESFAS LCO 3.3.2, ESF Response Time surveillance. The ESF Response Time testing requires the ,

channel response from the sensor to and including the actuated component be measured. 'Ihe CTS MSIV surveillance 4.7.1.5 is intended to meet the requirements of the Inservice Testing Program to detect valve degradation over time l

l by verifying valve closure within 7 seconds. The time required to be met by the

'bpter 3.7 ES-4-B April,1999

FNP TS Conversion Enclosure 5 - JD from STS Chapter 3.7 - Plant Systems STS 3.7.2 MAIN STEAM ISOLATION VALVES FNP ITS 3.7.2 MAIN STEAM 50LATION VALVES (MSIVs) ,

.N - JD NUMBER JUSTIFICATION ,

CTS surveillance does not include the two additional seconds of the ESFAS l 6pl Response Time requirement (9 seconds specified in FSAR Table 7.3-16) for the main steam isolation function which accounts for the associated ESF actuation l electronics and sensor. Therefore, deletion of the STS requirement to use a ,

simulated or actual (ESF) signal in SR 3.7.2.1 to measure the MSIV stroke time is l acceptable and consisteht with the intent of the corresponding CTS surveillance and the Inservice Testing Program which require the valve closure time be measured but do not specify an actuation signal be present or simulated in order to accomplish the measurement.

t l

j t

\

i l

l

~

Chapter 3.7 ES-5-B April,1999 l I

j

c l

FNP TS Conversion Enclosure 2 - Discussion of Changes to CTS l Chapter 3.7 - Plant Systems g CTS 3/4.7.1.4 ACTIVITY l FNP ITS 3.7.16 SECONDARY SPECIFIC ACTIVITY l

DOC EQ SHE DISCUSSION 1 A The CTS 3/4.7.1.4 Action statement is revised consistent with the STS.

The specific activity limit contained in the CTS Action statement repeats the LCO requirement and is deleted. Since this limit is specified in the LCO, the Action is revised to simply reference the " limit". This revision

~

does not introduce a technical change and is made to confonn with the format and presentation of the STS. Therefore, this change is considered

]

administrative.

2 M The CTS surveillance 4.7.1.4 is revised consistent with the STS. The CTS surveillance is simplified to verify the LCO limit (0.10 ci/gm Dose Equivalent I-131) is met every 31 days. In addition, the CTS Table 4.7-2 and references to it are deleted from the CTS. The CTS Table 4.7-2 contained conditional frequencies for the performance of surveillance 4.7.1.4 based on the results of a more frequent gross activity determination.

These CTS conditional frequencies are deleted consistent with the STS.

The frequency of surveillance 4.7.1.4 is revised to a straight 31 days without conditions based on the gross activity level. The requirement to perform this surveillance at the more restrictive frequency of every 31 days provides additional assurance the LCO limit is met. However, this change results in a requirement to always perform this surveillance at 31 day j intervals instead of the CTS allowance for 6 month intervals (depending on I the results of the gross reactivity determination). Therefore, this change is considered more restrictive. .

3 L The CTS surveillance 4.7.1.4 is revised consistent with the STS. The CTS :

surveillance requirement on Table 4.7-2 to perform a gross activity determination every 72 days is deleted. 'Ibe results of the gross activity surveillances are used in the CTS to determine the frequency at which the Dose Equivalent I-131 surveillance must be pufunued (31 days or 6 nonths). In the applicable dose analyses, the radioiodines and the resulting tb /roid dose are limiting not the noble gases and the whole-body dose. As rach, the Specific Activity LCO requirement is based on Dose Equivalent I-131 not gross activity. Therefore, the deletion of the CTS gross activity requirement is acceptable based on the more restrictive requirement imposed on the Dose Equivalent I 131 surveillance which must always be performed every 31 days regardless of the gross activity. The Dose Equivalent I-131 surveillance performed every 31 days pmvides adequate Chapter 3.7 E2-1-D April,1999 m J

(, < . y i c FNP TS Conversion Enclosure 2 - Discussion of Changes to CTS

) Chapter 3.7 - Plant Systems N

./

y CTS 3/4.7.1.4 ACTIVITY 4l FNP ITS 3.7.16 SECONDARY SPECIFIC ACTIVITY DOC N_Q SEE DISCUSSION assurance that the LCO limit is being met and that the assumptions of the applicable dose analyses regarding radioiodines and thyroid dose remain ,

valid.

f i

d Chapter 3.7 E2-2-D April,1999

ML20206H000

Text

Navigation menu

Search