ML20086G549

Proposed Tech Specs Section 3.6.1.2 Re Allowable Leak Rate for MSIVs

ML20086G549
Person / Time
Site:	Hope Creek
Issue date:	11/25/1991
From:	Public Service Enterprise Group
To:
Shared Package
ML20086G503	List: ML20086G499 ML20086G534 ML20086G549
References
NUDOCS 9112050149
Download: ML20086G549 (55)

Text

.

. INCE(

l.IMIi!NC CCNCIIIONS FOR CPERAIICN AND SURVEILL4NCE DEQUIREMENT5

_5 E C T :CN P t.0 E 3/4.4.6 PRESSURE / TEMPERATURE LIMITS React:r Cociant System. . . . . . . . ..

3.'4 4-21 Figure 3.4.6.1-1 Minimum Reactor Pressure Vessel Metal Temperature Versus React 4r Vessel Pressure. ........ .......... 3/4 4-23 Table 4.4.6.1.3-1 Reactor Vessel Material Surveillance Program Witncrawal Senedule... .. 3/4 4 24 Reactor Steam Dome.... ........ ...... ......... .. .. 3/4 4-25 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES... ... ........ .. . . 3/4 4-26 3/4.4.8 STRUCTURAL INTEGRITY.. . ........ ............... .. ... 3/4 4-27 3/4.4.9 RESIDUAL HEAT REMOVAL Hot Shutdown................ .. ..

................. . . 3/4 4-29 Cold Shutdown.......... ..... ... .................. ..... 3/4 4-29 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5,1 ECCS - OPERATING..... .. ..... ............. .., . ...... 3/45-1 3/4.5.2 ECCS - SHUTDOWN.............. ................ .... .. . 3/4 5-6 3/4,5.3 SUPPRESSION CHAMBER. .... ....................... . .. 3/4 5-3 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Primary Containment Integrity...................... ... . 3/4 6-1 Primary Containment Leakage. .....................,....... 3/4 6-2 Primary Contai nment Ai r Loc ks . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-5

"!!" k:1 4; Sy:td. ~

=_ ~~ w" DELETE Primary Containment Structural Integrity.................. 3/4 6-3 Drpeell and Suppression Chamber Internal Pressure. . . . . . . . . 3/4 6-9 HOPE CREEK xi 1117.05014o 9111; 3 PDR /sDOCK OE.O Mr.4 F POR

l . CONTAINMENT SYSTEMS PRIMARY CONTAINMENT i.EAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2 Primary containment leakage rates shall be limited to:

- An overall5 integrated leakage rate of less than or equal to La , 0.5 per-INSERT 1ja. cent by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at P,, 48 y p,jg,

b. A combined leakage rate of less than or equal to 0.60 L for all penetrationsandallvalveslistedinTable3.6.3-1,ex8eptformain steam line isolation valves", valves which form the boundary for the long-term seal of the feedwater lines, and other valves which are hydrostatically tested per Table 3.6.3-1, subject to Type B and C ests when pr to _, wNy, v

INSERT 2 .

L c.

w'Lc::g mesm"+N th :

the r egn. t: 1 0 ccfA c A combined leakage rate of less than or equal to 10 gpm for all con-17 9 d.

tainment isolation valves which form the boundary for the long-term seal of the feedwater lines in Table 3.6.3-1, when tested at 1.10 Pa, 52.9 psig,

e. A combined leakage rate of less than or equal to 10 gpm for all other containment isolation valves in hydrostatically tested lines in Table 3.6.3-1 which penetrate the primary containment, when tested at Pa, 48.1 psig ap.

APPLICABILITY: When PRIMARY CONTA1HMENT INTEGRITY is required per Specification 3.6.1.1.

ACTION:

With;

a. The measured overall integrated primary containment leakage rate exceeding 0.75 L, or

b. The measured combined leakage rate for all penetrations-and all valves listed in Table 3.6.3-1, except for main steam line isolation valves?, valves which form the boundary for the long-term seal of the feedwater lines, and other valves which are hydrostatically tested ,-

per Table 3.5.3-1, subject to Tvpe B andmmv

" C tests

=- exceeding 0.60 L , or INSERT 3 C'

q' *Cdi"

[3 f

d. The measured combined leakage rate for all containment isolation valves which form the boundary for the long-term seal of the feedwater lines in Table 3.6.3-1 exceeding 10 gpm, or

e. The measured combined leakage rate for all other containment isolation valves in hydrostatically tested lines in Table 3.6.3-1 which penetrate the primary containment exceeding 10 gpm, restore: 3"

a. The overall integrated leakage rate (s) to less than or equal to 0.75 La'

" Exemption to Appencix "J" of 10 CFR 50.

HOPE CREEK 3/4 6-2 I l

l CONTAINMENT SYSTEMS i LIMITING CONDITION FOR OPERATION (Continued)

ACTION (Continued)

b. The combined leakage rate for all penetrations and all valves listed in Table 3.6.3-1, except for main steam line isolation valves", valves which form the boundary for the long-term seal of the feedwater lines, anc other valves which are hydrostatically tested per Table 3.6.3-1, subject to Type B and C tests to less than or equa L,, and f

c. T g g ge g g ' qual t P n t rap-

d. The combined leakage rate for all containment isolation valves which form the boundary for the long-term seal of the feedwater lines in Table 3.6.3-to less than or equal to 10 gpm, and

e. The combined leakage rate for all other containment isolation valves in hydrostatically tested lines in Table 3.6.3-1 which penetrate the primary containment to less than or equal to 10 gpm, prior to increasing reactor coolant system temperature above 200*F.

SURVEILLANCE REQUIREMENTS 4.6.1.2 The primary containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria specified in Appendix J of 10 CFR 50 using the methods and provisions of ANSI N45.4 - 1972:

a. Three Type A Overall Integrated Contain e nt Leakage Rate tests shall be conducted at 40 + 10 month intervals during shutdown at P,,

48.1 psig, during each 10 year service period. The third test of each set shall be conducted during the shutdown for the 10 year plant inservice inspection,

b. If any periodic Type A test fails to meet 0.75 L,, the test schedule for subsequent Type A tests shall be reviewed and approved by the Comission. If two consecutive Type A tests fail to meet 0.75 L,,

a Type A test shall be performed at least every 18 months ur,til two consecutive Type A tests meet 0.75 L,, at which time the above test schedule may be resumed,

c. The accuracy of each Type A test shall be verified by a supplemental test which:

.l . Confirms the accuracy of the test by verifying that the dif ference between the supplemental data and the Type A test data is within 0.25 L,.

2. Has duration sufficient to establish accurately the change in leakage rate between the Type A test and the supplemental test.

3. Require's the quantity of gas injected into the containment or bled from the containment during the supplemental test to be between 0.75 }., and 1.25 L,.

" Exemption to Appendix "J" of 10 CFR 50.

HOPE CREEK 3/4 6-3

CONTAINMENT SYSTEMS LIMITING CONDITION FOR OPERATION 3.6.1.4 Twe inwepen&nt hS!V 'c li s ;y;te: 'MSIVSO ; e lystem; shell be ^

m iwisnusu.

fAOff T T S/ . A h f' M A T 1 All A t P Akin t T i nair t og __a s

%nua64 ii VI kan a vlin g h vgiv a i a vgi.g 4, 4 nily g.

AGT40 W 4h-one- IV-sealing system-subsystem ~inoperatda r -restore-the_.i parable sub- )

syste: t:

_a 0.

9 b .,,.BM-natus-withio40-dapbe-4*-at-leas ERA

.m2 1- P Al N f L1. ! (T fiAU, L1 . ,, i s k.1 , k t-HOT- 00ht witM n-y i. i v .,. 2, m m 9 m

awirL ,- i rs n v 1. 6 i , w rw 1 1 vv.- <1 --- . vi 11 6. i 1.m #, m

,1 y k, mv v.i,g-MN

• pg"- ~-

c mA A nnPMAni VI knnw h*

INSERT $

wem m ,

V bt m a I8 7 45 C rvg wv vv

• N M kC[ O k' { ", ( hb e v'S I OS Sb 0 dOf nh Oh0EOk On IbI:;hh O least ene ceg 'et+--cy:'O ^ . .' ' trn:b \

c Stamuse per a the 3 ssystem thrcu-um rmam-u*uonaute+t#

00;uence, and '>er fying that. ) i c::b i ter4eckn and b.utrits Operat.Nadgned operatoc ac and each antinmatic g/

t^d t: correct positio d /

• d, valve  ::teate:/

Of 'idI! b^ n$ irO COO IO) 'O'bIMOOnbd OI I ' bO I a .

n a rfn.t mm_n..e.n. . f. .n v' 1 n F. U. 8. , M. f l.. f. u.

.. .C. P. V..m .+ i. n. .s. .e, t. ..vn .n..r a 7.m., .S. A, ,k, ai v o. ..

l 9

6. wi Et f tIMPT T AM A I TfeT 11.A D U w w ivuw1 a vii is iw<>

.6 uw-1-.,6 usues vriu s ys.-.-

S A, a,. fw.y . ,wiiw .J l-se i1msus- , 6, --- --- in -- *L

. P.U A U U.r. l. P..A. f. T. O. D. . Avii

. T T. Aw.M view s yui 4v mvi e wis a .

w L

I HOPE CREEK 3/4 6-7 l

l

TABLE 3.6.3-1 (Continued) a PRIMARY CONTAINMENT ISOLATION VALVES l

g MAXIMUM h PENETRATION ISOLATION TIME Q NUMBER (Seconds) NOTE (5) P&,ID VALVE FUNCTION AMD HUMBER

== . . .

M-72-1 (c) 5 stem Isolation Valves DEllyg Outside: 45 1 }

Line A HV-5834A (KP-V010) 45 1 Line B HV-5835A PIB PIC 1 Line (KP-V008) d PID 45 ine D HV-5837A (KP-V007) -A % %[

2 Group 2 - Reactor Recirculation Water Sample System M-43-1 w

(a) Reactor Recirculation Water Sample Line Isolation Valves 2 Inside: 88-SV-4310 P17 15 3 P17 15 3 4 Outside: 88-SV-4311 o

3. Group 3 - Residual Heat Removal (RNR) System (a) RHR Suppression Pool Cooling Water & System Test M-51-1 Isolation Valves

( Outside: 4 HV-F024A (BC-V124) P2128 180 l Loop A: 4 P2128 180 HV-F010A (BC-V125) f Outside: 4 P212A 180 toop B: HV-F0248 (BC-V028) 4 P212A 180 HV-F0108 (BC-V027)

M-51-1 (b) RHR to Suppression Chamber Spray Header Isolation Valves Outside: 3 Loop A: HV-F027A (BC-V112) P2148 75 P214A 75 3 Loop B: HV-F0278 (BC-V015)

3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 PRIMARY CONTAINMENT INTEGRITY PRIMARY CONTAINMENT INTEGRITY ensures that the release of radioactive mate-rials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the accident analyses. This restriction, in conjunction with the leakage rate limitation, will limit the site boundary radiation doses to within the limits of 10 CFR Part 100 during accident conditions.

3/4.6.1.2 PRIMARY CONTAINMENT LEAKAGE The limitations on primary containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the accident analyses at the peak accident pressure of 48.1 psig, P,. As an added conserva-tism, the measured overall integrated leakage rate is further limited to less than or equal to 0.75 L during performance of the periodic tests to account for possible degradation of athe containment leakage barriers between leakage tests.

Operating experience with the main steam line isolation valves has indicated that degradation has occasionally occurred in the le- tightness of the valves; therefore the special requirement for testing thes "alves.

The surveillance testing for measuring leakage rates is consistent with the requirements of Appendix "J" of 10 CFR Part 50 with the exception of exemptions granted for main steam isolation valve leak testing and testing the airlocks after each opening.

3/4.6.1.3 PRIMARY CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the primary containment air locks are required to meet the restrictions on PRIMARY CONTAINHENT INTEGRITY and the primary containment leakage rate given in Specifications 3.6.1.1 and 3.6.1.2. The specification makes allowances for the fact that there may be long periods of time when the air locks will be in a closed and secured positien during reactor operation. Only one closed door in each air lock is required to maintain the integrity of the containment.

3/R1. 4 MSIV SEALING SYSTEM m -. =

g\

Calculate result'ng from the maximum leaka e wance for the' main steamline isolation va 13 the postulated LOC ations would be a small fraction of the 10 CFR 100 gu es, pr the main steam line system from DELETE -7 the isolation valves up to and in -

he turbine condenser remains intact. d Operating experience has i # ed that degr on has occasionally occurred  ;

in the leak tight he MSIV's such that the s ied leakage requirament '

have not alwa een maintained continuously. The seali b m will reduce 1 theged leakage from the MSIVs "

when isolation of the prim (ar)NPjstem Oj and cohtainment is reouired.

HOPE CREEK B 3/4 6-1

, HCPE CREE ( GENERAi!NG STAi!ON l

ENCLO5URE 2 Public Service Electri L LnjLqn LICENSE NO. NPF 57 DOCKET NO. 50-354 SUPPORTING INFORMATION AND ANALYS[S

1.0 INTRODUCTION

AND EQMMARY OF RESULTS The proposed Technical Specification amendment involves an increase in the allowable leakage rate from 46,0 total scfh to 200 scfh per main steam line, and deletion of the MSIV Sealing System, and exemption of the .

downstream main steam piping and condenser from the seismic requirements specified in Appendix A of 10CFR100.

Section 2.0 of this Enclosure provides a summary of background information; Section 3.0 discusses the justifications for the proposed changes; Section 4.0 provides a summary of the plant specific radiological dose assessment, and Section 5.0 summarizes the potential benefits for a Technical Specification MSIV allowable leak rate of 200 scfh, and the deletion of the MSIV Sealing Syste:n.

The BWROG report, NEDC-31358P Rev. 1, "BWROG Report for increasing MSIV

'eakage Rate Limits and Elimination of Leakage Control Systems",

October 1991, provides the justification for increasing MSIV leakage limits, and for eliminating the requirements fo- the MSly Sealing System.

With concurrence from the valve manufacturers, this report concludes that MSIV leakage rates up to 200 scfh are not an indication of substantial mechanical defects in the valve which would challenge the isolation capability of the valve to fulfill its safety function. Therefore, the proposed increase in the allowable leakage rate to 200 scfh for the MSIVs will not inhibit the isolation capability of the valve.

HOPE CREEK GENERATING 3Mi'CN l

The BWROG has evaluated several methods and has recommended the isolated condenser as an alternate method to the LCS for MSIV leakage treatment.

The isolated condenser method takes advantage of the large volume in the main steam lines and the condenser to hold up the release of fission products leaking from the closed MSIVs. The main steam drain lines are employed to convey leakage to the condenser. PSE&G proposes to delete the MSIV Sealing System from the Technical Specification and to incorporate the isolated condenser as an alternate method for MSly leakage treatment.

The BWROG has evaluated the availability of main steam system piping and condenser alternate treatment pathways for processing MSIV leakage, and has reviewed the potential combinations of Loss-Of Coolant Accidents (LOCAs) and seismic events of interest:

(1) LOCA WITHOUT NEAR COINCIDENT SEISMIC EVENT, For this occurrence the pressure in the piping system downstream of the MSIVs is rapidly reduced to atmospheric pressure, and since there is no seismic event, the alternate flow path through main steam system piping to the condenser is assured.

(2) 5EISMIC EVENT WITHOUT NEAR COINCIDENT LOCA. Without a LOCA and the potential associated core degradation, the radioactivity transported with MSly leakage is of no radiological significance.

(3) LOCA WITH NEAR COINCIDENT SEISMIC EVENT. For this occurrence (also assuming significant core damage) the consequences are of interest because a seismic induced failure in the main steam or condenser system could aliow MSIV leakage to bypass the alternate treatment pathway. It has been previously well documented that the probability of a near coincident LOCA and seismic event is extremely small (design basis earthquake probability is approximately 0.001 per reactor per year; core melt probability is plant specific and typically ranges from 0.00001 to 0.0001 per reactor per year).

HOPE CREEK GENERATING $IAil0N l

It is also noted that a LOCA does not induce a seismic event, and that a seismic event has a very low probability of causing a LOCA because the primary pressure boundary and emergency core cooling systems are designed to seismic requirements (NUREG/CR 4792 Volume 4 reported probability of seismic induced LOCA to be less than 5 x 10 7 per reactor per year).

Considering that the probabilit; of a near coincident LOCA and seismic event is much smaller than other plant nafety risks (less than 1 x 10-7 per reactor per year for coincident events, less than 5 x 10-7 per reactor per year for seismic induced LOCA), the concern for main steam piping or condenser damage is of little significance. Nevertheless, because main steam piping and condenser systems designs are extremely rugged, this equipment is expected to remain intact following design basis seismic events. The evolution cf design codes and regulatory requirements is documented in Appendix 0 of NEDC 31858P Rev. 1. It is noted that ANSI-831.1 design requirements have been extensively used for nuclear power plant system design and that this code contains a good deal of margin. In additic' specific seismic design provisions have been incorporated into BWR main steam and condenser systems such as Hope Creek Generat! . ion.

In order to f# 'er justify the capability of the main steam system piping and condenser alternate treatment pathway, the BWROG has reviewed limited earthquake experience data on the performance of non-seismically designed piping and condensers (in past earthquakes). The study summarizes data on the performance of main steam piping and condensers in past strong motion earthquakes and compares these piping and condensers with those in typical U.S. GE Mark I, 11, and 111 nuclear plants. This limited earthquake experience data and similarity comparisons are then used to further strengthen the conclusions on how the GE piping and condensers would maintain their pressure retention function in a design basis earthquake in conjunction with a LOCA occurring just prior to or after the seismic event.

HOPE CREEK GENERAT;NG STAi DN The earthquake experience data are derived from an extensive database on the performance of power plants and industrial facilities, compiled by EQE International (previously EQE Engineering) for the Seismic Qualification Utility Group, the Electric Power Research Institute, and many other E0E clients. This study summarizes the performance of over 100 power plant units (turbines, associated condensers, and main steam piping) in 19 earthquakes around the world from 1934 to the present.

The piping and condensers in the eartnquake experience database exhibited substantial seismic ruggedness, even when they are not designed to resist earthquakes. This is a common conclusion in studies of this type on other plant items such as welded steel piping in general, anchored equipment such as motor control centers, pumps, valves, structures, and so forth. That is, with limited exceptions, normal industrial construction and equipment typically have substantial inherent seismic ruggedness, even when they are not designed for earthquakes. No failures of main steam piping were found. Anchored condensers have also performed well in past earthquakes with damage limited to minor internal tube leakage.

Comparisons of piping and condenser design in example GE Mark I, 11, and 111 plants with those in the earthquake experience database reveal the GE plant designs are similar to or more rugged than those that exhibited good earthquake performance, in addition, each utility will perform a verification of seismic adequacy of the main steam piping and condenser, consistent with th? guidelines discussed in Section 6.7 of NEOC 31858P Rev. 1, to provide reasonable ass'urance of the structural integrity of the'3 components. The BWROG concludes that (1) the possibility of significant failure in GE BWR main steam piping or condensers in the event of an eastern U.S. design basis earthquake is highly unlikely and that (2) any such failure would also be contrary to a large body of historical earthq' lake experience data, and thus unprecedented.

1

HOPE CREEK GENERATING STATICN e

The design basis LOCA has been re analyzed for radiological impacts

- utilizing the isolated condenser as an alternate method for MSIV leakage treatment. The_ analysis demorstrates that a maximum MSIV leakage rate of 200 scfh per main steam line results in an accer.t.ble increase to the dose exposures calculated for a design basis t.0CA. In addition, the analysis demonstrates that MSIV leakage rates of approximately 500'scfh per main; steam line will not result in dose exposures in excess of the regulatory limits. ,

2.0 BACKGROUND

The safety function of the MSIVs is to isolate the reactor system in the event of a LOCA or other avents requiring containment isolation. The design of the MSIVs and its isolation requirements are described in Sections 5.4.5'and 6.2.4.3.1.1 of the Updated Final Safety Analysis Report (UFSAR). The allowable leakage rate from MSIVs is included in the-LOCA radiological analysis evaluated 'in Section 15,6.5.5 of the VFSAR,-

The safety-related MSIV Sealing System is designed to eliminate the

~

release of fission products through the MSIVs that would bypass reactor building-recirculation and filtration after a LOCA. This is accomplished

-by pressurizing the sections of the main steam lines between the inboard and the outboard MSIVs, and between the outboard MSIVs and the main steam stop valves, to a pressure above.that of the reactor vessel. The MSIV Sealing System is described in Section 6,7 of the UFSAR, and consists of inboard and outboard subsystems. The inboard subsystem contains-Lisolation valves which provide for containment integrity in the event that the' corresponding MSIV fails'to close.

Oper. ting experience: indicates that'MSIVs frequently exceed the Technical Specification allewable leak rates. Some of these valves repeatedly fail the localLleak rate tests despite frequent disassembly and refurbishment.

As a result of increasing MSIV leakages and _the inability of- the LCS to function at high MSIV leakages, the Nuclear Regulatory Commission prioritized Generic Issue C-8, "MSIV Leakage and LCS Failure" as a high L

' priority item in January 1983. This issue was closed in 1990.

- - ~ = ~ _ . . _ . - . - . , . . , . . _ - . . . . . - . . - . - - , . . - , _ - . . -,

HOPE CREEK GENERAi!NG STAT GN The BWROG formed a M51V Leakage Committee in 1982 to address the increasing MSIV leakage rates, and a follow on MSIV Leakage Closure Committee in 1986 te address alternate actions to resolve on-going, but less severe MSIV problems. The MSly Leakage Committee identified contributors which cause MSIVs to fail the leak rate tests by large margins, developed recommendations to minimize leakages, evaluated alternates for MSIV leakage treatments, and compiled recent history of MSIV leakages and LCS operating experience.

3.0 JUSTIFICATIONS FOR THE PROPOSED CHANGE PSE&G proposes to increase the Technical Specification allowable leakage rate for the MS!Vs from 46.0 scfh to 200 scfh per main steam line and to delete the MSIV Sealing System requirements from the Technical -

Specifications.

The current Technical Specification allowable MSIV leak rate is extremely limiting and routinely requires unnecessary repair and re-test of the MSIVs. This significantly impacts the maintenance work load during plant outages and often contributes to outage extensions. In addition, the needless dose exposures to maintenance personnel are inconsistent with As low As Reasonably Achievable (ALARA) principles. From a safety perspective, calculations using standard conservative assumptions for considering the offsite consequences of a postulated design-basis LOCA confirm that offsite and control room doses will be within the regulatory guidelines for t> allowable MSly leakage rate. This calculation is described in Se:, ion 15.6.5.5 in tne UFSAR. However, if MSIV leakages are only moderately higher than the allowable limit, the calculated doses will exceed the regulatory guidelines. Furthermore, as documented in Generic Issue C-8, the LCS will not function if MSIV exceeds the leakage limit by a moderate amount.

HOFE CREEA ;ENEGI N C m ia MSIV's failure to meet the current Technical Sper.ification limit have been documented in response to surveys conducted by the Nuclear Regulatory Commission during the early 1980s and by the BWROG during the middle and late 1980s. As high as 50% of the total "as found" HSiv local leak rate tests were reported in the early NRC survey to exceed the leakage rate limit.

The GWROG has studied the iswes regarding MSIV leakage rates, their causes, and available alternatives. The results of the BWROG study are prnvided in NECC 31858P Rev. I and are summarized in NUREG ll69, in response to Generic issue C 8, the BWROG has recommended corrective a:tions and maintenance practices to reduce the M31V leakage rates.

A recent survey conducted by the BWROG of MSly leakage tests performed between 1984 and 1988 indicates that the implementation of industry c.,d WROG actions has been offe:,'ve in reducing the leakage rates, and, in particular, a reduction in the number of valves which experienced

] substantial h leakage rates. The survey concludes that about 23% of the total "as id" MSIV leakages still exceeded the limit of 11.5 scfh and alout 10% exceeded 100 scfh.

The leakage performance at HCGS is consistent with the recent survey by the BWROG.

Despite the recent improvement in leakage performance, MSly leakage rates still frequent's exceed the current Technical Specification limit and the safety and maihtenance problems related to high MSIV leakage rates, although lesc severe, iemain as a significant issue.

Furthermore, based on the extensive evaluation of valve leakage data, the BWROG has found that disassembling and refurbishing the MSIVs to meet very low leakage limits frequently contribute to repeated failures. By

OPE CREEA GENERAilNG LIAi'ON not having to disassefrble the valves and refurbtsh them for minor leakage, the utility may avoid introducing one of the root auscs of recurring valve leakage problems. These unnecessary repairs can lead to later leak test f ailures and possibly congromise plant safety.

The current Technical Specification allowable leakage rate is established by excessively conservative LOCA radiological analysis as described in Section 15.6.5.5 of the UFSAR. The valve's large physical size, fast (_

acting characteristics, and the availability of existing turbine building equipment were not considered at the time the leakage limit was established. Based on the in depth evaluation of MSly leakages, the BWROG has concluded the MSIV leakage rates up to 500 scfh are not an indication of substantial mechanical defectt in the v41ve which would challenge the isolation capability of the valve to fulfill its safety function. Furthermore, valve manufacturers have stated that leakage rates up to 200 scfh can occur without having a major valve defect.

Therefore, the proposed increase from 46.0 scfh total to 200 scfh per main steam line will not inhibit the MSIV's performance of the isolation function and will not compromise the safety of HCGS.

This proposed increase provides a mort realistic, but still conservative, 4 limit for the MSIVs. Based on the BWROG study, the proposed increase in the allowable leak rate will increase the chance for a successful local leak rate test to greater than 90%, up from the 77% success rate at the ,

current typical limit of 11.5 scfh. The increase in successful local leak rate testing will significantly reduce MSIV maintenance cost, reduce dose exposure to maintenance personnel, reduce outage durations, extend effective service life of the MSIVs, and minimize the potential for outage extensions at HCGS.

A safety-related LCS was requir$d oy Regulatory Guide 1.96 in order to reduce the radiological consequences of MSIV leakages. As discussed earlier, Generic issue C-8 identified the safety concern that MSIV leakage rates, as determined by conservative local leak rate tests, were 1

HOPE CREE *. GENEuilNG STAi!GN too high and that the LCS would not function at high MS!V leakage rates.

The 1981 NRC survey indicated that 33 percent of the total tests exceeded leakage rates of 100 scfh. Since the process capability of the LCS (MS!V Sealing System at HCGS) is designed for MS!V leakage rates of no more than 100 scfh. the potential existed for the LCS not to function as analyzed for a design basis LOCA as described in Section 15.6.5.5 of the UfSAR. Consequently, the conservatively calculated dose contribution from MSIV leakage would exceed the regulatory limits for offsite and control room doses.

PSE&G proposes to delete the MS!V Sealing System requirements from the Technical Specifications. The proposed changes involve a replacement of the existing MSly Sealing System with the more reliable and effective main steam drain line and condenser for MSIV leakage treatment. This treatment method is effective to treat MSIV leakage over an expanded operating range. Except to the requirement to establish a proper flow path from the main steam isolation valves to the condenser, the proposed method is passive and does not require any logic control and interlocks.

The method is consistent with the philosophy of protection by multiple leak tight barriers used in containtnent design for limiting fission product release to the environment. Therefore, the proposed method is highly reliable for MSIV leakage treatment.

The existing MSIV Sealing System has limitation for mitigating MSIV leakage. Operation of the systen increases containment pressure and thereby increases the containment leakage. The MSIV Sealing System requires multiple logic controls, interlocks, timers, containment isolation valves, and other equipment to ensure containment integrity and protection from_the high pressure main steam lines. Based on plant operating experience, the System does not provide a high degree of reliability. Also the System has limited capacity and does not function at moderate MSIV leakage rates above 100 scfh.

HOPE CREli. GENE w N ;?n ;a Even though the resulting effsite doses may be slightly higher for low HSIV leakage rates, the effectiveness af the proposed method even for leakage rates greater t un the proposed increased allowable limits.

ensures off site dose limits to the public are not exceeded. Overall, the proposed treatment method can handle MSIV leakage over an expanded operating range, and will thereby resolve the safety conc 6rn that the LCS will not function at M$ly leakage rates higher than the LCS capacity.

Thus, a margin of safety ex ,ts, furthermore, it is clearly a safety improvement to replace a system with known limitations with the alternate main steam piping and condenser treatment pathway, which has been shown to have excellent reliability.

Furthermore, PSE&C will incorporate the applicable alternate leakage treatment methods, consistent with GE document NED0 30324 " Potential Operator Actions to Control MSIV Leakage", into the Operational Procedures and Emergency Operational Procedures.

In addition to improving plant safety, the proposed deletion of the MSly Sealing System requirements in the Technical Specifications will result in significant operational and maintenance benefits. The BWROG has evaluated recent LCS performance data. Results of this evaluation are shown in NEDC 31858P Rev. 1. The evaluation indicates that leakage control systems are difficult to maintain. Plant shutdowns and start up delays have occurred. The MSly Sealing System performance at HCGS is consistent with the recent survey by the BWROG.

In conclusion, PSE&G proposes to increase the Technical Specification allowable MSly leakage rate from 46.0 scfh total to 200 scfh per main steam line, and to eliminate tne requirements for MSly Sealing System in the Technical Specifications. The proposed increase in the MSly leakage limit should significantly reduce recurring valve leakage problems, and will minimize needless valve repair which can compromise plant safety.

The proposed deletion of the safety related MSIV Sealing System and the 21-

HOP [ (R([A CEN(RATING STAi!CN proposed alternate method (main steam lines and condenser) will resol,e the safety concern regarding LCS effectiveness at higher M$ly leakage rate.

4.0 ANALYSIS OF MSIV LEAKAqLCONTRIBUTION TO RADIOLOGICALELLCALCVL ATIONS 4.1 Selection of Alternate lealige Treatment Method The BWROG has evaluated several alternate MS!V leakage treatment methods and has recommended the isolated condenser for MSIV leakage treatment.

This leakage treatment method takes advantage of the large volume in the isolated main condenser to hold up the release of fission products leaking from the closed MSIVs. The main steam drain lines are employed to convey leakage to the condenser. .

As previously discussed in Section 1.0, the BWROG has evaluated the availability of main steam system piping and condenser alternate treatment pathways for processing MSIV leakage. The BWROG has determined that the probability of a near coincident LOCA and a seismic event is much smaller than for other plant safety risks. The BWROG has also determined that main steam piping and condenser designs are extremely rugged, and that the ANSI 831.1 design requirements typically used for nuclear plant system design contain a good deal of margin. In order to further justify the capability of the main steam piping and condenser alternate treatment pathway, we have reviewed limited earthquake experience data on the performance of non seismically designed piping and condensers (in past earthquakes). This study concluded the possibility of a failure which could cause a loss of steam or condensate in BWR main steam piping or condensers in the event of a design basis earthquake is highly unlikely, and that such a failure would also be contrary to a large body of historical earthquake experience, and thus unprecedented.

This conclusion is consistent with NUREG/CR 4407, " Pipe Break Frequency Estimation for Nuclear Power Plaats", dated May 1987, which reported no 22-

HOPE CREES. GENERAi!NG SIA!iGN observed failures in the main steam piping over 313 years of reactor operating years. Therefore, the isolated condenser alternate MSIV leakage treatment path at HCGS is considered appropriate for the reduction of radiological consequences of a design basis LOCA.

4.2 Rjtdiolooical Analysis and Raigll The radiological dose methodology developed by General Electric for the BWROG is documented in Appendix C of NEDC-31858P Rev. 1.

The radiological analysis calculates the effects of the proposed allowable MS!V leak rate in terms of control room and off site doses.

The revised LOCA doses are the sum of the LOCA doses (as described in Section 15.6.5.5 of the VFSAR) and the calculated MS!V leakage doses.

This method of calculating the revised dose exposures is very conservative since the LOCA doses already include the dose contribution from MSIVs at the maximum leakage rate permitted in the current Technical Specifications.

Table 1 shows the calculated dose exposures from the BWR00 radiological analysis for HCGS. Regulatory limits and calculated doses from LOCA radiological analysis are also included in Table 1 for comparison purpose. This analysis demonstrates that a MSIV leakage rate of 200 scfh per main steam line results in an acceptable increase to the dose exposures previously calculated for the control room, EAB, and the LPZ.

The revised LOCA doses remain well within the guidelines of 10CFR100 for offsite doses and 10CFR50, Appendix A (General Design Critaria 19) for the control room doses. Furthermore, the calculation th' that MSIV leskage rates up to approximately 500 scfh per steam lin,. .ould not exceed the regulatory limits.

Therefore, the proposed method provides a substantial safety margin for-mitigating the radiological consequences of MSIV leakage beyond the proposed Technical Specification leak rate limit of 200 scfh,

c HOPE CREEK G[NERAT]NG STATION l

5.0 BENEFITL OF THE PROPOS(D CHANGLS As discussed in NEDC 31858P Rev. 1, recent MSIV leakage performance has significantly improved since the early BWROG survey in 1984 and the NRC l survey in the early 1980s. Despite the recent improvement, MSIV leakage rates exceeding the current Technical Specification limits still frequently occur. The BWROG evaluation of the recent MSIV leakage performance concludes that the proposed change will improve the chance ,

for a successful local leak rate test to greater than 90%, up from the 77% success rate at the current Technical Specification limit of 200 l

scfh. l Implementation of BWROG recommended changes to hardware and maintenance activities has improved BWR MSIV leakage experience and this specifically applies for Hope Creek Generating Station. However, even with the changes, MSly leakage frequently exceeds the current limits and this excessive leakage has recently resulted in outage extensions and unnecessary radiation exposures for refurbishment of the valves.

Deleting the MSly Sealing System will reduce the overall dose rates, and eliminate the system's impact on refueling and maintenance outage activities at HCGS. Although the revised LOCA doses are slightly higher for low MSIV leakage ra.te, the effectiveness of the proposed method, even for leakage rates greater than the proposed increased MSIV allowable leak rate, ensures that off-site and control room doses are not exceeded.

HCPE C;EE( GENERai'NG 5 TAT:CN Table 1 CONTRIBUTION TO THE LOCA DOSE EXPOSURES FOR A MAXIMUM MSIV LEAK RATE OF 200 scfh HOPE CREEK GENERATING STATION Whole Body Thyroid Beta

[tamJ frem) _ M Exclusion Area 10CFR 100 Limit 25 300

• A)

Boundary (2 Hour) B) Previous Calculated 0.6 76.7 Domes **

C) Contribution from 0.1 2.6 MSIVs at 200 scfh D) New Calculated Doses 0.7 79.3 Low Population 10CFR 100 Limit 25 300

• A)

Zone (30 Day)

B) Previous Calculated 0.08 7.7 Doses **

C) Contribution From 0.34 57.5 MSIV at 200 scfh D) New Calculated Doses 0.42 65.2 Control Room A) GDC 19 5 30 30/75***

(30-Day)

B) Previous Calculated 0.04 0.26 .91 Doses **

C) Contribution From 0.10 2.71 1,64 MSIVs at 200 scfh D) .New Calculated Doses 0.14 2.97 2.55

• No limit specified.

• • UFSAR Section 15.6.5.5 ano 6.4 (includes MSly leak rate at a total of 45 scfh for the first 20 minutes; control room dose assumes 100% per day reactor building inleakage).

• • • 75 if prior commitment has been made to use protective clothing.

25-

HOPE CREEA GENERATING STAilGN Enclosure 3 Public Service Electric and GAs LLCENSE NO. NPF 57 DOCKET NO. 50-354 NO SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS PSE&G proposes to replace the MS!V Sealing System with the more reliable and effective main steam drain line and condenser for MSIV leakage treatment.

This treatment method is effective to treat MSIV leakage over an expanded operating range. Except for the requirement to assure that certain valves are opened to establish a proper flow path from the main steam isolation valves to the condenser, the proposed method is passive and does not require any logic control and interlocks. The method is consistent with the philosophy of protection by multiple leak tight barriers used in containment design for limiting fission product release to the environment. Therefore, the proposed method is highly reliable for MSly leakage treatment.

Public Service Electric and Gas proposes an amendment to the Technical Specifications as follows:

1) Revise Section 3.6.1.2(c) to permit an increase in the allowable leak rate for the main steam isolation valves (MSIVs) from the current 46.0 standard cubic feet per hour (scfh) to 200 scfh per main steam line.

2) Delete Sections 3/4.6.1.4 and 83/4.6.1.4 to reflect the deletion of the MSIV Sealing System from the Technical Specifications.

3) Revise Index pages xi and xii, Section 3.6.1.2(c) and Sections 3,6.1.4, 4.6.1.4 and B3/4.4.6.1.4

4) Revise Section 3.6.1.2(a) to exempt MSIV leakage from the acceptant criteria for the overall integrated leakage rat"

, "CTE CREEA GENERAilSG STAi:04 s

The proposed amendment to add a footnote to Section 3.6.1.2(a) clarifies that MSly leakage is exempt from the overall integrated containment leakage rate test acceptance criteria in Appendix J of 10CFR50. As defined in the Bases Section 3/4.6.1.2 of the Technical Specifications, the purpose of these requirements is to ensere that the overall integrated containment leakage will not exceed the designed containment leak rate assumed in the design basis loss of-coolant accident (LOCA) radiological analysis. Since en allowable leak rate is specifically allocated for the MSIVs in the Technical Specifications, and since the radiological analysis has been revised to analyze MSiv leakage path separately from those of the containment leakage rates, the proposed exemption is, therefore, appropriate and justified. This proposed change does not exempt the MSIVs from the test schedules as required in the Technical Specifications and 10CFR50 Appendix J.

(Pursuant to 10CFR50.12(a), PSC&G bas applied for an exemption request). This application also provides a detailed justification for exempting the MSIV leakages from the acceptArce criteria of 10CFR50 Appendix J, and demonstrates that the proposed exemption will not present an undue risk to the public health and safety. Therefore, based on the above consideration, the proposed amendment to Section 3.6.1.2(a) is considered an administrative change.

This application provides a detailed justification for exempting the downstream main steam piping and condenser from the seismic requirements specified in Appendix A of 10CFR100, and demonstrates that the proposed exemption will not pre:ent an undue risk to the public health and safety.

Therefore based on the above consideration, the proposed amendment to Section s

3.6.1.2(a) is considered an administrative change.

PSE&G has, pursuant to 10CFR Part 50.92, reviewed the proposed amendment to determine whether our request involves a significant hazards considerations.

The ooeration of Hoce Creek Generatino Station. in accordance with the orocosed amendment, will not involve a sionificant increase in the orotability or conseouences of an accident oreviously evaluated.

27-

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

HOPE CREE ( SEMRAihG miRN The proposed amendment to Section 3,6.1.2(c) does not involve a change to structures, components, or system. that would affect the probability of an accident previously evaluated in the Jpdated Final Safety Analysis Report (UFSAR).

The proposed amendment to delete ,ections 3/4.6.1.4 and Bases Sections B 3/4.6.1.4 involves eliminating the MS!V Sealing System requirements from the Technical 90ecifications. As described in Section 6.7 of the UFSAR, the MSIV Sealing System is manually initiated in about 20 minutes following a design-basis LOCA. Since the MSIV Sealing System is operated only after an accident. has occurred, this proposed amendment has no effect on the probability of an accident.

Since MSIV leakage and operation of the MSIV Sealing System are included in the radiological analyais for the design-basis LOCA as described in Section 15.6.5.5 of the UFSAR, the proposed amendments will not affect the precursors of other analyzed accidents. The proposed amendments result in acceptable radiological consequences of the design basis LOCA previously evaluated in Section 15.6.5.5 of the UFSAR.

The Hope Creek Generating Station has an inherent MS!V leakage treatment capability. PSE&G proposes to use the main steam lines and condenser as an alternate to Regulatory Guide 1.96 " Design of Main Steam isolation Valve Leakage Control System For Boiling Water Nuclear Power Plants" for MSIV leakage treatment. PSE&G will incorporate this alternate method in the Operational Procedures and Emergency Operational Procedures.

The BWROG has evaluated the availability of main steam system piping and condenser alternate treatment pathways for processing MSly leakage, and has determined that the probability of a near coincident LOCA and a seismic event is much smaller than for other plant safety risks. -

The BWROG has also determined that main st6am piping and condenser designs are extremely rugged, and that the ANSI B31.1 design requirements typically used for nuclear plant system design ccntain a good deal of margin.

28-

. - - - - - = . ___ . _- - - - - - - - - - .. . ~ _ - - -

HOPE CREEK GENERA 1hG ST Ai!Cf4 in order to further justify the capability of the main steam piping and condenser alternate treatment pathway, the BWROG has reviewed limited earthquake experience data on the performance of non seismically designed piping and condensers (in past earthquakes). This study concluded that the possibility of a failure which could cause a loss of steam or condensate in BWR main steam piping or condensers in the event of a design basis earthquake is highly unlikely, and that such a failure would also be contrary to a large body of historical earthquake experience data, and thus unprecedented.

PSE&G will also perform a verification of seismic adequacy of the main steam piping and condenser, consistent with the guidelines discusss_' in Section 6.7 of NE00 31858P Rev.1, to provide reasonable assurance of the structural integrity of these components.

A plant specific radiological analysis has been performed to assess the effects of the proposed increase to the allowable MSIV leak rate in terms of contrcl room and offsite doses following a postulated design basis LOCA. This analysis utilizes the hold up volumes of the main steam piping and condenser as an alternate method for the MSIV leakages. As discussed earlier, there is reasonable assurance that the main steam piping and condenser will remain intact following a design basis earthquake. The radiological analysis uses standard conservative assumptions for the release of source terms consistent with Regulatory Guide 1.3 Revision 2, " Assumptions Used for Evaluating the Potential Radiological Consequences of-a loss-Of Coolant Accident for Boiling Water Reactors", dated April 1974.

The analysis demonstrates that dose contributions from the proposed MSly leakage rate limit of 200 scfh per steam line and from the proposed deletion of the MSIV Sealing System result in an acceptable increase to the LOCA doses previously evaluated against the regulatory guidelines for the offsite doses and control room doses as contained in 10CFR100 and 10CFR50, Appendix A (General Design Criteria 19), respectively. The LOCA doses previously evaluated are discussed in Section 15.6.5.5.5 of the UFSAR. The revised LOCA

.- . ~ . - . . . ~ - - . - - - . - . . - - . - - - - - - - . - - - - . - - . - - - . . - - . - . . .

+

HOPE CREEK OENERAi!NG STAT:CN doses are the sum of the LOCA doses previously evaluated in the Uf5AR and the additional MS!V doses calculated using the alternate treatment method. This method of calculating the revised doses is very conservative since the LOCA doses previous evaluated already included dose contributions from M51V at the maximum leakage rate permitted in the current Technical Specifications. The I l

attached table shows the previous calculated doses and the new calculated doses. j i

t s

30-

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

HOPE CREE ( SENERAilNG SIAIiCN CONTRIBUTION TO THE LOCA DOSE irr, m ' FOR A MAXIMUM M51V LEAK RATE OF 200 scfh HOPE CREEK GENERATING STATION l'

Whole Body Thyroid Beta irJLml frem) frem)

Exclusion Area 10CFR 100 Limit 25 300

• A)

Boundary (2 Hour) B) Previous Calculated 0.6 76.7 Doses **

C) Contribution From 0.1 2.6 MSIVs at 200 scfh D) New Calculated Doses 0.7 79.3 Low Population A) 10CFR 100 Limit 25 300

• Zone (30 Day)

B) Previous Calculated 0.08 7./

Doses **

C) Contribution from 0.34 57.5 MSIV at 200 scfh D) New Calculated Doses 0.42 65.2 Control Room A) GDC 19 5 30 30/75***

(30-Day)

B) Previous Calculated 0.04 0.26 .91 Doses **

C) Contribution-From 0.10 2.71 1.64 MSIVs at 200 scfh D) New Calculated Doses 0.14 2.97 2.55 No limit specified.

• • UFSAR Section 15.6.5.5 and 6.4 (includes MSly leak rate at a total of 45 scfh for the first 20 minutes; control room dose assumes 100% per day reactor building inleakage).

• • • 75 if prior commitment has been made to use protective clothing.

31-

HOPE CR(( A G!N[pATING $TA'l;'<

I 1

The whole bouy dose at Se low Population Zone (LPZ) and the control room is increased from 0.08 to 0.42 rem and from 0.04 to 0.14 rem respectively. These increases are acceptable because the revised doses are well within the regulatory guidelines (0.42 versus 25 rem at the LPZ and 0.14 versus 5 rem at the control room). The associated whole body dose at the exclusion area boundary (EAB) increased insignificant 1y from 0.6 to 0.7 rem.

The thyroid dose at the LPZ increased from 7.7 to 65.2 rem. This increase is 1 acceptable because the revised dose of 65.? rem is significantly less than of the regulatory guideline (300 rem). The EAB thyroid dose increased slightly from 76.7 to 79.3 rem, whereas the control room thyroid dose increased from 0.26 to 2.97 rem. The increase in control room thyroid dose is acceptable because the revised dose remains a small fraction (9.9%) of the limit. The control room beta dose is increased from 0.91 to 2.55 rem, which remains insignificant relative to the regulatory guideline of 30 rem.

It is important to note that the resulting doses are dominated by the organic iodine fractions which occur because of the ultraconservative source term assumptions used in this analysis. For 200 scfh per steam line, more than 85%

of the off-site iodine and control room doses are due to organic iodine from the RG 1.3 source term and organic iodine converted from the elemental iodine deposited in main steam piping systems, if the actual iodine composition from the fuel release (cesium iodide) is used in the calculations, essentially all of this organic iodine dose would be eliminated.

In summary, the proposed changes result in an acceptable increase to the radiological consequences of a LOCA previously evaluated in the UFSAR. The revised LOCA doses are well within the regulatory guidelines. Although the revised LOCA doses are slightly higher for low MSIV leakage rates, the effectiveness of the proposed method, even for leakage rates greater than the proposed increased MSly allowable leak rate, ensures that off site and control room doses are not exceeded.

32-

HOP [ CR((*, G[%[UilNS $iAIlON The proposed amendment to Table 3.6.3-1 of the Technical Specifications involves the deletion of MSIV Sealing System valves from the list of containment isolation valves. This proposed change is consistent with the proposed deletion of the HSly Sealing System. The inboard system lines which are connected to the main steam piping will be disabled with blind flanges to assure that containment integrity is maintained.

This proposed change does not involve an increase in the probability of an l accident previously evaluated in the UFSAR. This proposed change has no effect in the consequences of an accident since the inboard MSIV Sealing System lines will be disabled with blind flanges, thus assuring that the containment integrity, isolation, and leak test capability are not compromised for the postulated accident.

The proposed change to 3.6.1.2(a) is administrative in nature and has no effect on any accident.

The proposed changes to Index pages xi and xii, Section 3.6.1.2 and Sections 3.6.1.4, 4.6.1.4 and 83/4.4.6,1.4. are administrative in nature and have no effect on any accident. These changes provide new section and page number designations due to the deletion of Sections 3.6.1.4 and Bases Sections B 3/4.6.1.4.

The operation of HQDe Creek Generat10.0 Station. in accordance with the DroDosed amendment will not create the oossibility of a new or different kind of accident from any accident oreviou}ly evalut.ith The= proposed amendment to Section 3.6.1.2(c) does not create the possibility for a new or different kind of accident from any accident previously evaluated.

The BWROG evaluated MSIV leakage performance and concluded that MSly leakage rates up to 200 scfh will not inhibit the capability and isolation performance of the valve to isolate the primary containment. There is no new modification HOPE CREEA SENERA!;NG STA7lCN which could impact the MSIV operability. The LOCA has been analyzed using the main steam piping and condenser as a treatment method to process MSIV leakage at the proposed maximum rate of 200 scfh. Therefore, the proposed change does not create any new or different kind of accident from any accident previously evaluated in the UFSAR.

The proposed amendment to delete Sections 3/4.6.1.4 and Bases Sections B3/4.6.1.4 does not create the possibility of a new or different kind of accident from any accident previously evaluated because the removal of the MSIV Sealing System does not affect any of the remaining systems at HCGS and the LOCA has been analyzed using the alternate method to process MSIV leakage.

The proposed amendment to delete the MSIV Sealing System isolation valves from Table 3.6.3 1 does not create the possibility of a new or different kiri of accident. The inboard MSlV Sealing System piping will be isolated with a blind flange to assure that the primary containment integrity, isolation, and leak testing capability are not compromised, therefore eliminating the possibility for any new or different kind of accident.

The proposed change to Section 3.6.1.2(a) is administrative in nature, and does not create a possibility of a new or different kind of accident from any accident previously evaluated in the FSAR The proposed changes to the index pages, and the revision of section numbers are administrative in nature, and do not create the possibility of a new or different kind of accident from any accident previously evaluated.

The operation of Hooe Creek Generatina Station in accordance with the crocosed amendment. will not involve a sianificant reduction in the marain of safety.

The proposed amendment to Section 3.6.1.2 does not involve a significant reduction in the margin of safety. As discussed in the Bases of the Technical Specification 3/4.6.1.2, the allowable leak rate limit specified for the MSivs

-34

HOPE CREE < GENERATING STA*!34 is used to quantify a maximum amount of bypass leakage assumed in the LOCA radiological analysis. Results of the analysis are evaluated against the dose guidelines contained in 10CFP.100 for the offsite doses and 10CFR50, Appendix A (General) Design Criteria 19) for the control room doses. The margin of safety is considered to be the difference between the calculated doses and the guidelines as contained in 10CFR100 and GDC 19.

Results of the radiological analysis demonstrate that the proposed change does not involve a significant reduction in the margin of safety. The margins of safety with respect to whole body doses, are insignificant 1y reduced by 1.4%

at the LPZ, 2.0% in the control room, and 0.4% at the EAB. The thyroid dose '

margin of safety is reduced by 19.2*. at the LPZ, 9.0% in the control room, and 0.9% at the EAB. The beta dose margin of safety is insignificantly reduced by 5.5% in the control room. The margins of safety are net significantly -

affected because the absolute margins of safety remain well below the guidelines (lowest whole body margin of safety is 97.2% at the EAB, lowest thyroid margin of safety is 73.6% at EAB).

Therefore, the ?roposed amendment does not involve a significant reduction in the margin of safety at HCGS.

The proposed amendment to delete Sections 3/4.6.1.4 and B3/4.6.1.4 does not reduce the margin of safety. In fact, the overall margin of safety is increased. The function of the MSly Sealing System for MSIV leakage treatment will be performed by the alternate main steam lines and condenser equipment.

This treatment method is effective in reducing dose consequences of MSIV leakage over an expanded operating range and will, thereby, resolve the safety concern that the LCS will not function at MSIV leakage rates higher than the LCS design capacity. Except for the requirement to establish a proper flow path from the MSIVs to the condenser, the proposed method is passive and does not require any logic control and interlocks. The method is consistent with the philosophy of protection by multiply leak tight barriers used in containment design for limiting fission product release to the environment.

Therefore, the proposed method 4 highly reliable and effective for MSIV leakage treatment.

35-  !

HCFE ;REE=; GENERAi:N3 t'ai::N The revised LOCA doses remain well within the regulatory limits for the site and control room. Furthermore, the calculation shows that MSIV leakage rates up to approximately 500 scfh per steam line would not exceed the regulatory limits. Therefore, the proposed method provides a substantial safety margin for mitigating the radiological consequences of MSIV leakage beyond the proposed Technical Specification leak rate limit of 200 scfh.

The proposed amendment to delete MSIV Sealing System valves from Table 3.6.3-1 does not involve a significant reduction in the margin of safety. Isolation of the inboard system lines via use of a blind flange assures that the primary containment integrity, isolation, and leak testing capability are not compromised; therefore, it does not result in a reduction in the margin of safety.

The proposed change to Section 3.6.1.2(a) is administrative in nature and does not affect the margin of safety.

The proposed amendment to the index pages, and the revision of section and page numbers is administrative in nature, and does not have any impact on the margin of safety.

HOPE CREEN GENE;.ATING STAi:CN (N1LQ1VRE 4 Public Service Electric and Qn LICENSL NO. NPF 57_

QQ Eli NO. 50 354 APPLICATION FOR EXEMPTION TO APPENDIX A 0F 100FR100 Public Service Electric and Gas (PSE&G), holder of Facility Operating License No. NPF 57, hereby requests an exemption of the downstream main steam piping and condenser from the seismic requirements specified in Appendix A of 10CFR100, " Seismic and Geologic Siting Criteria for Nuclear Power Plants "

Specifically, PSE&G proposes to employ probability analysis, existing design margin, seismic experience, and a plant specific seismic adequacy verification, as alternate methodology to the dynamic analysis or qualification test specified in Paragraph Vl(a)(1) of 10CFR100 Appendix A, to provide reasonable assurance that the existing main steam piping and condenser will remain functional following a design basis accident coincident with a significant seismic event.

The exemption would allow the existing, non seismically' designed main steam piping and condenser to be used for mitigating and radiological consequences of MSIV leakage during the duration of a Design Basis Accident, such that the resulting doses are within the guidelines of 10CFR100.

PSE&G recognizes that there is no provision in 10CFR100 for exemption; however, the Nuclear Regulatory Commission (NRC) has the authority to grant this exemption.

PSE&G proposes to replace the MSIV Sealing System with the more reliable and effective main-steam drain line and condenser for MSIV leakage treatment.

This treatment method is effective to treat MSIV leakage over an expanded 37-

HCFE CREEA H NE;AiiNG STA*:CN operating range. Except for the requirement to assure that certain valves are opened to establish a proper flow path from the main steam isolation valves to the condenser, the proposed method is passive and does not require any logic control and interlocks. The method is consistent with the philosophy of protection by multiple leak tight barriers used in containment design for limiting fission product release to the environment. Therefore, the proposed method is highly reliable for MSly leakage treatment.

The existing MSIV Sealing System has limitations for mitigating MSly leakage.

Operation of the system increases containment pressure and thereby increases the containment leakage. The MSly Sealing System requires multiple logic controls, interlocks, timers, containment isolation valves, and other equipment to ensure containment integrity and protection from the high pressure main steam lines. Based on plant operating experience, the System does not provide a high degree of reliability. Also the System has limited capacity and does not function at moderate MS!V leakage rates above 100 scfh.

Even though the resulting off-site doses may be slightly higher for low MSIV leakage rates, the effectiveness of the proposed method even for leakage rates greater than the proposed increased allowable limits, ensures off site dose limits to the public are not exceeded. Overall, the proposed treatment method c:.n handle MSIV leakage over an expanded operating range, and will thereby resuive the safety concern that the LCS will not function at MSIV leakage rates higher than the LCS capacity. Thus, a margin of safety exists.

Furthermore, it is clearly a safety improvement to replace a system with known limitations with the alternate main steam piping and condenser treatment pathway, which has been shown to have excellent reliability.

In conjunction with this application for exemption request, PSE&G has transmitted to the NRC an application for a license amendment pursuant to 10CFRSO.90. This license amendment involves a proposed change to Section 3.6.1.2 of the Technical Specifications to permit an increase in the allowable leak rate for the MSIVs from the current 46.0 standard cubic feet per hour 38-

HOPE CREEK SENERAi!NG 5IAilCN (scfh) total to 200 scfh per main steam line, and a proposed change to Section 3.6.1.4 for eliminating the requirements for the MSIV Scaling System, the safety analysis has been revised to assess the radiological effects of MS!V leakage following a postulated design basis LOCA. PSE&G has demonstrated that the proposed change does not involve a significant hazards consideration.

This proposed exemption is c result of the extensive work performed by the BWR Owners' Group (BWROG) in support of the resolution of Generic Issue C 8 "MS!V Leakage and LCS failure".

The following discussion provides a detailed justification and evaluation of the proposed exemption. While recognizing this exemption criteria is specifically applicabic to 10CFR50, PSE&G has evaluated the proposed exemption in accordance with the criteria specified in 10CFR50.12(a). The proposed exemption will not present an undue risk to the public health and safety and is consistent with the common defense and security. Furthermore, special circumstances are present that warrant the granting of this exemption.

The proposed exemption will not cause additional operational activities that may significantly affect the environment. It does not result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Impact Statement Operating License Stage, a significant change in effluents or power levels, or effect any matter not previously reviewed by the NRC that may have a significant adverse environmental impact.

Upon the NRC approval of the license amendment and exemption requests, PSE&G will perform a verification of seismic adequacy of the main steam piping and cond9nser, consistent with the guidelines discussed in Section 6.7 of NE00-31858P Rev. 1, to provide reasonable assurance of the structural integrity of.these components.

Therefore, PSE&G hereby requests an exemption to the seismic requirements of 10CFR100 Appendix A for Hope Creek Generating Station to permit the use of existing, non-seismically designed main steam piping and condenser to mitigate the radiological consequences of MSly leakage.

39-

HOPE CREIK GENEM i aG siAt::N A. Anu fic at ion Parajraphs VI(a)(1) of 10CFR100 Appendix A ieauires that structures, systems and components, which assure the capability to prevent or mitigate the consequences of accidents which could result in potential off-site exposures of 10CFR100, be designed to remain functional following a safe shutdown earthquake (SSE) and concurrent loads. The engineering method used to assure that the required safety functions are maintained following the SSE shall involve the use of either dynamic analysis or a suitable qualification test to demonstrate that structure, systems, and components can withstand the seismic and other concurrent loads.

The BWROG has evaluated the capability of main steam piping and condensers to process MSIV leakage following a design basis accident coincident with a .

seismic event. Based on this comprehensive evaluation, the BWh0G has concluded there is reasonable assurance that the main steam piping and condenser will remain functional following a design basis accident coincident with a seismic event, as great as the design basis earthquake, to mitigate the radiological consequences of MSIV leakage. The following conclusions provide the bases for this assurance:

(1) Probability for which the resulting dose from MSIV leakage is significant is extremely low. This requires a design basis LOCA, a degraded core where ECCS are not functional, and a significant seismic event.

(2) Main steam piping and condensers are designed to strict inc.strial standards and building codes; thus, a significant design margin exists.

(3) Main steam piping and condensers exhibit substantial seismic ruggedness. Comparisons of piping and condenser design in GE plants with those in the earthquake experience database reveal that the GE plant designs are similar to or more rugged than those that have exhibited good earthquake performance.

HOPE GEEK GENE:AIDiG STAT:CN (4) Possibility of significant f ailure in GE BVR main steam piping or condensers in the event of design basis earthquakes is highly unlikely, and any such failure would also be contrary to a large body of historical earthquake experience data, and thus unprecedented.

(5) A plant specific verification of seismic adequacy of the main steam piping and condenser will be performed to provide reasonable assurance of the structural integrity of these components.

In conclusion, there is a reasonable assurance that the existing, non seismically designed main steam piping and condenser will remain functional following a design basis accident coincident with a seismic event, as great as the design basis earthquake, to mitigate the radiological consequences of MSIV leakage.

In support of the above, the BWROG has reviewed the potential combinations of Loss Of Coolant- Accidents (LOCAs) and seismic events of interest:

(1) LOCA WITHOUT NEAR COINCIDENT SEISMIC _ EVENT. For this occurrence the pressure in the piping system downstream of the MSIVs is rapidly reduced to atmospheric pressure; and since there is no seismic event, the alternate flow path through main steam system piping to the condenser is assured.

(2) SEISMIC EVENT WITHOUT NEAR COINCIDENT LOCA. Without a LOCA and the potential associated core degradation, the radioactivity transported with MSIV leakage is of no radiological significance.

(3) LOCA WITH NEAR COINCIDENT SElSMIC EVENT, For this occurrence (also assuming significant enre damage) the consequences are of interest because a seismic induced failure in the main steam or ccndenser system could allow MS!V leakage to bypass the alternate treatment pathway, it has been previously well documented that

HOPE CREEK GENERAT[V STAi!CN the probability of a near coincident LOCA and seismic event is extremely small (design basis earthquake probability approximately 0.001 per reactor per year; core melt probability is plant specific and typically ranges from 0.00001 to 0.0001 per reactor peryear). It is also noted that a LOCA does not induce a seismic event, and that a seismic event has_a very low probability of causing a LOCA because the primary pressure boundary and emergency core cooling systems are designed to seismic requirements (NUREG/CR 4792-Volume 4 reported probability of seismic induced LOCA to be less than 5 x 10*7 per reactor per year).

Considering that the probability of a near coincident LOCA and seismic event

-is much smaller than other plant safety risks (less than 1 x 10*7 per reactor per year _for coincident events, less than 5 x 10*7 per reactor per year'for seismic induced LOCA), the concern for main steam piping or condenser damage is of little significance. Nevertheless, because main steam piping and condenser systems designs are extremely rugged, this equipment is expected to remain intact following design basis seismic events. The evolution of design codes and regulatory requirements is documented in Appendix 0 of NEDC 31858P Rev. 1. It is noted that ANSI 831.1 design requirements have been extensively used for nuclear power plant system design and that thi . ode contains a good deal of margin, in addition, specific seispic design provisions have been incorporated into some newer BWR main steam and condenser systems.

'To further justify the capability of the main steam system piping and condenser alternate treatment pathway, the BWROG has reviewed limited earthquake experience data on the performance of non seismically designed piping and condensers (in past earthquakes). The study, documented in Appendix D of NEDC 31858P Rev. 1, summarizes data on the performance of main steam piping and condensers in past strong motion earthquakes and compares these piping-and condensers with those in typical U.S. GE Mark I, 11, and-l!!-

-nuclear plants _This limited earthquake experience data and similarity comparisons are tten used to further strengthen the conclusions-on how the GE piping and condensees would maintain their pressure retention function in a design basis earthquake in conjunction with a LOCA occurring just prior to or after the seismic event.

42

--.-----_--,s,----_ _ - - ---__-,a---u-_.---------- --,---- u. - - - - - - - - - - . _ , - - - - - - - . - - - - - - - . - - - - - - --__u. .,_.----.-----------u------u----su-,---,-----w_.--

HOPE C W SENE UT:NG 5 Ai!0N ,

The earthquake experience data are derived from an extensive database on the performance of power plants and industrial facilities, compiled by E0E for the Seismic Qualification Utility Group, the Electric Power Research Institute, and many other E0E clients. This study st.mmarizes the performance of over 100 power olant units (turbines, associated condensers, and main steam piping) in 19 earthquakes around the world from 1934 to the present.

The piping and condensers in the earthquake experience database exhibited substantial seismic ruggedness, even when they are not designed to resist earthquakes. This is a common conclusion in studies of this type on other plant items such as welded steel piping in general, anchored equipnient such as motor control cet.ters, pumps, valves, structures, and so forth. That is, with limited exceptms, norrr.al industrial construction and equipment typically have substartial inherent seisn.ic ruggedness, even when they are not designed for earthquakes. No failures of mail steam piping were found. Anchored '

condensers have also performed well in past earthquakes with damage limited to minor internal tube leakage.

Comparisons of piping and condenser design in example GE Mark I, II, and 111 plants with those in the earthquake experience database reveal the GE plant designs are similar to or more rugged than those that exhibited good earthquake performance. The BWROG concludes that (1) the possibility of ,

significant failure in GE BWR main steam piping or condensers in the event of an eastern U.S. design basis earthquake is highly unlikely and that (2) any such failure would also be contrary to a large body of historical earthquake experience data, and thus unprecedented.

Earthquake experience methodology has been applied in seismic equipment qualification issues-associated with Unresolved Safety Issue A 46 (Seismic Qualification of Equipment in Operating Plants). Piping performance data are presented in NUREG 1061-(a report from the NRC piping Review Committee), and this report proposes changes to criteria that are directed toward the recognition of the superior performance of piping in earthquakes and establishing more realistic seismic criteria for piping qualification. The

HOPE CREES. GENE;Ai:NG STAi::N NRC has published NUREG 1030 and NUREG 1211 " seismic Qualification of Equipment in Operating Nuclear Power Plants." nhich conclude that the seismic experience data approach provides the most reasonable and preferred alternative to other current equipment qualification methcds.

The rapidly growing use of the seismic experience data approach is further illustrated by the fact that this method of analysis is now referenced in:

A. Draft RG 1.100, Revision 2 " Seismic Qualification of Electrical and Mechanical Equipment in Nuclear Power Plants"

8. Recent approved version cf IEEE Standard 344 1997, " Recommended Practice for Seismic Qualificction of Class IE Equipment for Nuclear Power Generating Stations" C. Draft report of ASME Standard " Recommended Practice for Seismic Performance Qualification of Mechanical Equipment Used in Nuclear Power Plants."

The SQUG earthquake experience database .icludes a large number and variety of piping systems. In fact, piping is probably the strongest area in this regard (compared to areas like electrical or mechanical equipment, cable trays, etc.), it has been concluded that the earthquake experience data on piping, and in particular data on main steam piping, are applicable to main steam piping in BWRs.

In both nuclear and conventional power plants, the condenser is designed to reduce the low psssure turbine outlet pressure (thereby increasing turbine efficiency) and to condense the steam. The nuclear environment does not impose additional significant design considrirations on the condenser. With the exception of hotwell size, a conventionti plant and nuclear plant with similar performance parameters have similar condensers.

44-

HCFE CREE ( LENERA?!NG STAilCN None of the condensers within the seismic experience database has seismic design criteria. However, in view of the performance of the condensers within the database, it is concluded that the condensers have an inherent seismic ruggedness and that the earthquake experience data on condensers are applicable to condensers in BWRs.

Another recent study to develop, by datt collection and statistical analysis, updated estimates of pipe breaks in commercial U.S. nuclear power plants was completed in 1987. This study evaluates both LOCA sensitive systems and non LOCA sensitive systems. For BWR non LOCA sensitive systems, ten pipe failures have occurred over 313 years of operating experience. None of these failures occurred in the main steam piping. Based on the observed failure rates, this study estimated the failure rate for the main steam system piping 1 to be 0.0007 failure / year /BWR with an upper bound of 0.0096 failures / year /BWR, ,

These results are consistent with the conclusion from the SQUG databases and NUREG 1169: BWR main steam piping is designed to withstand severe plant transients such as turbine trips and is expected to remain intact following accidents as severe as a design basis LOCA. Thus, the non seismically designed main steam piping and the main condenser can be used to mitigate the consequences of MSIV leakage.

PSE&G will also perform a verification of seismic adequacy of the main steam piping and condenser, consistent with the guidelines discussed in Section 6.7 of NEDC-31858P Rev. 1, to provide reasonable assurance of the structural integrity of these components.

In conclusion, there is reasonable assurance that the existing, non-seismically designed main steam piping and condenser will remain functional following a design basis accident coincident with a seismic event, as great as the design basis earthquake, to mitigate the radiological consequences of MSly leakage.

45-

HOPE CREEK GENERA!ING STAT GN B. No Undue Risk to Public Health and Safety The BWROG has evaluated the capability of main steam piping and condensers to process MSIV leakage following a d sign basis accident coincident with a i seismic event. Based on this comprehensive evaluation, the SWROG has conclude there is reasonable .3: urance that the main steam piping and condenser will remain functional tve10 wing a design basis accident coincident with a seismic event, as great as the design basis earthquake, to mitigate the radiological consequences of MSlV leakage. The following conclusions provide the bases for this assurance:

(1) Probability for which the resulting dose from MSIV leakage is significant is extremely low. This requires a design basis LOCA, p a degraded core where ELCS are not functional, and a significant seismic event.

(2) Main steam piping and cc.7densers are designed to strict industrial standards and building codec; thus, a significant design margin exists.

(3) Main steam piping and condensers exhibit substantial seismic ruggedness. Comparisons of piping and condenser design in GE plants with those in the earthquake experience database reveal m that the GE plant designs are similar to or more rugged than these that have exhibited good earthquake performance.

(4) Possibility of significant failure in GE BWR main steam piping or ,

condensers in the event of design basis earthquakes '.s highly unlikely, and any such failure would also be contrary to a large body of historical earthquake experience data, and thus unprecedented.

(5) A plant-specific verification of seismic adequacy of the main steam piping and condenser will be performed to provide reasonable assurance of the structural integrity of these componants.

1 l

. HOPE CREE ( GENEUT!NG siAi;CN The treatment method for HSlV leakages is recommended by the BWROG in support of the resolution to Generic issue C-8 "HSIV Leakage and LCS Failure". The proposed changes involve a replacement of the existing LCS with the more reliable and effective main steam piping and condenser for MSiv leakage treatment. This treatment method is effective to reduce dose consequences of MSIV leakage over an expanded operating range and will, thereby, resolve the safety concern that the LCS will not function at MSIV leakage rates higher than the LCS design capacity. Except for the requirement to assure certain values are opened to establish a proper flow path from the MSIVs to the condenser, she proposed method is passive and does not require any logic ,

control and interlocks. This utility specific requirement is discussed in detail in Section 6 of NEDC-31858P. The method is consistent with the philosophy of protection by multiple leak-tight barriers used in containment design for limiting fission prcduct release to the environment. Therefore, the proposed method is highly reliable and effective for MSIV leakage treatment, in conclusion, the proposed exemption presents no unoue risk to public health and safety. ',

C. Consistent with Comon Defense and Security With regard to the "comon defense and security" standard, the grant of the ,

requested exemption is consistent with the comon defense and security of the United States. The Comission's Statement of Considerations in support of the exemption rule note with approval the explanation of this standard as set fortn in Lona Island liahtina Comoany (Shoreham Huclear Power Station, Unit 1), LBP 84-45, 20 NRC 1343, 1400 (October 29,1984). There, the term "comon defense and security" refers principally to the safeguarding of special nuclear material, the absence of foreion control over the applicant, the protection of Restricted Data, and the availability of special nuclear material for defense needs. The granting of the requested exemption will not affect any of these matters and, thus, such grants are consistent with the comon defense and security.

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

HOPE CREE =; 3ENEW'N3 iW::N D. Soecial Circumstances Are Present Special circumstances are present which warrant issuance of this requested exemption. These special circumstances are discussed in accordance with the classification contained in 10CFR50.12(a)(2):

(ii) Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule.

Compliance with Appendix A of 10CFR100 for the downstream main steam piping and condenser is not necessary to achieve the underlying purpose of the rule.

The underlying purpose of the rule is to limit releases to within the off-site dose limits of 10CFR100. The regulation requires components that mitigate the consequences of an accident to within the dose limits of 10CFR100 be designed to the seismic requirements of 10CFR100 Appendix A. The regulation is intended to provide a reasonable assurance that the components will remain functional for the mitigating function. For the purpose of mitigating the radiological consequences of MS!V leakage, it is not necessary to apply the seismic requirements of 10CFR100 Appendix A to the main steam piping and condenser in order to achieve the underlying purpose of the rule because:

(1) There is reasonable assurance that the existing, non-seismically designed main steam piping and condenser will remain functional following a design basis accident coincident with a seismic event, as great as the design basis earthquake, to mitigate the radiological consequences of MSIV leakage. This assurance is based on methodology using probability analysis, margins in the existing design codes, seismic experience, and a plant specific verification of seismic adequacy.

(1) The safety analysis has been revised to assess the radiological consequences of MSIV leakage following a design basis LOCA. The analysis has demonstrated that the revised doses are well within the off-site dose guidelines of 10CFR100.

, HOPE CREE ( 'ENERAi!NG STATIGN furthermore, the seismic approach is consistent with the current resolution of the seismic and equipment qualifin tion issues. Earthquake experiences data have applied in seismic equipment qualification issues associated with Unresolved Safety issues A-46 (Seismic Qualification of Equipment in Operating Plants). Piping performance data have been presented in NUREG 1061, e report from the NRC Piping Review Committee, which proposes changes to criteria that are directed toward the recognition of the superior performance of piping in earthquakes and establishing more realistic seismic criteria for piping qualification. The NRC has published NUREGs 1030 and 1211 " Seismic Qualification of Equipment in Operating Nuclear Power Plants," which conclude that the seismic experience data approach provides the most reasonable and preferred alternative to other current equipment qualification methods.

(iii) Compliance would result in undue hardship or other costs that are significantly in excess of those contemplated when the regulation was adopted, or that are significantly in excess of those incurred by others similarly situated.

The proposed MSIV leakage treatment method utilizes the existing main steam piping and condenser for the mitigating function. Compliance with the seismic requirements of 10CFR100 Appendix A for the main steam piping and condenser would require significant upgrade of the existing equipment, lead to unnecessary long-term plant shutdown for modification, and significantly increase maintenance requirements and the associated costs in order to meet seismic qualification requirements. 1 (iv) The exemption would result in benefit to the public health and safety that compensates for any decrease in safety that may result from the granting cf the exemption.

PSE&G has transmitted to the NRC an application for a license amendment which involves proposed changes to the Technical Specifications to increase the allowable MSIVs leak rate from 46.0 scfh total to 200 scfh per steam line and to-delete the requirements for the MSiv Sealing System. For the MSIV leak I

l

l , HOPE CR[E( GENE W 'M i W '04 rate limits, this application is partly based on the f act that the current limit is too restrictive, and results in excessive MSiv maintenance and repair, leading to additional MSIV failures, which in turn result in higher leakages. The proposed limit will benefit the public health and safety by reducing the potential for MSIVs failures, and thus keeping the MSIVs . leakage within the radiological analysis values.

For the MSIV Sealing System, the proposed changes involve a replacement of the existing System with the more reliable and effective main steam piping and condenser method for MSIV leakage treatment. The effectiveness of the proposed method even for leakage rates greater than the proposed increased allowable limits, ensures off-site dose limits to the public are not exceeded.

Overall, the proposed treatment method can handle MSIV leakage over an expanded operating range, and will thereby resolve the safety concern that the ,

M51V Sealing System will not function at MSIV leakage rates higher than the System capacity. Thus, a margin of safety exists. Furthermore, it is clearly a safety improvement to replace a system with known limitations with the alternate main steam piping and condenser treatment pathway, which has been shown to have excellent reliability. The exemption from 10CFR100 Appendix A seismic requirements for the downstream main steam piping and condenser is regaired so that Hope Creek Generating Station can operate with the proposed Technical Specifications limit of 200 scfh per steam line and with the alternate MSly treatment method. This benefit will compensate for any decrease in safety that may result from the-granting of this exemption.

Thus, special circumstances exist warranting the granting of the exemption.

E. Environmental Imoact The proposed exemption has been analyzed and determined not to cause additional construction or operational activities which may significantly affect the environment. It does not result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Impact Statement-0perating License Stage, a significant change in effluents or s

-S0-l

HOPE CREEK GENERAiiNG STATION l 1

~

1 1

power levels, or a matter not previously reviewed by the Nuclear Regulatory Commission which may have a significant adverse environmental impact.

The proposed exemption does not alter the land use for the plant, any water uses or impacts on water quality, air or ambient air quality. The proposed action does not affect the ecology of the site and v1cinity and does not affect the noise emitted by station. Therefore, the proposed exemption does not affect the analysis of environmental impacts described in the environmental report.

HOPE CREEK GENERAilNG STAi!ON

[NC LOSU_RL5 fMBLIC SERy_1[LE1LCIB.1C ANJ,pji LICENSE NO. NPF 57 DOCKET NO. SQfl51 APPLICATION FOR EXEMPTION TO APPENDIX J OF 10CFR50 Pursuant to Section 50.23(a) of the Regulatior.s of the Nuclear Regulatory Commission, PSE&G, holder of f acility Operating License No. NPF-57, hereby requests specific exemptions to Appendix J of 10CFR Part 50 " Primary Reactor Containment Leakage Testing For Water-Cooled Power Reactors".

Specifically, PSE&G requests that leakages from the main steam isolation valves (MSIVs) be exempted from the acceptance criteria for the overall integrated leak rate-test (Type A), as defined in the regulations of 10CFR50, Appendix J, Paragraphs !!!.A.5(b)(1) and Ill.A 5(b)(2).

The purpose of the test acceptance criteria is to ensure that the measured leak rate from the containme.it volume will not exceed the designed containment leak rate assumed in the safety analysis for a postulated design basis Loss-Of-Coolant Accident (LOCA).

In conjunction with this application for exemption request, PSE&G has transmitted to the Nuclear Regulatory Commission an_ application for a license amendment pursuant to 10CFR50.90. This license amendment involves a proposed change to Section 3.6.1.2 of the Technical Specifications to permit an increase in the allowable leak rate for the MSIVs from the current 46.0 total standard cubic feet per hour (scfh) to 200 scfh per main steam line, and a proposed change to Section 3.6.1.4 for eliminating the requirements for the MSIV Sealing System. The safety analysis has been revised to assess the radiological effects of MSIV leakage following a postulated design basis LOCA.

PSE&G has demonstrated that the proposed change does not involve a significant hazards consideration.

l HOPE CREEK GENERATING 57ATION This proposed exemption is a result of the extensive work performed by the BWR Owners' Group (BWROG) in support of the resolution of Generic issue C-8 "MS!V Leakage and LCS Failure". ,

The following discussion provides a detailed justification and evaluation of the proposed exemption. The proposed exemption is found to be authorized by law, will not present an undue risk to the public health and safety, and is consistent with the common defense and security. Furthermore, special ,

circumstances are present that warrant the granting of this exemption.

The proposed exemption will not cause additional operational activities that may significantly affect the environment. It does not result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Impact Statement-0perating License Stage, result in a significant change in effluents or power levels, or affect any matter not previously reviewed by the Nuclear Regulatory Commission that may have a significant adverse environmental impact.

Therefore, pursuant to 10CFR 50.12(a), PSE&G hereby requests an exemption for HCGS for MSIV leakages from the acceptance test criteria specified in Appendix J of 10CFR50.

A. Justification The regulation of 10CFR50, Appendix J. Paragraphs Ill.A 5(b)(1) requires the overall integrated leakage rate, as measured during containment pressure tests (Type A), to meet the acceptance criterion of less than or equal to 0.75 of the maximum allowable containment leak rate.

As described in the Bases Sections B 3/4.6.1.2 of the Technical

' Specifications, the limitations on primary containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the accident analyses at the peak accident pressure. As an added HOPE CREE ( 3ENERATING siai!GN conservatis.5 the me sured leak rate is further limited to less than or equal to 0.75 of the maximum allowable leak rate during the performance of the periodic tests to account for possible degradation of the containment leakage barrier between leakage tests.

The maximum containment leakage rate was included in the radiological analysis of a postulated design basis LOCA as evaluated in Section 15.6.5 of the Final Safety Analysis Report (FSAR), The radiological analysis calculated the effect of the maximum leakage rate from the containment volume in terms of control room and off site doses, which were evaluated against the dose guidelines of 10CFR50, Appendix A (General Design Criteria 19) and 10CFR100, respectively. Leakages from the containment volume were contained in the reactor building (secondary containment), filtered by the Standby Gas Treatment System, and then released *.o the environment. The maximum containment leakage rate includes leakages through structures, all penetrations identified as Type B, and all containment isolation valves identified as Type C.

The safety analysis has been revised to account for the radiological effect from HSIV leakages and from those of other containment leakages following a postulated design basis LOCA, Unlike the treatment path for other containment leakages, the treatment of MSIV leakages employs the main steam drain piping and the condenser. Fission products are removed by plate-out and hold-up in the relatively large volumes of the main steam piping and condenser.

The treatment method for HSIV leakages is recomended by the BWROG in support of the resolution to Generic Issue C-8. The BWROG has evaluated the availability of main steam system piping and condenser alternate treatment pathways for processing MSIV leakage, and has determined that the probability of a near coincident LOCA and a seismic event is much smaller than for other plant safety risks. The BWROG has also determined that main steam piping and condenser designs are extremely rugged, and that the ANSI-831.1 design requirements typically used for nuclear plant system design contain a good deal of margin.

+

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

HOPE CREEK GENERATING STAi!CN >

In order to further justify the capability of the main steam p1 ping and condenser alternate treatment pathway, the BWROG has reviewed limited-earthquake experience data on the performance of non seismically designed piping and condensers (in past earthquakes). This study concluded.the possibility of a failure which could cause a loss of steam or condensate in BWR main steam piping or condensers in the event of a design basis earthquake is highly unlikely, and that such a failure would also be contrary to a large body of historical earthquake experience data, and thus unprecedented.

Leakage should not be included in the Type A acceptance criterion because -the treatment path for MSIV leakage is different from that of containment leakage.

Potential leakage from the containment is contained in the reactor building (secondary containment), treated by the SGTS, and released to the environment.

MSIV leakage is contained, plated out, and delayed in the main steam piping .

and the condenser, and released via the turbine building.

The delet' .. ' the MSIV Sealing System is proposed partly in response to the safety co..cern identified by Generic Issue C 8 that the MSIV Sealing System would not function at high MSIV leakage rates since the process capability of-the System at HCGS is' designed for_MSIV leakage rate of no more than 100 scfh.

MSIV leakage is treated- separately from other containment leakages. therefore any exemption which was-previously_ granted in accordance with Paragraph III.C 3_ of Appendix J of 10CFR50 should ' remain applicable.

As discussed earlier, the basis _for the containment leakage tests and the -

acceptance criteria is to ensure that the measured leak rate will not exceed the maximum leak rate assumed in the safety analysis. The safety analysis for

.a. design basis LOCA has been revised to include the maximum MSIV leak rate

l. separately-from the maximum containment leak rate. MSIV leakages will' be o tested as part of the local leak rate test in accordance with the requirements L -in Section 3.4.6.1.2. of the-Technical Specifications. This test ensures that the measured MSIV leak rate will not-exceed the allowable leak rate assumed in the-safety analysis.

L

, HOPE CREEK GENERATING STA!!ON There is sufficient conservatism in the maximum allowable MSIV leak rate to account for possible degradation of the MSIV leakage barrier between leakage tests. As discussed in the application for the license amendment, PSE&G proposes a maximum allowable MSIV leak rate of 200 scfh per main steam line; whereas, the analysis demonstrates that MS!V leakage rates up to approximately 500 scfh per main steam line will not result in dose exposures in excess of the regulatory limits. Thus, a safety margin exists. Furthermore, PSE&G will institute into the MSIV maintenance and test program, the requirement that any MSIV exceeding the proposed 200 scfh limit, will be repaired and re-tested to meet a leakage rate of less than or equal to 11.5 scfh. This will assure continuation of high quality repair and refurbishment efforts to improve the overall performance and reliability of the MSIVs-Therefore, the proposed exemption from the acceptance criteria of 10CFR50, Appendix J will not defeat the underlying purpose of the regulation, and is consistent with the safety analysis.

B. Authorized By Law The proposed axemotion is consistent with Section 3.6.1.2 of the Standard Technical Specification (NUREG-0123). The reason for this exemption is provided in the Technical Specifications Bases B 3/4.6.1.2. A review of the Technical Specifications for BWRs indicates that such an txemption has been granted to the following plants: Fermi 2, Hatch I & 2,, Limerick 1, Shoreham, LaSalle 1 and 2, Hanford, Clinton, Grand Gulf 1, Perry, Dresden 2 and 3, Monticello, Quad Cities 1 and 2, Brunswick 1 and 2 and Nine Mile Point 2.

Therefore, the proposed exemption is authorized by law.

C. No Undue Risk to Public Health and Safety The proposed exemption presents no undue risk to public health and safety.

The revised MSIV leakage rate has-been incorporated in the radiological analysis for a postulated LOCA as an additien to the designed containment leak

. =- . . ..- -

HOPE CREEK GENERATING STATICN rate. The analysis demonstrates an acceptable increase to the dose exposures previvJsly Calculated for the Control room and off-site. The revised LOCA doses remain well within the guidelines of 10CFR100 for off site doses and 10CFR50, Appendix A, (General Design Criteria 19) for the control room doses.

In addition, Section 3.6.1.2 of the Technical Specification has provided for allowable MSIV leak rates, which assure that the HSIVs isolation function is not compromised. Finally, potential MSIV leakage is subjected to plate-out, and hold-up in the main steam piping and condenser, thus minimizing their effect en the total dose released. As discussed in Section F of this application, the proposed change will not adversely affect the conclusions of the previously issued FES-OL. Therefore, the proposed exemption presents no undue risk to public health and safety.

D. Cgnsistent with Common Defense and Security With regard to the " common defense and security" standard, the granting of the requested exemption is consistent with the common defense and security of the United States. The Commission's Statement of Considerations in support of the exemption rule note with approval the explanation of this standard as set forth in lona Island Lichtino Company (Shoreham Nuclear. Power Station, Unit 1), LBP-84-45, 20 NRC 1343, 1400 (October 29,1984). There, the term " common defense and security" refers principally to the safeguarding of special nuclear material, the absence of foreign control over the applicant, the protection of Restricted Data, and the availability of special nuclear material for defense needs. The granting of the requested exemption will not affect any of these matters and, thus, such grants are consistent with the common defense and security.

E. Special Circumstances Are Present Special circumstances are present which warrant issuance of this requested exemption. These special circumstances are discussed in accordance with the classification contained in 10CFR50.12(a)(2):

1 t

HOPE CREEK GENERAT!NG iTATION (ii) Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule.

The underlying purpose of the rule is to limit releases to within the off site and control room dose guidelines of 10CFR100 and 10CFR50, Appendix A (GDC 19),

respectively. Compliance with Appendix J of 10CFR50 for Type A test acceptance criteria is not necessary to achieve the underlying purpose of the rule since MSIV leakage is not directed into the reactor primary containment.

Instead, the MSIV's leakage is directed through the main steam drain piping into the condenser. Since Type A tests are intended to measure the primary containment overall integrated leak rate (ILRT), the MSIV's leakage rate should not be included in the measurement of the ILRT.

The safety analysis has been revised to assess the radiological consequences of MSIV leakage following a design basis LOCA. The analysis has demonstrated that the revised LOCA doses are well within the off-site and control room dose guidelines of 10CFR100 and GDC 19.

(iii) Compliance would result in undue hardship or other costs that are significantly in excess of those contemplated when the regulation was adopted, or that are significantly in excess of those incurred by others similarly situated.

Compliance with Appendix J of 10CFR50 Type A test acceptance criteria results in undue hardship or other costs that are significantly in excess of those contemplated when the regulation was adopted. The proposed increase in the MSIV allowable leak rate will not be possible if the MSIV leak rate results are included in the Type A test acceptance criteria.

Compliance requires unnecessary repair and re-testing of the MSIVs. This significantly impacts the maintenance work load during plant outages and often contributes to outage extensions. The frequent MSIV disassembly and refurbishing, which is required to meet the low leakage limits, contributes to repeated failures.

r 1 o HOPE CREEK GENERATING STAi!ON l Examples of these maintenance induced defects include machining induced seat cracking, machining of guide ribs, excessive pilot valve seat machining, and mechanical defects induced by assembly and disassembly. By not having to disassemble the valves and refurbish them for minor leakage, HCGS avoids introducing one of the root causes of recurring leakage. Industrial experience suggests that, by attempting to correct non-existing or minimal defects in the valves, it is likely that some actual defects may be introduced that lead to later leak test failures, In addition, the frequent maintenance work results in needless dose exposures to maintenance personnel leading to additional economical burdens, and are inconsistent with As low As Reasonably Achievable (ALARA) principles.

(iv) The exemption would result in benefit to the public health and safety that compensates for any decrease in safety that may result from the granting of the exemption.

PSE&G has transmitted to the NRC an application for a license amendment which involves proposed changes to the Technical Specifications to increase the allowable MSIVs leak rate from 46.0 total scfh to 200 scfh per main steam line and to delete the MSIV Sealing System. For the MSIV leak rate limit, this application is partly based on the fact that the current limit is too restrictive, and results in excessive MSIV maintenance and repair, leading to additional MSIV failures, which in turn result in higher leakages. The proposed limit will benefit the public health and safety by reducing the potential for MSIVs failures, and thus keeping the MSIV leakage within the radiological analysis values.

For the MSIY Sealing System, the proposed changes involve a replacement of the existing System with the more reliable and effective main steam piping and condenser method for MSIV leakage treatment. The effectiveness of the proposed method even for leakage rates greater than the proposed increased allowable limits, ensures off-site dose limits to the public are not exceeded.

Overall, the proposed treatment method can handle MSIV leakage over an O

HOPE CREEK GENERATING S1ATION 4

expanded operating range, and will thereby resolve the safety concern that the MSIV Sealing System will not function at MSIV leakage rates higher the System capacity. Thus, a margin of safety exists. Furthermore, it is clearly a safety _ improvement to replace a system with known limitations with the alternate main steam piping and condenser treatment pathway, which has been shown to have excellent reliability. The exemption from 10CFR50 Appendix J requirements for MSIV leakage rates is required sn that HCGS can operate with the proposed Technical Specifications limit of 200 scfh and with the alternate MSIV leakage treatment method. This benefit will compensate for any decrease in safety that may result from the granting of this exemption.

Thus, special circumstances exist warranting the granting of the exemption.

F. Environmental Imoact .

The proposed exemption has been analyzed and determined not to cause additional construction or operational activities which may significantly affect the environment, it does not result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Impact Statement-Operating License Stage, result in a significant change in effluents or power levels, or affect any matter not previously reviewed by the

-Nuclear Regulatory Commission which may have a significant adverse environmental impact.

The proposed exemption does nct alter the land use for the plant, any water uses or impacts on water quality, air or ambient air quality. The proposed action does not affect the ecology of the site and vicinity and does not affect the noise emitted by station. Therefore, the proposed exemption does not affect the analysis of environmental impacts described in the environmental report.