ML20236T791

Proposed Tech Specs Revising MSL Leakage Requirements & Eliminating Msli Valve Leakage Control Sys

ML20236T791
Person / Time
Site:	Perry
Issue date:	07/22/1998
From:	CENTERIOR ENERGY
To:
Shared Package
ML20138K431	List: ML20236T791 ML20236T793 PY-CEI-NRR-2299, Supplemental Application for Amend to License NPF-58, Revising MSL Leakage Requirements & Eliminating Msli Valve Leakage Control Sys.Proprietary & non-proprietary Rev 1 to Calculation Psat 08401T.03,encl.Proprietary Info Withheld
References
NUDOCS 9807290016
Download: ML20236T791 (30)

Text

,. . . . . .

l Attachment 3 PY-CEI/NRR-2299L Page1of3 PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY hbni[requiredtob e in MbbkS 5 h 2.. I S h h kl%i94frI5oftleth.k b.--

t i Verify the combined leakage rate for all l In accordance secondary containment bypass leakage with the I

paths is s 0.0504 L, when pressurized to Primary l

a P, . Containment Leakage Rate g;[ Testing p Program i

SR 3.6.1.3.10 ------------------NOTE-------------------

! Only required to be met in MODES 1, 2.

and 3.

Verify leakage brough each main In accordance l

steam line is cfh when tested at with the e total leakage rate Primary

@ a P,.Mac _.tnr igh 6 1 Tour main steam lines Containment is s scfh, wkm tu+,.A at 2: P., Leakage Rate 2W Testing Program (continued) t 9807290016 980722 PDR ADOCK 05000440 P pop PERRY - UNIT 1 3.6-18 Amendment No. e ,W ,87

(AcIn 6 W m.S h + O W ly M , ,,

8 8v 4 v LyJ J.6.1.9 i i

Attachment 3 3.6 CONTAINMENT SYST N PY-CE!/NRR-2299L Page 2 or3 l

! s k sA- d'& W *- 0 3.6.1.9 Main Stea Ismc tion valve (tiSI") Lc;k:gc Contrc' System (LCS)

LC0 3.6.1.9 T.ec "cI'/ LCS . subs,u_c + om..~e e .h. _e '. ' _h ^ n..o r n_a "' r.

1 t TL Main Ste N+oM Vets s kutt 6 cPE.M ALE. . .

,, APPLICABILITY:

- - - - - - - DES 1. MO 2. and 3.

- - - - - - - - - N o v e_ -

s f ede Con 6Mron eey is att w,A 4.c ac.h pc.n, Ara +ren 46 v4.

. CTIONS CONDITION REQUIRED ACTION COMPLETION TIME (Ermere Main Stam Sh *M Vdv25)

A. ~ OnbMSI'! LCS subsystem A.1 ==

!.'OTE "

inoperable. -LCO 3.0.'. i; not 1 --

ama'ic;bic for the "-

[ A.1 Close the inoperable Ma bbcardIISI'!LCS "~

Steam Shutoff Valve. '

-Cubsystem '

--~

--_- js t

1 Tkis  %

s n~m ,,,,,,,- +

'5 IM ~, h - ,wu, w u u .> ays

.e i--sj. +wcmw .+m

. nn-n+~ ~

L -status . " ~

wv w LIVWLL y

&s r,.-

~ . u.e.ru.

. cc o w.,

n o_m - + . m_. . e .r u, , ~ -

.o.s ,,mym . d -

uw subg ;tcm: ncper;ble.%_ ~

-cub yrtem to OPEPJ."LE e statur. D ,

& B i

R. Required Action and associated Completion

/1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Time not met. AND l

/.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> l

1 i

l OfD '

f ^

cha,3c, A *ste p Valve " -Fo S h d e@ V d vs- -thr,$ ed k -h. mar k up, s ~ A ~

PERRY - UNIT 1 3.6-27 Amendment No. E,,E-

Nn sw.- Shaon- Wlves t 19 I"'

.U.C T ._.-w, t-

.E.1.

Attachment 3 PY-CEl/NRR-2299L age 3 d3 SURVEILLANCE REQUIREMENTS 1

SURVEILLANCE FREQUENCY 2.n o e , n ,

. .v u. i re, u x,. v~o.- -

,o v.v.4.s.. . vyw

% _,+n. u w w nw, w. - u .u.e, ,s o, .

ww. w. -

v. a , , , .,

--s s eav. 4.m.... .+u m. ,.w .,. . s -ws.

. c,n. oe , no ,1

. v.v...s.u s.fw.n . 4. #. g,mi.n,.+

. ww.. 4.

,.nm&

. m v . . 4..m. . , , 4. +v.

, e r -n -, .r h. o, 31 A,,,,. m __

v. -ws-4 m.uvu.b n-s r. A u. .c, T. t.t t t.'e r, . .h e or,+ nm.. . , n , + n r. n1 n.

.-,s.. . ww.m - . nm.

m i ~ .3.+.a- .-

4. & e s , o 2w4. m ,o-e. .g. -_

i I ~

%6 kr.c of'" A.a.nea w ik I SR 3.6.I. . n

.wi

_.u..,

c .t u m

cv ._....

, rur+

.,.1,.n.m..

w c ,+

5. . 4. vm .m.w,1

+n,+

.w,. ,m, f, n e ,. h. .

i

'3 N mInserv.*ca

.0 e r + L ,.

h+;n

, . r, J.,, ,,,, +.n m. . ,m.w i . . . . . . . . ,

. w . c.e

. . . . . . M %. ,

a i

l WrM h isola h h e eE u<-k vdvc is w;&in h its.

i T M s.

{

h suep\wat { 1 1

l l

l PERRY - UNIT 1 3.6-28 '

Amendment -

l

___a

Attachment 4 PY-CEI/NRR-2299L Page 1 of 9 The following information is a dupIlcate of the information provided in Attachment 1 to the originalsubmittal dated August 27,1996, except that necessary changes as a result of the current Supplement are included. Added or revised text is identified by italic text and revision bars. Deleted text is shown by use of revision bars.

SUMMARY

Removal of the Technical Specification requirements for the Main Steam Isolation Valve j Leakage Control System (MSIV-LCS), and increasing the allowable leak rate specified for the Main Steam lines is the result of applying the revised accident source term (as documented in NUREG-1465 and the Nuclear Energy Institute (NEI) document " Generic Framework for

. Application of Revised Accident Source Term to Operating Plants") to the design basis Loss of Coolant Accident (LOCA) off-site and Control Room dose analysis for the Perry Nuclear Power Plant (PNPP) l The proposed amendment was the subject of a meeting with the NRC staff on May 30,1996.

MSIV-LCS Technical Specification 3.6.19 identifies the operability requirements for the MSIV-LCS.

In 1984, a program was initiated by the NRC to make regulatory requirements more efficient by eliminating those with marginal impact on safety. An industry survey resulted in a list of 45 l candidates for potential regulation modification. The survey results and analyses of the selected candidates were published in NUREG/CR-4330 " Identification of Regulatory Requirements that may have Marginal Importance to Risk", Volumes 1,2, and 3. One of the candidates was to l eliminate the requirement for the MSIV-LCS in Boiling Water Reactors (BWRs) per Regulatory l Guide 1.96," Design of Main Steam Isolation Valve Leakage Control Systems for Boiling Water l Reactor Plants". This submittal is a result of continuing efforts to achieve the goal of eliminating that system.

Removing the Technical Specification requirements for the MSIV-LCS is based on reanalysis of I off-site and Control Room doses, where the MSIV-LCS is not credited in the calculation. As

)

. noted above, the reanalysis utilizes the revised design basis accident (DBA) source terms. The  ;

limiting reanalysis case assumes that main steam line leakage is attenuated in the main steam line from the reactor vessel out to the outboard MSIV. This is the limiting scenario since the worst case single failure, and hence the most limiting analysis case, involves a failure to close j the valve downstream of the outboard MSIV in each main steam line, i.e., the Main Steam  ;

ShutoffValves (INI1F0020A,B,C and D). Although this most limiting analysis case assumed l a failure to close the Main Steam ShutoffValves, retention of OPERABILITY requirements on these valves is appropriate to ensure the single failure analysis remains valid.

l L

Attachment 4 PY-CEl/NRR-2299L Page 2 of 9 Not crediting the MSIV-LCS in the design basis accident analysis is consistent with the approach taken by several BWR licensees, which have applied for NRC approval of this change using an approach developed by the Boiling Water Reactor Owners Group (BWROG). The BWROG methodology involves seismically qualifying the main steam lines out to and including the non-safety related, non-seismic drain lines and main condenser, and then using that volume to attenuate leakage past the MSIVs. At PNPP, the existence of safety related, ,

seismically qualified piping leading to the safety r&ted, seismic, Class 1E powered Main Steam

{

ShutoffValves, together with the characterbtics of the revised accident source term j l

(i.e., predominantly aerosol which is largely retained in the drywell, containment and main

{

steam lines) provides the option of taking credit only for the volume within the main steam lines j

for leakage attenuation.

{

Although the requirements for the MSIV-LCS are being removed (since credit is no longer taken for the system as part of the design basis accident analysis), OPERABILITY requirements on the Main Steam ShutoffValves are being retained since the valves meet }

l '

Criterion 3 of 10CFR50.36 (c)(2)(ii). Specifically, a technical specification limiting condition for operation must be established for each item that is "A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier." The only necessary surveillance requirement is to ensure the valves will stroke closed on a remote-manual demand by the operators. Leak test l requirements are not necessary to ensure the assumptions of the dose calculation methodology are met for the main steam lines, since leakage flow characteristics used in the analyses are affected only by the turbulence caused by an open ended pipe (i.e., the Main Steam Shutoff l Valves fail to close).

Implementation of this license amendment prior to restart from the next refueling outage (refuel outage 7) will fulfill a commitment made to the NRC in a letter dated April 25,1995 l (PY-CEI/NRR-1934L). The commitment is to resolve issues which result in declaring the inboard MSIV-LCS subsystem inoperable during plant startups. Currently, an exception to l Limiting Condition for Operation (LCO) 3.0.4 is incorporated into the Technical Specificat:on for the MSIV-LCS, because until plant power levels are increased above 50 percent, the inboard MSIV-LCS subsystem is considered to be inoperable (further details on this issue are in the April 25,1995 letter and in License Amendments 63, 71 and 89). The existing 3.0.4 exception was issued by Amendment 89 to the Operating License. Elimination of the MSIV-LCS Technical Specification requirements beginning with RFO 7 (as proposed by this  ;

l licensing submittal) will resolve the issues which result in declaring the inboard MSIV-LCS inoperable during operations below 50 percent rated thermal power and, consequently, will l climinate the necessity for the 3.0.4 exception.

I The physical isolation of the MSIV-LCS from the Main Steam system will eliminate leakage  !

pathways. This modification will be performed as part of the PNPP design change process. i 1

l

Attachment 4 PY-CEI/NRR-2299L Page 3 of 9 MAIN STEAM LINE LEAKAGE RATE This amendment request proposes to limit the leakage through each main steam line to less than or equal to 100 scfh, as long as the combined leakage rate through the four main steam lines is less than or equal to 250 scfh. Technical Specification 3.6.1.3 currently requires that the main steam line leakage rates shall be limited to less than or equal to 35 scfh through each main steam line, as long as the totalleakage rate through allfour main steam lines is less than or equal to 100scfh.

The purpose fbr limiting the main steam line leakage rate is to ensure isolation of the reactor  ;

coolant system in the event of a design basis LOCA. Industry operating experience has shown '

that these valves invariably exhibit some level of minor leakage. The current Technical Specification allowable leakage rate is extremely small considering the physical size and operating characteristics of the MSIVs (i.e., large size and fast acting). Based on an in-depth evaluation of MSIV leakage (refer to NEDC-31858P, "BWROG Report for Increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems", Revision 2, and the summary in NUREG-1169 " Technical Findings Related to Generic Issue C-8; Boiling Water Reactor Main Steam Isolation Valve and Leakage Treatment Methods"), the BWROG has concluded that leakage rates of up to 500 scfh are not indicative of substantial mechanical I defects in the valves which would challenge the valves capability to fulfill their safety function ofisolating the steam lines. Therefore, as demonstrated in the design basis LOCA radiological reanalysis, the proposed increased allowable main steam line leakage rate (each line less than or equal to 100 scfh, and total leakage less than or equal to 250 scfh, when tested at Pa) will not affect each MSIV's isolation function capability. Therefore, the overall level of plant safety will .

be maintained, while reductions can be achieved in: radiation dose to maintenance workers; the {

maintenance workload during plant outages; and the potential for outage extensions.

Per agreements reached with the NRC through the BWROG program for increasing main steam line leakage rate limits, a commitment is made that if the leakage rate on a main steam line exceeds 100 scfh during Technical Specification required testing, the leakage rate for that line will be restored to 5 25 scfh when tested at Pa (the originallicensed value for leakage), prior to l plant restart.

DESCRIPTION OF TIIE PROPOSED TECIINICAL SPECIFICATION CIIANGE Technical Specification Surveillance Requirement (SR) 3.6.1.3.10 will be revised to limit the  !

leakage through each main steam line to less than or equal to 100 scfh when tested at Pa, as long as the combined leakage rate through the four main steam lines is less than or equal to 250 scfh when tested at Pa. In addition, the Specification pertaining to the MSIV-LCS will be deleted. In

! its place, a new Specification entitled " Main Steam ShutogValves" will be created, which l

retains OPERABILITY requirements on the Main Steam ShutoffValves to ensure that the Main Steam ShutogValves remain OPERABLE during MODES 1,2 and 3. Refer to Attachment 3 for a marked-up copy of the affected Technical Specification pages.

Attachment 4 PY-CEl/NRR-2299L Page 4 of 9 For additional Technical Specification changes proposed by this Supplement, see Attachment 1.

l

[ Note: The Bases will be revised under the PNPP Bases Control Program to reflect the changes identified above. In addition, the Bases will be revised to provide clarifying text to the discussions on the Containment Spray (B 3.6.1.7), and Combustible Gas Mixing (B 3.6.3.3)

Systems, in order to reflect their contribution to the post-LOCA removal of airbome radionuclides. The proposed Bases and Table of Contents mark-ups are contained in Attachment 5 "for infomiation only".]

l This proposed change has been developed for implementation during refuel outage 7.

l SAFETY ANALYSIS Removal of the Technical Specification requirements for the MSIV-LCS, and increasing the allowable leak rate specified for the Main Steam lines is the result of applying the revised accident source term (as documented in NUREG-1465 and the NEI Generic Framework document (Reference 1)) to the design basis accident (DBA) LOCA off-site and Control Room I dose analysis for the Perry Nuclear Power Plant (PNPP).

As noted in the NEI Generic Framework Document and in the PNPP LOCA reanalysis (see j Reference 9), the revised DBA source terms of NUREG-1465 are comparable in conservatism l {

to the DBA source terms previously used at PNPP. The noble gas and iodine release fractions (which are the main determinants of the whole body and thyroid dose evaluations specified in 10CFR100) are about the same. The revised accident source term timing and chemical form, while different from the previous source terms, are nonetheless conservative compared to what is expected under actual accident conditions, (e.g., the 1979 accident at Three Mile Island) and provide a more physically correct representation of activity release to the containment.

Knowledge of the more physically correct representation of the timing and chemical form provides the opportunity to develop the most appropriate mechanisms for mitigating radiological releases.

Furthermore, in terms of activity transport within and through the containment system and )

release to the environment, there are many other conservatism included in the LOCA I reanalysis. The following provides a brief summary of some of the activity transport conservatism that exist within the dose reanalysis. These are explained more fully in  !

Reference 9. l l 1. Earlier gap release start time than required for a BWR

2. Underestimated volumetric flow from the drywell during core damage / debris quench i

3. Neglected suppression pool scrubbing l

4. Neglected natural aerosol removal in unsprayed regions of containment

5. Underestimated containment spray effectiveness

6. Most conservative break location

7. Most conservative distribution of total main steam line leakage

)

Attachment 4 J PY-CEI/NRR-2299L Page 5 of 9

8. Most conservative main steam line valve single failure

9. Underestimated natural aerosol removal in main steam line

10. Underestimated drywell aerosol natural removal

11. Instantaneous iodine release in main steam lines These are significant conservatism, which result in calculated offsite and Control Room doses higher than those that would exist following a release of a NUREG-1465 source term from the reactor core. These activity transport conservatism in the LOCA re analysis are judged to be comparable to the conservatism utilized in the original analyses.

Although (as stated above) the revised source term is considered "more physically correct" than the previous source tenns used in performing design basis LOCA radiological analyses, it should be noted that in an actual design basis LOCA, no fuel damage would occur. The plant operating limits, such as the Maximum Average Planar Linear Heat Generation Rate l

(MAPLHGR), are chosen to ensure that post-LOCA fuel cladding temperatures remain low enough to maintain fuel pin integrity. As discussed in USAR section 15.6.5.5.2 (the Realistic Analysis), "the only activity released to the drywell is that activity contained in the reactor coolant plus any additional activity which may be released as a consequence of reactor scram and vessel depressurization." Therefore, off-site and Control Room doses would be minimal (significantly less than the calculated values presented below, which utilize the NUREG-1465 source term and other inherent conservatism such as the activity transport conservatism described above). The results of the DBA LOCA off-site and Control Room dose reanalysis are provided below.

DOSE RESULTS (REM)

Proposed Existing USAR USAR Regulatory Dose

• Dose Limit # j Control Room Whole Body 0.1 0.4 5  ;

Thyroid 16.2 29.2 30 i Skin 4.8 12.5 30 EAB Whole Body 1.9 3.6 25 Thyroid 157.9 140.8 300 LPZ Whole Body 1.7 1.9 25 Thyroid 130.3 144.7 300

• rounded to nearest tenth

1. Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) dose limits are per 10CFR100.11. Control Room dose limits are per 10CFR50 Append:x A, General Design Criterion (GDC) 19 and NUREG 0800 Standard Review Plan (SRP) Section 6.4

attachment 4 )

PY-CEI/NRR-2299L Page 6 of 9

{

As noted in the NEl Generic Framework Document (Reference 1), the acceptability of applications utilizing the revised accident source terms "may be judged by the same licensing acceptance limits (e.g., dose limits in 10CFR100) in use with the TID-14844 source term.

That is, the licensee would show that the revised design basis, with either selective or essentially complete application of NUREG-1465 together with the plant changes under evaluation, results in doses no greater than these licensing acceptance limits." The off-site dose licensing acceptance limit for PNPP is 10CFR100.11 (see Question 3 of the Significant Hazards Consideration for details on the source of this PNPP licensing acceptance limit). As seen in the above Table, the newly calculated radiological doses are lower than the current analysis for six of the seven factors evaluated. For the one factor which was higher, i.e., at the -

EAB for thyroid dose (from 140.8 REM to 157.9 REM), the dose remained significantly below the 10CFR100 limit of 300 REM to the thyroid. Consequently, the results of the LOCA reanalysis constitute a basis for demonstrating compliance with the requirements of 10CFR100 and with 10CFR50, Appendix A, GDC 19.

A significant amount ofinformation regarding the assumptions, conservatism, and methodology of the dose calculations is provided in the non-proprietary dose calculations (Attachment 6 ofthe original August 27,1996 submittal). The majority af that information is not repeated in this attachment. However, several points are emphasized in relation to the mitigation techniques employed in response to the postulated design basis Loss of Coolant Accident. As noted above, knowledge of the more physically correct source term timing and chemical form permits use of more appropriate mitigation techniques. Specifically, natural forces such as gravitational settling of aerosol (particulate) have been credited inside the drywell and in portions of the main steam lines, which significantly reduces the amount of radionuclides that could escape from the containment and into the environment. Also, based on ,

I a high radiation signal in the Control Room, the Containment Spray system would be operated post-LOCA for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (previous analyses assumed 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of spray operation), in order to scrub released radionuclides from the containment atmosphere and into the suppression pool, and thus reduce the post-LOCA off-site and Control Room dose. Once the containment sprays have been successful in sweeping the iodine to the suppression pool, the iodine must be retained in the water. To achieve this, the pH level of the suppression pool will now be raised to 7 or above following the accident, and then maintained at 7 or above. This prevents significant fractions of the dissolved iodine from being converted to elemental iodine and then re-evolving l to the containment atmosphere. During the course of the accident the pH of the suppression pool can decrease due to radiolysis of reactor coolant and chloride-bearing electrical insulation, which would create acids. The method for pH control will use the existing Standby Liquid Control (SLC) system for raising (and maintaining ) long term post-accident suppression pool pH levels to 7 or above. Calculations have shown that the contents of one tank of the Standby Liquid Control solution will be effective in raising and maintaining pH levels for 30 days following the DBA. Controls over SLC system operability are already included in the Technical Specifications. In addition, a backup method for pH control will be developed for use if post-accident suppression pool sampling identifies that the primary pH control method is not being effective.

Attachment 4 PY-CE!/NRR-2299L Page 7 of 9 Necessary procedure changes and appropriate training will be made in conjunction with implementation of this change to reflect these revised mitigation techniques.

Post-accident operator actions are minimized. The operator action associated with initiating the Containment Spray system does not change. Containment Spray is initiated via a push button in the Control Room. The previously required manual initiation of the MSIV-LCS involved multiple operator actions to open and close numerous valves and start the blowers, which will no longer be required. Replacing these actions, the new analysis assumes the operator closes the Main Steam ShutoffValves (which was previously one of the steps in manually initiating the l MSIV-LCS ), and, based on post-accident pH samples of the suppression pool, initiates the Standby Liquid Control system, which is accomplished via two key lock switches in the Control l

Room. These operator actions are less complex than those previously required, and minimize the probability of an error.

The Control Room dose calculations were performed utilizing the PNPP-specific atmospheric dispersion factors (Chi /Q's) currently listed for the Control Room in the Updated Safety Analysis Report (USAR). These Chi /Q values were determined based on a special study performed at PNPP (Reference 10), which was included for NRC stafTreview as Attachment 5 to the original submittal dated August 27,1996.

The new Main Steam ShutofValve specification 3.6.1.9 (see Attaclunent 3 for the marked-up l pages) retains appropriate controls over the operability of these valves. Previously, operability of the Main Steam ShutofValves was an integral part of the MSIV-LCS system functional test l (SR 3.6.1.9.3), since closure of the valves was necessary for the outboard MSIV-LCS to function. The new LCO requires the valves to be OPERABLE. The ACTIONS taken if a Main Steam ShutofValve is inoperable require the affected main steam line to be isolated in a fashion that would create apost-accident holdup volume for the leakage past the MSIVs. This must be perfomied within 30 days, or the plant is required to be shutdown. The 30 day Completion Time is consistent with the Completion Time that was provided for the MSIV-LCS. The new leakage control method (using the MSSVs) therefore uses that same duration. During this 30 day Completion Time, the remaining OPERABLE MSIVs in that main steam line are adequate to perform the required leakage holdup function shoulda LOCA occur. However, the overall l

reliability is reduced because a single failure of an MSIV in that line could result in a loss of the MSlV leakage holdup function, becausepost-accident, a single MSIVfailure wouldprevent 1 establishment ofa " holdup volume". The purpose of closing the A/SSVin a main steam line is I based on the characteristics of the revised design basis accident source term (i.e., predominantly aerosol). Closing the AISSVwill provide a single-failure-proofholdup volume within the main l steam line after an accident, for deposition of the aerosol on the inner walls of the main steam line. The Required Action does notpermit closure ofan AfSSVand an AISIVat the same time, or both MSIVs at the same time, since doing so duringplant operation could result in differential cooldown ofthat particular Afain Steam Line, with restdtant damage to some small-bore drain 1 piping that is interconnected to the ether Afain Steam Lines. If an MSSV is " inoperable", but closed, credit can be taken for it in meeting the ACTION. Leak tightness of the MSSVs is not necessary to ensure the assumptions of the dose calculation methodology are met for the main i

Attachment 4 PY-CEI/NRR-2299L Page 8 of 9 steam lines, since leakage flow characteristics used in the analyses are affected only by the turbulence caused by an open ended pipe (i.e., the Main Steam ShutoffValves fail to close). l The 30 day Completion Time is based on the redundant capability afforded by the remaining OPERABI.E MSIVs on that line, and the low probability of a DBA LOCA occurring during this period. After 30 days, when the MSSVon that line has been closed, a holdup volume can be established and the plant can continue operation. The ACTIONS are modified by a Note allowing separate condition entry for each penetration flow path because an inoperable MSSV l in a main steam line does not affect the ability to provide a post-accident holdup volume in l

the affected line or in the other lines (between the MSIVs or between an MSIVand its l MSSV). The Required Actions provide appropriate compensatory actions for each inoperable l MSSV. Complying with the Required Actions may allow for continued operation, and l subsequent inoperable MSSVs are governed by subsequent Condition entry and application of associated Required Actions.

i REFERENCES

1. NEl document (prepared by the Electric Power Research Institute) entitled " Generic Framework for Application of Revised Accident Source Term to Operating Plants",

l EPRI TR-105909, Interim Report, November 1995.

2. PSAT 04212H.03, " Ultimate Iodine Decontamination Factor for Perry DBA"

3. PSAT 04202H.04, " Aerosol Decay Rates (Lambda) in Drywell"

4. PSAT 04202H.05, " Aerosol Decay Rates (Lambdas) in Containment with Spray" l 5. PSAT 04212H.06, " Mixing Between the Sprayed and Unsprayed Portions of the Perry i

Containment"

6. PSAT 04202H.08, "Steamline: Particulate Decontamination Calculation"

7. PSAT 04202H.09, " Steam Line: Elemental Iodine Decontamination Calculation" l 8. PSAT 04202H.12, " Calculation of Fraction of Containment Aerosol Deposited in Water"

9. PSAT 04202H.13, "Off-site and Control Room Dose Calculation"

10. NUS-4792, "Results of the Atmospheric Tracer Study Within the Building Complex at the PNPP" COMMITMENTS WITHIN THIS LETTER l

Identified below are the actions committed to in this letter.

• Per agreements reached with the NRC through the BWROG program for increasing main steam line leakage rate limits, a commitment is made that if the leakage rate on a main steam line exceeds 100 scfh during Technical Specification required testing, the leakage rate e

for that line will be restored to s 25 scfh when tested at Pa (the originallicensed value for l  ;

leakage), prior to plant restart.

= Based on a high radiation signal in the Control Room, the Containment Spray system would be operated post-LOCA for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (previous analyses assumed 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of spray I

1

_ _ _ _ _ _ _ _ _ _ _ _ a

Attachment 4 PY-CEI/NRR-2299L Page 9 of 9 operation), in order to scrub released radionuclides from the containment atmosphere and into the suppression pool, and thus reduce the post-LOCA off site and Control Room dose.

• The pH level of the suppression pool will be raised to 7 or above post-LOCA, and then maintained at 7 or above. The method for pH control will use the existing Standby Liquid Control (SLC) system.

• A backup method for pH control will be developed for use if post-accident suppression pool sampling identifies that the primary pH control method is not being effective.

. Necessary procedure changes and appropriate training will be made in conjunction with implementation of this change to reflect these revised mitigation techniques.

ENVIRONMENTAL CONSIDERATION The proposed Technical Specification change request was evaluated against the criteria of 10CFR5:.22 for environmental considerations. The proposed change does not significantly increase individual or cumulative occupational radiation exposures, does not significantly change the types or significantly increase the amounts of effluents that may be released off-site and, as discussed in Attachment 2, does not involve a significant hazards consideration. Based on the foregoing, it has been concluded that the proposed Technical Specification change meets the criteria given in 10CFR51.22 (c)(9) for categorical exclusion from the requirement for an Environmental Impact Statement.

l 4

l  !

L t

__.____._____.___________a

Attachment 5 PY-CE!/NRR-2299L TABLE OF CONTENTS Page1of18 8 3.5 CORE ISOLATION COOLING (RCIC) SYSTEMEMERGENCY 6 " "l}ICO B 3.5.1 ECCS -0)erating ......

B 3.5.2 ECCS - S iutdown . . . . . . .

. . . . . . . . . . . . B 3.5-1 8 3.5.3 RCIC System ............. B 3.5-15 B 3.5-21 3.6 CONTAINMENT SYSTEMS 3.6.1.1 Primary Containment-Operating 3.6.1.2 Primary Containment Air Locks ............. 3.6-1 3.6.1.3 Primary Containment Isolation Valves ........... 3.6-3 3.6.1.4 Primary Containment Pressure . . . . (PCIVs) . . . .... . 3.6-9 3.6.1.5 Primary Containment Air Tem .......... 3.6-20 3.6.1.6 Low-Low Set (LLS) Valves . perature .......... 3.6-21 3.6.1.7 ............... 3.6-E2 Residual Heat Removal (RHR) Containment Spray Syst 3.6.1.8 Feedwater a' age rol System (FWLCS) . 3.6-24 3.6.1.9 Main Steam 20' ati er L'"^ '"""' ' ^"^ ". . ^1. ."~.--

. . 3.6-26

.i - n.

3.6.1.10 Primary Containment-Shutdown . . . . . . . . . .

. . . 3.6-27 3.6.1.11 Containment Vacuum Breakers

... 3.6-29 3.6.1.12 .........

Containment Humidity Control . . . . . . . . . . . . . 3.6-31 3.6.2.1 Suppression Pool Average Temperature . . . . .

.... 3.6-34 3.6.2.2 Suppression Pool Water Level . . . . . . . . . . . . . 3.6-36 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool

.... 3.6-39 Cooling 3.6.2.4 ......................

Suppression Pool Makeup (SPMU) System .......

3.6-40 3.6.3.1 . 3.6-42 Primary Containment Hydrogen Recombiners . . . . . . 3.6-44 3.6.3.2 Primary Containment and Drywell Hydrogen 3.6.3.3 Igniters . . . . . . . . . . . . . . . . . . . . . .

Combustible Gas Mixing System 3.6-46 3.6.4.1 Secondary Containment ...... ............. 3.6-49 3.6.4.2 ........... 3.6-51 Secondary Containment Isolation Valves (SCIVs) . . . . . 3.6-53

, 3.6.4.3 Annulus Exhaust Gas Treatment (AEGT) System 3.6.5.1 Drywell ............. ...... 3.6-56 3.6.5.2 Drywell Ai r Lock . . . . . . . ............ ............ 3.6-59 3.6.5.3 Orywell Isolation Valves . 3.6-61 3.6.5.4 Drywell Pressure . . . . . . . . . . . . . . . . . . . 3.6-65 I 3.6.5.5 Drywell Air Temperature .............. 3.6-69 3.6.5.6 . .... .......... 3.6-70 Drywell Vacuum Relief System . . . . . . . . . . . . . 3.6-71 B 3.6 CONTAINMENT SYSTEMS B 3.6.1.1 Primary Containment-Operating .............

B 3.6.1.2 Primary Containment Air Locks ............. B 3.6-1 B 3.6.1.3 Primary Containment Isolation Valves (PCIVs B 3.6-7 B 3.6.1.4 Primary Containment Pressure . . . . . . . ). ...... B 3.6-17 B 3.6.1.5 Primary Containment Air Tem

...... B 3.6-33 B 3.6.1.6 Low-Low Set (LLS) Valves . perature

........ B 3.6-36

.... ........... B 3.6-39 l

l (continued)

PERRY - UNIT 1 iv I Revision No. 0 '

Attachment 5 PY-CEI/NRR.2299L TABLE OF CONTENTS B S.6 B 3.6.1.7 CONTAINMENT SYSTEMS (continued)

Residual Heat Removal (RHR) Containment

.gg[*y{jN190"T " * " "

Spray Sy eom -. ...................

8 3.6.1.8 Feedwater eaka 01 System (FWLCS) ........ B 3.6-43 B 3.6-48 8 3.6.1.9 Main Stea "^' " ^" '"^'"c"" ' ^& ^ ^ "^*c V B 3.6.1.10 Syst a 'LCS) . . . .

Primary Containment-Shutd 6.Q .M. l

.... B 3.6-51 B 3.6-55 B 3.6.1.11 Containment Vacuum Breakers ............ . B 3.6-59 B 3.6.1.12 Containment Humidity Control . . . . . . . . . . . . . B 3.6-65 B 3.6.2.1 Suppression Pool Average Temperature . . . . . . . . . . B 3.6-70 B 3.6.2.2 Suppression Pool Water Level . . . . . . . . . . . . . . B 3.6-75 8 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling ...................... B 3.6-79 B 3.6.2.4 Suppression Pool Makeu (SPMU) System ......... B 3.6-83 B 3.6.3.1 Primary Containment Hy rogen Recombiners . . . . . . . . 8 3.6-90 B 3.6.3.2 Primary Containment and Drpell Hydrogen Igniters . . . . . . . . . . . . . . ....... B 3.6-95 B 3.6.3.3 Combustible Gas Mixing System ............. B 3.6-101 B 3.6.4.1 Secondary Containment .

................ B 3.6-106 B 3.6.4.2 Secondary Containment Isolation Valves (SCIVs) . . . . . B 3.6-111 B 3.6.4.3 Annulus Exhaust Gas Treatment (AEGT) System . . . . . . B 3.6-118 B 3.6.5.1 Drywell ...................... B 3.6-123 B 3.6.5.2 Drp ell Air Lock . . . . . . . . . . . . . . . . . . . . B 3.6-128 Drpell Isolation Valves . . . . . . . . . . . . . . . .

8 3.6.5.3 B 3.6-136 B 3.6.5.4 Drywell Pressure . . . . . . . . . . . . . . . . . . . . . B 3.6-145 B 3.6.5.5 Drpell Air Tem)erature ................ B 3.6-148 B 3.6.5.6 Drpell Vacuum 1elief System . . . . . . . . . . . . . . B 3.6-151 3.7 PLANT SYSTEMS 3.7.1 Emergency Service Water (ESW) System-Divisions 1 and 2 ................. 3.7-1 3.7.2 Emergency Service Water (ESW) System-Division 3 .... 3.7-3 3.7.3 Control Room Emergency Recirculation (CRER) System . . . 3.7-4 3.7.4 Control Room Heating. Ventilation, and Air Conditioning (HVAC) System . . . . . . . . . . . . . 3.7-8 3.7.5 Main Condenser Offgas ................. 3.7-11 3.7.6 Main Turbine Bypass System . . . . . . . . . . . . . . . 3.7-13 3.7.7 Fuel Pool Water Level ................. 3.7-14 3.7.8 Fuel Handling Building . . . . . . . . . . . . . . . . . 3.7-15 3.7.9 Fuel Handling Building Ventilation Exhaust System ... 3.7-16 3.7.10 Emergency Closed Cooling Water (ECCW) System . . . . . . 3.7-19 8 3.7 PLANT SYSTEMS B 3.7.1 Emergency Service Water (ESW) System-Divisions 1 and 2 . ................ B 3.7-1 B 3.7.2 Emergency Service Water (ESW) System-Division 3 ... B 3.7-7 B 3.7.3 Control Room Emergency Recirculation (CRER) System . . . B 3.7-10 (continued)

PERRY - UNIT 1 v Revision No. O

Primary Containment-0perating 8 3.6.1.1 B 3.6 CONTAINMENT SYSTEMS ^""d"""

PY-CEI/NRR-2299L B 3.6.1.1 Primary Containment-Operating Page 3 or 18 BASES . bhNi)!b b h[h[

BACKGROUND The function of the primary containment is to isolate and contain fission products released from the Reactor Coolant System following a design basis Loss of Coolant Accident (LOCA) and to confine the postulated release of radioactive material to within limits. The primary containment consists of a free standing steel cylinder with an ellipsoidal dome, secured to a steel lined reinforced concrete mat. which surrounds the Reactor Coolant System and provides an essentially leak tight barrier against an uncontrolled release of radioactive material to the environment.

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

The isolation devices for the penetrations in the primary containment boundary are a part of the primary containment leak tight barrier. To maintain this leak tight barrier:

a. All primary containment penetrations required to be closed during accident conditions are either:

, 1. capable of being closed by an OPERABLE primary l containment automatic isolation system, or

2. closed by manual valves, blind flanges, or i

de-activated automatic valves secured in their i closed positions, except as provided in LC0 3.6.1.3. " Primary Containmerit Isolation Valves (PCIVs)";

l b. Primary containment air locks are OPERABLE except as

]rovided in LC0 3.6.1.2. " Primary Containment Air

_ocks":

c The equipment hatch is closed and sealed:

d. The leakage control systems associated with penetrations are OPERABLE, except as provided in" LCO 3.6.1.8. "Feedwater leakage Control Syste sCO J.' .1.5. "Ha m Stcom Iscution Wlvo (MSIV)

Leaka;c Centrol Syste: 'LCS)fi (contir.ned)

PERRY - UNIT 1 B 3.6-1 Revision No. 1

Primary Containment-0perating B 3.6.1.1

..1 gn c. 3 cp c, t Attachment 5 BASES

.k/p g$ bN1 bdd.vy;jpj( PY-CEl/NRR-2299L Page 4 of18 BACKGROUND e. The containment leakage rates are in compliance with (continued) the requirements of Specification 3.6.1.1 and Specification 3.6.1.3: .

f. The suppression pool is OPERABLE: and

g. The sealing mechanism associated with each primary containment penetration, e.g., welds, bellows. or 0-rings, is functional.

This Specification ensures that the performance of the primary containment, in the event of a DBA. meets the assumptions used in the safety analyses of References 1 and 2. SR 3.6.1.1.1 leakage rate requirements are in conformance with 10 CFR 50. Appendix J. Option B (Ref. 3),

as modified by approved exemptions.

APPLICABLE The safety design basis for the primary containment is that SAFETY ANALYSES it must withstand the pressures and temperatures of the limiting DBA without exceeding the design leakage rate.

l The DBA that postulates the maximum release of radioactive material within primary containment is a LOCA. In the analysis of this accident, it is assumed that primary containment is OPERABLE such that release of fission products to the environment is controlled by the rate of primary containment leakage.

Analytical methods ar.d assumptions involving the primary containment are presented in References 1 and 2. The safety analyses assume a(disikechanistic' fission product release following a DBA.fwTftCh forms the basis for determination of l

L

[bwADottsite coses. The fission product release is, in turn, I based on an assumed leakage rate from the primary I containment. OPERABILITY of the primary containment ensures I that the leakage rate assumed in the safety analyses is not exceeded.

l The maximum allowable leakage rate for the primary l

' containment (L drywell air pe.)24r hours at the design basis LOCA maximumis 0.20 peak containment pressure (P,) of 7.80 psig (Ref. 4).

Primary containment satisfies Criterion 3 of the NRC Policy Statement.

l l

PERRf - UNIT 1 B 3.6-2 Revision No.1 l

4 PCIVs I B 3.6.1.3 8 dOd2af/Iri

[ M $ G W i 7)U;p; g,7 og g; ptj g 3 ty Attachment 5 BASES PY-CEI/NRR-2299L Page 5 on 8 SURVEILLANCE SR 3.6.1.3.9 (continued)

RE0llIREMENTS device. If both isolation devices in the penetration are j A sw^ A N o+e- closed, the actual leakage rate is the lesser leakage rate '97s of the two devices.

45 i +

OQ l  %+ em.m 5% A Note is added to this SR which states that these valves

are only re1, 2 and 3. quired to meet this leakage rate limit in MODES In the other conditions, the Reactor Coolant U

• l'* Y

! i

' System is not pressurized and specific primary leakage rate n d nob 6 l

limits are not required. The Frequency is required by the 97- '

gg % g rimary Containment Leakage Rate Testing Program. Gy

' e,n.t.cy' g,, wt SR 3.6.1.3.10 b f"5 7 The analyses than theinspecified References 1 and rate. 2 are based on leakage th l  % +a.\, 3.nc { is less leakage Leakage through

• 4 "^ i each main steam line must be s 35 scfh when tested at a P . D-g

^*' and the total leakage rate through all four main steam lifies issMjscfh The Frequency is required by the Primary N w k ;s .,

2h-Conta'inment Leakage Rate Testing Program. 'O

$ Abs. sea CR 5'f g.3 ;, g 'The outboard MSIVs must have a safety related aie source available for use following an accident in order for leakage 4 rnAidj u \ Aose, to be within limits. Therefore, anytime that this air source from the "B" train of P57 Safety Related Air System 053 d d aMons)Is is not available, the outboard MSIVs may not be able to meet j ,g g this surveillance requirement. .

l g u,gd) A only Notere is added to this SR which states that these valve are 4 tawA.

• A land 3. quired to meet this leakage rate limit in MODES 1. 2.

In other conditions, the Reactor Coolant System is 1

%g g;g not pressurized and specific primary containment leakage rate limits are not required.

hpus LLu;g-

{

isi d h58- SR 3.6.1. M 1 N i' Y Surveillance of hydrostatically tested lines provides wwreA 6 ' assurance that the calculation assumptions of References 1 97-S n. u ,l.3 ,10. nd 2 are met. The combined leakage rate must be 09f k

(continutd)

If h le4 9 rade o^ anj N in Stum Li ne, ,

e-you M t oo s.c4 h, h l4.4.a. p rn.te . uM k-(r(stord +c w i w's 25 sc.f L wh 4atQ 5t 2 Pa.

PERRY - UNIT 1 B 3.6-31 Revision No. 1 '

RHR Containment Spray System B 3.6.1.7 Attachment S B 3.6 CONTAINMENT SYSTEMS PY-CEl/NRR-2299L B 3.6.1.7 Residual Heat Removal (RHR) Containment Spray System BASES

-e M BACKGROUND The 3rimary containment is designed with a suppression pool so tlat. in the event of a loss of coolant accident (LOCA),

steam released from the primary system is channeled through the suppression pool water and condensed without producing significant pressurization of the primary containment. The primary containment is designed so that with the pool initially at the minimum water volume and the worst single failure of the primary containment heat removal systems.

suppression pool energy absorption combined with subsequent operator controlled pool cooling will prevent the primary containment pressure from exceeding its design value.

However, the primary containment must also withstand a postulated bypass leakage pathway that allows the passage of steam from the drywell directly into the airspace bypassing the suppression pool. primaryThe primarycontainment containment also must withstand a low energy steam release into the primary containment airspace. The RHR Containment.

• Spray System is designed to mitigate the effects of bypass leakage and low energy line breaks.

INSEET Aa-o-There are two redundant 100% capacity RHR containment spray subsystems. Each subsystem consists of a suction line from the suppression pool, an RHR pump. two heat exchangers in series, and three spray saargers inside the primary containment (outside of tie drywell). Dispersion of the spray water is accomplished by 346 nozzles in subsystem A and 344 nozzles in subsystem B.

~

F The RHR containment spray mode will be automatically initiated _if required, followina a LOCA. gcattina t%9T SfRAy .

AS M AH u4LL4 IHrn AT6D FM C6NTnirmerr AmosPMRF l (foe-McA po6e. Mrr\6/rricO_tF REQUIRcR -

APPLICABLE Reference 1 cohti1 tit 1t1 F results of analyses that predict SAFETY ANALYSES the primary containment pressure response for a LOCA with the maximum allowable bypass leakage area.

The equivalent flow path2 area for bypass leakage has been specified to be 1.68 ft . The analysis demonstrates that (continued)

Fo R coraTA W t4 EtJT PE550RE Reduction GAse.p M PRE 56URE. I NSTRUME.WT ATlo d )

PERRY - UNIT 1 B 3.6-43 m. .~ Revision No. 1

{')) [kh b

u !J i , I, f Attachment 5

- ML Page 7 jg JNSERT "A" The RHR containment spray mode is operated post-LOCA, for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, in order to scrub released radionuclides from the containment atmosphere and into the suppression pool, and thus reduce the post-LOCA off site and Control Room dose. Post-LOCA manual initiation for this function is based on a high radiation signal in the Control Room, .

l l

hhkikhjlNbb -

4 l

RHR Containment Spray System B 3.6.1.7

'}T 6

ATc,JL1by n a k IV }2f8 OL Attachment 5 PY-CEUNRR-2299L BASES Page 8 of I8 APPLICABLE with containment spray operation the primary containment SAFETY ANALYSES pressure remains within design limits.

(continued) ~

The RHR Containment Spray System satisfies Criterion 3 of the NRC Policy Statement.

LCO In the event of a Design Basis Accident (DBA), a minimum of one RHR containment spray subsystem is required to mitigate (drywclj potentiallbypass leakage paths and maintain the primary containment peak pressure below design limits F To ensure that these requirements are met, two RHR containment spray subsystems must be OPERABLE. Therefore, in the event of an accident, at least one subsystem is OPERABLE assuming the worst case single active failure. An RHR containment spray subsystem is OPERABLE when the RHR pump, two heat exchangers in series. and associated piping, valves, instrumentation, and contrnk are OPERAB h (LAD

• PMir4 fop- conApJmrM MW4HQY poSF_ RaDuC004 APPLICABILITY J

In MODES 1. 2. and 3. a DBA could cause pressurization of primary containment. In MODES 4 and 5. the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MONS. Therefore, maintaining RHR containment spray subsystems OPERABLE is not required in MODE 4 or 5.

- . ACTIONS A.1 ,

With one RHR containment s) ray subsystem inoperable the ino)erable subsystem must se restored to OPERABLE status witlin 7 days. In this Condition. the remaining OPERABLE RHR containment spray subsystem is ade y primary containment cooling function. However, yate to perform the the overall l-reliability is reduced because a single failure in the OPERABLE subsystem could result in reduced primary containment cooling capability. The 7 day Completion Time

! was chosen in light of the redundant RHR containment spray capabilities afforded by the OPERABLE subsystem and the low L probability of a DBA occurring during this period. 1 (continued)

R@HK@l @lD1 PERRY - UNIT 1 B 3.6-44 Revi.sion No. 1

m ham SLde Whes Nhh3bbO[$ b$b B3 .1 Attachment 5 B 3.6 CONTAINMENT SYSTE PY-CEUNRR-2299L 5 Abte s1ai c e..s Page 9 ofI8 8 3.6.1.9 Main Stea g:c. men V ,vc (MS"D Le&g: Centrol Sy;te ; (Lc;) -

BASES -

BACKGROUND e MSIV LCS su)plements the isolation function of the MSI by rocessing tne fission products that could leak throu '

the osed MSIVs after a Design Basis Accident (DBA) 1 s of coolan accident (LOCA).

The MSIV L consists of two independent subsyste : an inboard subs tem, which is connected between t inboard and outboard Vs: and an outboard subsystem which is connected immedi tely downstream of the out ard MSIVs.

Each subsystem is apable of processing 1 age from MSIVs following a DBA LOC Each subsystem co ists of blowers (one blower for the i card subs id two blowers for the outboard subsystem). valves.ystem I w.r4 B 9 the inboard subsystem on1 The pip' g. and heaters (for ur electric heaters in the inboard subsystem are p vid a to boil off any condensate prior to the proce flow passing through the blower.

Each subsystem operates i two proc s modes:

depressurization and bl doff. The 3 pressurization process -

reduces the steam lin pressure to wit the operating capability of equip nt used for the ble off mode. During bleedoff (long te , leakage control), the )

negative ]ressur in the main steam lines (Rowers maintain a

. 1). This i

ensures tlat akage through the closed MSIVs collected 4 by the MSIV S. In both process modes, the pro ss flow is discharge o the shield building annulus, which e loses a volume s ved by the Annulus Exhaust Gas Treatment tem (AEGTS .

Th MSIV LCS is manually initiated approximately 20 minut 11owing a DBA LOCA (Ref.1).

APPLICABLE 'c LCS mitigates the consequences of a DBA LW 7 ensuring ti 1

SAFETY ANALYSES ion products that may le the closed 1 MSIVs are diverted o .e ield b' ' g annulus and ultimately filtered by AEr ..

1 ses in Reference 2 provide the s alua ,o offsite dose co hurt C- operatio i e MSIV LCS prevents a release of un ces. The e for this type of event.

(continued)

PERRY - UNIT 1 B 3.6-51 Revision No.1

Attachment 3 PY-CEI/NRR-2299L i Page 10 of 18 l

l

\

, . : . e n. -. n INSERT B for " Background": ,buO,v$ $ $W D 3 The post accident function of the Main Steam Shutoff Valves (MSSVs) is to be remote-manually l

closed in order to provide a reduction of post accident dose associated with the main steam line leakage path. With the Shutoffvalves in a closed position, mitigation of the off-site and Control l

Room dose is achieved by taking credit for the deposition of particulate forms of released fission products (aerosols) on the inner walls of the four main steam lines. This removal process is based on a " plug flow" model for the Main Steam Isolation Valve (MSIV) leakage with relatively even, slow cooldown of the insulated main steam lines. The closed position of the MSSV supports the plug flow model by providing isolation of the spacejust upstream of the MSSV from external convection which could originate from the downstream nonsafety side of the MSSV. Therefore, the MSSVs are required to move to a closed position, but are not required or credited with any tightness against leakage.

The Afain Steam Shutoff Valves (IN11 F0020 A, B, C, and D) are safety-class, remote manual motor-operated valves. The operator response to provide the manually initiated closure for the MSSVs is 20 minutes post-LOCA (Ref.1). Failure of all four motor-operated MSSVs to close is taken as a single active failure based on a single operator error or on a loss of divisional power.

This failure is not coincident with a single MSIV failure to close.

INSERT C for " Applicable Safety Analysis":

The applicable safety analyses are the off-site and Control Room radiological dose calculations.

The MSIVs in the main steam lines are required to close when a design basis accident (DBA) occurs. Failure to close an MSIV would affect the retention of the fission product aerosols in that main steam line (and therefore the fission product release to the environment), unless a holdup volume could be established downstream of the closed MSIV in that line, i.e., using the Main Steam ShutoffValve (AfSSV). In determining the most limiting single failure case that would result in the l

highest (most conservative) calculated off-site doses, two cases were examined.

1. The single MSIV failure to close case resuhs in less off-site and Control Room dose (than failure of all four MSSVs to close) because of credit for particulate deposition in the i downstream volumes out to the closed MSSVs.

l

)

2. Failure of all four AISSVs to close is taken as a single active failure based on a single l operator error or on a loss of divisional power, and results in the most limiting dose consequences.

l This failure is not coincident with a single MSIV failure to close. This limiting case assumes that main steam line leakage is attenuated in the main steam line from the reactor vessel out to the outboard MSIV. Although this most limiting analysis case assumed a failure to close the A/SSVs, retention of l l OPERABILITY requirements on these valves is appropriate to ensure the single failure analysis l associated with the LOCA off-site and Control Room dose reanalysis remains valid. The Main l Steam Shutoff Valves meet Criterion 3 of 10CFR50.36 (c)(2)(ii). l l

I l

- }

\

h ain Swm 56doH W ecs ucto , ce i

) g3, v p- r, 7 m -

B 3$6.5[

ff - I Attachment 5 BASES PY-CEl/NRR-2299L Page 11 of 18 t

APPLICABLE

-The MSIV LCS ::tisfic: Critericr 2 cf the "9C Pclig SAFETY ANALYSES St:ta:nt.

(continued) -

^

LCO subsystem

• can provide the required nrnem:ing 3

• ofavailable.

the MSIVasmm4~; ea sure that th.a cepability is Lur4- D ..m a caseh 413"ra, two MSIV LCS )

I

@ ap tems must be OPERABLE.

APPLICABILITY In MODES release to primary 1. containment.

2. and 3. Therefore.

a DBA could ' lead D\hV to a4"""'on OPERABILITY is required during these MODES. M" 4 and 5. the probability and consequences of these events are reduced due to the pressure and temperatur 1~ itations in these MODES. Therefore, maintaining the " :/ '% PERABLE is not required in MODE 4 or 5 to ensure MSIV le- e is processed.

m SS V5 ACTIONS M

+

one MSIV LCS subsystem inoperable. the inoperable L sa-t E. LCS stem must be restored to OPERABLE status wi n 30 days. this Condition, the remaining OPE MSIV LCS subsystem is aquate to perform the require eakage control function. owever, the overall lability is reduced because a sin failure in remaining subsystem could result in a total e of leakage control NW4pM function. The 30 day Comple redundant capability aff ed by Time is based on the remaining OPERABLE MSIV I LCS subsystem and th ow probability a DBA LOCA occurring during s period. An LCO 3.0. axception is 3rovided to lit changes in MODES when the 1 ard MSIV-

_CS subs em becomes inoperable due to condensat ildup i betw the MSIVs when the plant is operated below 50 ' TED MAL POWER.

l l

With two MSIV LCS su .: 'no erabl 't est one subsystem must be restored .e c tus within 7 days.

D*l' O M The 7 day Comple ' one is based on the ow ' ity of the o . of a DBA LOCA.

(continued) l PERRY - UNIT 1 B 3.6-52 Revision No. 1

1 e

Attachment 5 PY-CEI/NRR-2299L Page 12 of 18 gg. .z -

v INSERT D for "LCO": M ha - - - - '

l The four MSSVs are part of the mitigation strategy for off-site and Control Room dose l

consequences. Failure of all four MSSVs to close is taken as a single active failure based on a single operator error or on a loss of divisional power, and results in the most limiting dose  !

consequences. Although this most limiting analysis case assumed a failure to close the MSSVs, l

retention of OPERABILITY requirements on these valves is appropriate to ensure the single failure analysis associated with the LOCA off-site and Control Room dose reanalysis remains valid.

INSERT E for " ACTIONS":

The ACTIONS are modified by a Note allowing separate condition entry for each penetration flow path because an inoperable Main Steam Shatoff Valve (MSSV) in a main steam line does not affect the ability to provide a post-accident holdup volume in the affected line or in the other lines (between the MSIVs or between an MSIVandan MSSV). The Required Actions provide appropriate compensatory actions for each inoperable MSSV. Complying with the Required Actions may allow for continued operation, and subsequent inoperable MSSVs are governed by subsequent Condition entry and application of associated Required Actions.

INSERT F for Required Action"A.1":

Each Main Steam Line has two Main Steam Isolation Valves (MSIVs) and a downstream Main Steam Shutoff Valve (MSSV), i.e., INIl-F0020 A, B, C, or D. With one or more MSSVs inoperable, the invrable MSSVmust be restored to OPERABLE status or the Main Steam Line must be isolat s ,sithin 30 days. During this 30 day Completion Time, the remaining OPERABLEMSIVs in that Main Steam Line are adequate to perform the requiredleakage holdupfunction should a LOCA occur. However, the overall reliability is reduced because a singlefailure ofan MSIVin that line could result in a loss ofthe MSIVleakage holdupfunction, becausepost-accident, a single MSIV failure wouldprevent establishment ofa " holdup vohime" Ifan MSSVis " inoperable", the action ofclosing the MSSVin that main steam line is based on the characteristics of the revised design basis accident source term (i.e., predominantly aerosol). Closing the MSSVwill providefor a post-LOCA single-failure-proofholdup volume within the main steam line, for deposition of the aerosol on the inner walls of the main steam Iine. The RequiredAction does notpermit closure ofan MSSVandan MSIVat the same time, or both MSIVs at the same time, since doing so duringplant operation could result in differential cooldown ofthatparticular Main Steam Line, with resultant damage to some small-bore drain piping that is interconnected to the other Main Steam Lines. ICan MSSV is i " inoperable", but closed, credit can be taken for it in meeting the ACTION. Leak tightness of the MSSVs is not necessary to ensure the assumptions of the dose calculation methodology are met for the main steam lines, since leakage flow characteristics used in the analyses are affected only by the turbulence caused by an open ended pipe (i.e., the Main Steam Shutoff Valves fail to close). l The 30 day Completion Time is based on the redundant capability afforded by the OPERABLE MS/Vs on that line, and the lowprobability ofa DBA LOCA occurring during this period After 30 days, when the MSSVon that line has been closed, apost-LOCA holdup volume can be established without concern over a singlefailure, thereforeplant operation may continue.

i

,1

(Y\ain 54em Shdoff

_fvGS m .

,,._. ...-.n. -o p B ,

%u ..n ak . t [. . d ' 3 Attachment 5 I BASES PY-CEl/NRR-2299L Page 13 or 18 ACTIONS .1 and . 2_ #

(continued)

If th m, mCS :2:yst= cannot be restore to OPERABLE statu ithin the required Completion Tim , the plant must

/fr- % My De brought to a MODE in which the LC0 does not apply. To

( M*" U^M achieve this status, the plant must be brought to at least i et 6 MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating tsch+&A experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.6.1.9.1 REQUIREMENTS V LCS blower is operated for 2 15 minutes to v r .,

OPERABil . 31 day Frequency was devel siderin the known reliabi the MSIV L r and controls, g bW+ b s the two subsystem redundan e low 3robability of a significant degra '

the MSI ) system occurring between ances and has been shown to table operating experience.

. 3.6.1.9.2 The e ctrical continuity of each inboard MSIV LCS subsyst .

heater verified by a resistance check, by verifying rate of te erature increase meets specifications, or y verifying t1 urrent or wattage draw meets specif' ations 1 (e.g.. the inbo d heater draws 8.28 10% amp s per f 5 phase). The 31 da Frequency is based on op ating Mdo" experience that has n that these comp ents usually pass this Surveillance when rformed at th' Frequency.

SR 3.6.1.9.3 A system functional test i erfo ed to ensure that the MSIV LCS will operate t ugl its o ating sequence. This t includes verifying t the automatic sitioning of the valves and the op ion of each inter o and timer are correct, that t blowers start and develop e required flow rate an he necessary vacuum (i .e. . inb d system:

15" H 0 a 100 scfm: outboard system: 15" H2 O a a 200 scfm)2 the u3 stream heaters meet current or wat e draw re ements (wlich may also be used to verify electr al tinuity in SR 3.6.1.9.2). The 18 month Frequency is (continued)

PERRY - UNIT 1 B 3.6-53 Revision No. 1

Attachment 5 PY-CEl/NRR-2299L Page 14 of 18

',..,..j...

.-..._m.. ,

- n ., , ,; s -

1 INSERT G for "SR 3.6.1.9.1": #L- - -

I "' -

The only necessary surveillance requirement is one to ensure the Main Steam Shutoff Valves will l 3 stroke closed on a manual demand by the operators. Leak test requirements are not necessary to ensure the assumptions of the dose calculation methodology are met for the main steam lines, since

)

l leakage flow characteristics used in the analyses are affected only by the turbulence c'aused by an open ended pipe (i.e., the Main Steam Shutoff Valves fail to close). The Frequency ofthis SR is in accordance with she Inservice Testing Program.

t l

l l

l

(A A h be-A A bbd ch B 6.1.9

{ 0 Kl?.l y

• 9 'C' 5 # i?J. g" ' Attachment 5 ggs3 D sed f$ 2 d5I'd 4('$'1 ba k\f BASES PY-CEI/NRR-2299L {

Page 15 of18

= l rE SR 3.6.1.9.3 (continued)

REQUIREMENTS based on .

to perform this ce under the Alde- )

conditions that app '

lant outage and the 3 )

potential p anned trans1 Surveillance erformed with the reactor at power.

REFERENCES 1. Section 6.7 a , Section .

Wh sek H I

r PERRY - UNIT 1 B 3.6-54 Revision No. 1

d Attachment 5 PY-CEl/NRR-2299L Page 16 of18 {

j INSERT H for " References": If.! !.iRidfJi0i!! GIF_V 1.

Calculation PSAT 0420211.08 "Steamline: Particulate Decontamination Control" 2.

Calculation PSAT 0420211.13 "Offsite and Control Room Dose Calculation"  !

3. USAR Section 15.6.5 l i

i l

1 1 l

l l

L _ --_- - - - _ - - - - - _ - - - - - _ - - - - - -

Combustible Gas Mixing System B 3.6.3.3 B 3.6 CONTAINMENT SYSTEMS 9

' 9aQ 7 2 n ,;maAW W W g (e @" ]= 7 Attachment 5 PY-CE!/NRR-2299L Page 17 of18 B 3.6.3.3 Combustible Gas Mixing System BASES BACKGROUND The Combustible Gas Mixing System ensures a uniformly mixed post accident containment atmosphere. thereby minimizing the 30tential for local hydrogen burns due to a pocket of lydrogen above the flammable concentration.

The Combustible Gas Mixing System is an Engineered Safety Feature and is designed to operate following a loss of coolant accident (LOCA) in post accident environments without loss of function. There are two redundant and independent combustible gas mixing subsystems, each consisting and piping. of Each a compressor and associated valves, controls, to pump 500 scfm. combustible gas mixing subsystem is sized Each subsystem is powered from a separate emergency power supply. Since each combustible gas mixing subsystem can provide 100% of the mixing requirements, the system will provide its design function with a worst case single active failure.

Following a LOCA, the drywell is immediately pressurized due to the release of steam into the drywell environment. This pressure is relieved by the lowering of the water level within the weir wall, clearing the horizontal vents and allowing the mixture of steam and noncondensibles to flow into the primary containment through the sup3ression pool, removing much of the heat from the steam. T1e remaining steam in the drywell begins to condense. As steam flow from i the reactor pressure vessel ceases, the drywell pressure {

falls rapidly. The combustible gas mixing compressors are manually started prior to the drywell hydrogen concentration exceeding 3.0 v/o. The compressors force air from the {

primary containment into the drywell. Drywell pressure l j

increases until the water level between the weir wall and -

the drywell is forced down to the horizontal pool vents forcing drywell atmosphere back into containment and mixing '

with containment atmosphere to dilute the hydrogen. While combustible gas mixing continues following the LOCA, hydrogen continues to be produced. Eventually, the 4.0 v/o u y )

limit is again approached and the primary containment

@[ 5 EAT 1 hydrogen recombiners are manually placed in operation.

(continued) 1 1

1

! PERRY - UNIT 1 B 3.6-101 Revision No. 1 u___________________ _ _ _ _ _ _ . J

Attachment 5 PY-CEI/NRR-2299L Page 18 of I8 gr.wn 9 a ;p,fd.en (d(py3**U s(2f'bi..h: %a INSERT I for " Background": 1S I The containment spray is credited with removal of airborne radionuclides. Post-LOCA operation of the mixing compressors also provides a transport of air between containment and the drywell.

Therefore, post-LOCA dose is reduced with mixing compressor operation.

l 2