ML20070U659
| ML20070U659 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 04/03/1991 |
| From: | Public Service Enterprise Group |
| To: | |
| Shared Package | |
| ML20070U644 | List: |
| References | |
| NUDOCS 9104090235 | |
| Download: ML20070U659 (47) | |
Text
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HOPE CRELK GENERATING STATION GCLOSURE 1 hhltLJfr.yjce. Electric and Gu Li((tiSE NO. NPF-51. ,
DOCKET NO. 50 351 i
PROPOSED __C ENGES TO TEC11NICAL SPECIFICATIONS Replace the following pages with the attached revised page(s). These pages are provided in their entirety with marginal marking to indicate the changes.
~1. Page xi
- 2. Page3/4 62
- 3. Page 3/4 63
- 4. Page 3/4 67-
- 5. Page 3/4 6 20 t
- 6. Page B 3/4 6 1 t
9104090235 9104o3 hDR ADOCK 05000354 PDR
.INDEX LIMITING CONDITIONS FOR OPERATICN AND SURVEILLANCE REQUIREMENTS
\
SECTION PAGE 3/4.4.6'- PRESSURE / TEMPERATURE LIMITS Reactor Coolant System................
................... 3/4 4 21 Figure 3.4.6.1 1 Minimum Reactor Pressure Vessel Metal Temperature Versus Reactor Vessel '
Pressure............................... 3/4 4 23 Table 4.4.6.1.3-1 R+ actor Vessel Material Surveillance Program Withdrawal Schedule........... 3/4 4 24 1
Reactor Steam 00me........................................ 3/4 4-25 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES.......................... 3/4 4-26 ,
13/4.4.8__ $TRUCTURAL INTEGRITY............. ......................... 3/% 4 27 3/4.4.9 RESIDUAL HEAT REMOVAL-Hot' Shutdown... 2..........................................
3/4 4 28 Cold Shutdown............................................. ~3/4 4 29 .
3/4.5 EMERGENCY CORE COOLING SYSTEMS j.
.3/4.5.1' ECCS:-
0PERATING...................................... ... 3/4 5 1 3/4. 5. 2 -- ECCS
-LSHUTD0WN............................................ 3/4 5-6 1 3/4.5.3 SUPPRESSION.CHAMBERh..................................... .
3/4 5 8 4 3/4.6 CONTAINMENT SYSTEMS ,
3/4.d.1 PRIMARY CONTAINMENT' P ri ma ry ' C on ta i nme n t I n t e'g r i ty. . . . . . . . . . . . . . . . . . . . . . . . . . . . . -3/4 6-1 Primary Containment l Leakage. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-2
' Prima+ ry Conta inment A i r Loc ks ' .-. . . . . . . . . . . . . . . . . . . . . . . . . . .= 3/4 6-5 e- '
u , !
DELLTE my3p7:me ; + c _ _.
c '; ~,"7 Primary Containment: Structural Integrity................... 3/4 6 8
-Orywell and Suppression Chamber Internal Pres sure. . .'. . . . . .- 3/4 6 HOPE CREEK . xi y *e -
up ' g.p gy qee-r-e-ry p-'..-vy9yera m g: s 4-e-9<ww.-W%ne--.gycy 1_/p-. ,m-qgpw--- . na.+r-pe'is & ..>upm-TW3g H. s.iv,_m4 g-ceg-y,av um-y9Wy.4yyLp.w w ,w4 >- -- mg &y .w---w w:--
3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 PRIMARY CONTAINMENT INTEGRITY PRIMARY CONTAINMENT INTEGRITY cnsures 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,the containment leakage barriers between leakage tests.
^
Operating experience with the main steam line isolation valves has indicated that degradation has occasionally occurred in the leak tightness of the valves; therefore the special requirement for testing these valves.
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 contairement air locks are required to meet the restrictions on PRIMARY CONTAINMENT 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 position during reactor operation. Only one closed door in each air tock is required to maintain the integrity of the containment, ww Nv -sv <ee ,
3/ M O .4 MSIV SEALING SYSTEM l Ca q resu1 Sng from the maximum leaka pH ance for the main steamline isolation va N in the postulated LO MC tions would be a small l fraction of the 10 CFR 100 guTtte s, pr,cx44c5 the main steam line system from DEtITE the isolation valves up to and in tfMhe turbine condenser remains intact.
Operating experience has 3 i dlesT that de'grada Qon has occasionally occurred in the leak tightne W the MSIV's such that the b e Qied leakage requirement have not jalw y een maintained continuously.
th yetren ed leakage from the MSlys when isolation - Theofsealin79 ysQm will the prima bryatem reduc containment is reouired. Uh' and HOPE CREEK B 3/4 6-1
s CONTAINMENT SYSTEMS
- ' PRIMARY CONTAINMENT LEAKAGE-1.lMITING CONDITION FOR OPERATION _
I 3.6.1.2 Primary containment leakage rates shall be limited to:
9 INSERTaj-a. An overa11 integrated leakage rate of less than or equal to La , 0.5 per-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,{g,
- b. A combined leakage rate of less than or equal to 0.60 t for all penetrations and a11 valves listed in Table 3.6.3-1, exfept 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 l hydrostatically tested per Table 3.6.3 1, subject to Type B and C INSERT 2 G np, , _ to ,4
(,
LO th c';ut te " . 5 0 9 c c-M - ' t F m ;'
- c. s*y:::Q : ,A g };;}:'j g 3(:::', r t & W
.d.- A combined leakage raie of less than or equal to 10 gpm for all con
- 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 4nich penetrate the primary containment, when tested at Pa, 48.1 psig Ap.
' APPLICABILITY: When PRIMARY. CONTAINMENT INTEGRITY is required per SpecifMation 3.6.1.1.
ACTION:
With:
- a. The measured overal_1 integrated primary containment leakage rate
-exceeding 10.75 L, or
- b. The' measured ' combined leakage rate ~ forl 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 tern seal-of the feedwater lines, and other valves which are hydrostatically tested.
per Table 3.6.3-1, subject to Type-B a d C tests' exceeding 0.60 L , or INSERT 3 ,, ,myrmd
.c. eyp xceedin a
_ d.1 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 gps, or.
e.- The measured c>mbined leakage rate-for:all other containment-isolation-va'ves in hydrostatically tested lines in Table 3.6.3-1 which penetrate .
i
.the primary containment exceeding 10 spm- :l restore!
- a. 'The overalifintegrated leakage rate (s) to less than or q ual to 0.75'La, and -
- Exemptio,ito-Appendix"J"of10CFR50.
3/4 6-2
. HOPE CREEK l l
.- _ __. _ _ _ _ _ ~ . _ -- . _ . _ _ . _ _ _ .
. CONTAINHE_NT SYSTEMS LIMITING CONDITION FOR OPERATION (Continuedl 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, and other valves which are hydrostatically tested per Table 3.6.3-1, subject to Type B and C tests to less than or equa o 0.60 L,, and
- c. T aka e g Q Q or equal t3*
g % . _. _
0 ee ns;::,. Mhetgh- )
- 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 3.6.3-1 to less than or equal to 10 gps, 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 gpe, prior to increasing reactor coolant system temperature above 200*F.
SURVEILLANCE REQUIREMENTS 1
- 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 Containment 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 Commission. If two consecutive Type A tests fail to meet 0.75 L,, a Type A test shall be performed at least every 18 months until two consecutive Type A tests meet 0.75 L,, at which time the above test schedule may be resu'oed.
- c. The accuracy of each Type A test shall be verified by a supplemental test which:
.1. Confirms the accuracy of the test by verifying that the difference 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 1., and 1.25 L,.
- Exemption to Appendix "J" of 10 CFR 50.
HOPE CREEK ,
3/4 6-3
d CONTAINMENT SYSTEMS
! LIMITING CON 01 TION FOR_ OPERATION
- 3. I- i;
-A CABit-14Vt--ODEftAT40 hat-GOND1440HS-1-e-and ACTIONi -
With -one -- IV-sealing system-subsystem. inoperable,-restore.the i parable sub-system R AB L E-s ta t us-w i t hi n40-day s-o r . be-i n-a t4 e a s t-HOT- TDOWN-within-tw-eent-le h, n :nd-4n COLD SHUT 00WN-w4 thin-the-fel4ew4*g-P -houne GUAVE444- W N- ,ws e u 4r6:-1 OPE AS' 4 r-lbne--
6-
!NSERT $
4
^:
, :.L L , . n
& team-6tep-Valves-(MGGVs)-+ -testab e-dwe4eg-operat4e wtheevgh-e least en; ec rlete-eyele -f 4- trav+1,-
c At least-once-per4 8-- 4ths-by-pe o rma nc e-o f-a-func ti ona l-tes t-o f-the-subsystem-tbrev 44 t+-opoeat equencer -and-Ver4fy.ing-that '.
each-iater40ck-end mar operatu.n gned and.r.ach motomet.ic.
valve-estwate st^ fts c^r-rect-pos4 tion .
- d. -By-veel fyf eg . e-cont-col-4 n&teumentat4on4 be-OPERABLE 4y-
-par.formance f-4+-
4- CH .NEL-CHECK-44-least-once-per-24-houe*r--
- a. == mCmo -im-+14eesi ,m n ; , e,d C HANNEL-cal 4 BRAT 40N-e t-4eas-t-ence-per-16-mont hs .
/
i HOPE CREEK 3/4 6-7.
i l
l 1
s TABLE 3.6.3-1 (Continued) x-a o A PRIMRY CONTAINMENT ISOLATION VALVES MAXIMUM
.k p PENETRATION ISOLATION TIME . t i VALVE FUNCTION AAD NupeER NUMBER (Seconds) NOTE (S) P&ID-
= -_ _-_ _ . ,_
(c) 5 stem Isolation Valves- M-72-1 J, DELETE Outside: 1 d
Line A HV-5834A (KP-V010) 45 1 l ;
Line B .-HV-5835A -
PIB 45 1
- Line (KP-V008)- Plc 1 ;
ine D HV-5837A (KP-V007) PID 45 t L ~-- -- .. !
- 2. Group.2'- Reactor Recirculation Water Sample System 4
w (a).' Reactor Recirculation Water Sample'Line Isolation Valves M-43-1 !
.1 .
i
. Inside: . 88-SV-4310 P17 15 3 ;
a Outside: 88-SV-4311~ P17 15 3 3.. Group 3 - Residual Heat Removal (RHR) System I i' s (a) RHR Suppression Pool Cooling Water & System Test i Isolation Valves M-51-1 t
. Outside:
) Loop A: HV-F024A (BC-V124) P2128 180 4
- HV-F010A (BC-V125) P2128 180 4 i
1 Outside:
Loop B: MV-F0248 (BC-V028) P212A 180 4 l j HV-F0108 (BC-V027) P212A 180 4 l
(b) RHR to Suppression Chamber Spray Header Isolation Valves M-51-1 Outside:
Loop A: HV-F027A (BC-V112) P2148 75 3 j Loop B: .HV-F027B (BC-VOIS) P214A 75 3 l
! j
Reference:
ICR 90-02 AMfklS._fDR_IIIMLCAlt Jf_K[11CMLON - 3 4,L2 lEEEk.L1 (except f or innin steam isolation valves')
IM11tL2 Less than or equal to 200 seth per main steam line when tested at 25 poig.
lEfifeT 3 200 ocf h f or any individual inain steam line, or INSERLi 11,5 seth on the affected main steam line, and TM13.T FOR TECilNICALSlQlFj{ CATION - 3,6,1,4 INE9T 5 Ti(IS PAGE INTENTIONALLY BLANK Coction 3.6.1.4 is reserved for future use and to preserve subsequent Section 1,umbering I
HOPE CREEK GENERATING STATION MCl0SURE l hblis_Etr.ylce Electrlt_and Gas , UCENSE NO. NPF El DOCKET NO. 50 354 SUPPORTING INFORMATION AND ANALYSES
1.0 INTRODUCTION
AND
SUMMARY
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 Scaling 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 sumnary of the plant. specific radiological dose assessment, and Section S.0 summarizes the potential benefits for a lechnical Specification MSIV allowable leak rate of 200
, -scfh, and the deletion of the MSIV Sealing System. The BWROG report, NEDC-31858P, "BWROG Report for increasing MSIV Leakage
- Rate Limits and Elimination of Leakage Control Systems", February 1991, provides the justification for,. increasing MSIV leakage limits, and for L eliminating the-requirements for the MSIV Scaling System.- With concurrence from the valve manufacturers, this report concludes that MS!V leakage rates up to 200 scfh
- are not'an indication of substantial mechanical' defects in -the valve which would challenge thi isolation capability of the valve to fulfill its safety function. Ther9 fore, the-proposed increase in the allowable leakage rate to 200 scfh for the MSIVs will not inhibit the isolation capability of the valve.
1
HOPE CREEK GENERA 11NG STA110N 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 MSlVs. The main steam drain lines are employed to convey leakage to the condenser. PSE&G proposes to delete the HSly Scaling System from the Technical Specification and to incorporate the isolated condenser as an alternate method for MSly leakage treatment. Tne BWROG has evaluated the availability of main steam system piping and condenser alternate treatment pathways for processing MSIV leakage. We have reviewed the potential combinations of loss Of-Coolant Accidents and seimic events of interest: (1) LQSS-0f COOLANT-ACCIDENT (lQMLW11110#T NfM_(QlE1QLtiL10ELC [y[lil. 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) MISMIC EVEPT WITJ0RT NEAR,1QJEIDIM.102 Without a LOCA and the potential associated core degradation, the radioactivity transported with HSly leakage is of no radiological significance, p (3) LOSS Of-00QlANT-ACCIDENT ILQCA) W11H NEAR COINCIDINT _ S11111LC 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-allow MSly leakage to bypass the alternate troatment 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 aproximately 0-001 . per L reactor per year; core melt probability is plant specific- and typically ranges from 0.00001 to 0.0001 per reactor per year). 2
l HOPE CREEK CENERATING $1A110N i lt 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 perreactorperyear), s Considering that the probability of a near coincident LOCA and scismic event is much smaller than other plant safety risks (less than 1 x 10 7p er reactor por year for coincident events, less than 5 x 10 7 per reactor per year for seismic induced LOCA), the concern for main steam piping and/or condenser damage is of little significance. Nevertheless, because main steam piping and condenser systems designs are extremely j 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. It is noted I that ANSI-B31,1 design requirements have been extensively used for ! nuclear power plant system design and that this code contains a good deal of margin, in addition, specific seismic design provisions have been incorporated into newer BWR main steam and condenser systems such as Hope Creek Generating Station. In order to further justify the capability of the main steam system 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). 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. 3
l ll0PE CREEK GENERATING STATION The earthquake experience data are derived from an extensive database on the performance of power plants and industrial f acilities, complied by EQE for the Seismic Qualification Utility Group, the Electric Power Resetrch Institute, and many other EQE clients. Tht: 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 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 equipment such as taotor control centers, pumps, valves, structures, and so forth. Tha+ is, with limited exceptions, normal industrial construction and .pment 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 carthquakes 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 tv or more rugged than those that exhibited good earthquake performance. We conclude 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 unpre;edented.
- 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 demonstrates that a maximum MSIV leakage rate of 200 scfh per main steam line results in an acceptable increase to the dose exposures calculated for a design basis LOCA. In addition, the 4
1 HOPE CREEK GENERATING STATION l l J 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 BACdiROUND - The safety function of the MSIVs is to isolate the reactor system in the event of a loss of Coolant Accident (LOCA) or other events 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 t'ne Updated Final Safety Analysis Report (VFSAR). The allowable leakage rate , from MSIVs is included in the LOCA radiological analysis evaluated in Section 15.6.5.5 of the UFSAR. The safety-related MSIV Scaling 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 MS!Vs 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 isolation valves which provide for containment integrity in the event that the corresponding MSIV fails to close. Operating experience indicates that MSIVs frequently exceed the Technical Specification allowable leak rates. Some of these valves repeatedly faii the local leak 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, 'MSly leakage and LCS Failure" as a high ! priority item in January 1983. This issue was closed in 1990. 5 i
HOPE CREEK GENERATING STATION The BWROG formed a MSIV Leakage Committee in 1982 to address the increasing M!lY leakage rates, and a follow-on MSIV Leakage Closure Committee in 1986 to address alternate actions to resolve on going, but less severe MSIV problems. The MSIV 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. 1 3.0 10STtil. CATIONS FOR THE PRQP0_SID CHANG 1 PSE&G proposes to increase the Technical Specification allowable leakage rate for the MSIVs f rom 46.0 scfh to 200 scfh per main steam line and to delete the MSIV Sealing System requirements from the Technical z 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 mair.tenance 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 the allowable HSIV leakage rate. This calculation is described in Section 15.6.5.5 in the VFSAR. 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 Issus C 8, the LCS will not function if MS1V exceeds the leakage limit by a moderate amount. 6 1
HOPE CREEK GENERATING STA110N MSIV's failure to meet the current Technical Specification 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" MSly local leak rate tests were reported in the early NRC survey to exceed the leakage rate limit. The BWROG has studied the issues regarding MSIV icakage rates, their causes, and available alternatives. The results of the BWROG study are l provided in NE00 31858P and are summarized in NUREG ll69. In response to Generic issue C 8, the BWROG has recommended corrective actions and maintenance practices to reduce the MSIV leakage rates. A recent survey conducted by the BWROG of MSiv icakage tests performed between 1984 and 1988 indicates that the implementation of industry and BWROG actions has been effective in reducing the leakage rates, and, in particular, a reduction in the number of valves which experienced substantial high leakage rates. The survey concludes that about 23% of the total "as found" MSIV. leakages still exceeded the limit of 11.5 tefh and about 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, MSIV leakage rates still frequently exceed the current Technical Specification limit and the safety and maintenance problems related to high MSIV leakage rates, although less severe, remain 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 repeating failure. By 7
H0pE CREEK GENERATING STATION not having to disassemble the valves and refurbish thui for minor leakage, the utility may avoid introducing one of the root causes of recurring valve leakage problems that led to later leak test f ailures and 1 the posssility of compromising 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, f ast 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 MS!V leakages, the BWROG has concluded the MSIV leakage rates up to 500 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. Furthermore, valve manuf acturers 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 MSlV's performance of the isolation function and will not compromise the safety of HCGS. This proposed increase provides a more realistic, but still conservative, 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 11.6 scfh. The increase in successful local leak rate testing will significantly reduct 'tly maintenance cost, reduce dose exposure to maintenance personnel, i . 2 outage durations, extend effective service life of the MSIVs, and minimize the potential for outage extensions at HCGS. A safety-related LCS was required by 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 MS!V leakage rates, as determined by conservative local leak rate tests, were 8 l
HOPE CREEK GENERATING S1A110N 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 (MSly Scaling System at HCGS) is designed for MSly 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 MSiv Sealing System requirements from the Technical Specifications and proposes the isolated condenser as an alternate method for MSIV leakage treatment. Since .mpler and less equipment is employed, the alternate method is more reliable than tne MSIV Sealing System. As supported by the BWROG, this prnposed change will resolve the safety concern associated with LCS performance capability at high MSIV leakage rates, and will assure that a reliable and effcctive resolution to Generic issue C-8 is maintained. Furthermore, PSE&G will incorporate the applicable alternate leakage treatment methods, consistent with GE document NE00 30324 " Potential Operator Act;ons to control MSIV Leakage", into the Operational Procedures and Emergency Operational Procedures. In addition to resolving the safety concern, the propose deletion of the MSIV Scaling System requirements in the Technical Specifications will result in significant operational and maintenance benefits. The BWROG has avaluated recen, rformance data. Results of this evaluation is shown " NEDC 31858P. The evaluation indicates that leaktge control s3 stems are difficult to maintain. Plant shutdowns and start-up delays have occurred. The MSIV Sealing System performance at HCGS is consistent with the recent survey by the BWR00. In conclusion, PSE&G proposes to increase the Technical Specification allowable MSIV leakage rate from 46,0 scfh total to 200 scfh per main I g 1
HOPE CREEK GENERATING STATION steam line, and to eliminate the requirements for HSiV Sealing System in the Techn' 31 Specifications. The proposed increase in the MSIV leakage limit should significantly reouce recurring valve leakage problems, and will minimize needless valve repair which can compromise plant safety. The proposed deletion of the safety related MSly Sealing System and the proposed alternate method (main steam lines and condenser) will resolve the safety concern egarding LCS effectiveness at higher MSIV leakage rate. 4.n 'NAJXJilS Of MSIV LLAKAGE CONTRJfj)llpN TO RADIOLOGICAL 00SERLQLAJ10Ei 4.1 Egl gtlon_of Alternate lejlag_e Treatment MthM The BWroG hn evaluated several alternate MSly leakage treatment methods and has recorunded the isolat(d condenser for MSiv leakage treatment, Trts leakage treatmen method d.akes advantage of the large volume in the isolated main condenser to hold up the release of fission products leaking from the closed MStVs. The main steam drain lines are employed to convey leakage to the condenser. As previously discussed in Section 1.0, the BWR00 has evaluated the availability of main steam system piping and condenser alternate treatment pathways for processing MSIV leakage. We have 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 B31.1 design requirements typically used for nuclear plant system design contain a good deal of margin. In order to further justi' the capability of the main steam piping and condenser altornate treatment ) pathway, we have reviewed limited earthquake experience data on the performance of non-seismically designed piping and condensors (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 1
HOPE CREEK GENERATING STATION condensers in the event of a design basis earthquake is highly unlikely, and 'that such ' 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 Plants", dated May 1987, which reported no observed failures in the main steam piping over 313 years of reactor anerating 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 fladioloolsal Analys1 Land Rgtylt The radiological dose methodology developed by General Electric for the BWROG is documented in Appendix C of NEDC-31858P. The radiological analysis calculates the effects of the proposed allowable MSIV leak cate in terms of control room and offsite doses. The revised LOCA doses are the sum of the LOCA doses (as described in Section 15.6.5.5 of the UFSAR) and. the calculated MSIV 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 ; ulated _ dose exposures from the BWROG radiological analysis for HCG. : egulatory 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 ecceptable 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 Criteria 19) for
HOPE CREEK GENERATING STATION-the control room doses. Furthermore, the calculation shows-that MSIV
=
leakage rates 1 up to approximately 500 scfh per steam line would not exceed the regulatory limits. 5,0--RENEFITS FOR THE PROPOSED CHANGES As discut, sed in NEDC 31858P, recent MSIV leakage performance has
- significantly_. improved. since the early BWROG survey in 1984 and the NRC survey in the_early 1980s. Despite the recent improvement, MSly leakage
-rates- exceeding -the current Technical Specification 11. nits 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 f 77% success-- rate at- the. current Technical Specification limit of 46.0-scf h .- _ ,
- . Specifically, MSIV leakage -experience at Hope Creek Generating Station b
has improved.due to the- implementation of BWROG recommended changes to
- hardware and' maintenance activities, However, even with-the changes, MSIV leakage -frequently exceeds the current limits and this excessive
-loakage hasi.recently resulted in_ outage extensions and unnecessary
~
radiation exposures for refurbishment of the valves, Deletinghtho'MSIV _ Sealing System will reduce the overal1 dose ? rates, and eliminats the isystemts_ impact on refueling and maintenance outage activities:at HCGS.=
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l HOPE CREEK GENERATING STATION Table 1 CONlRIBUTION TO THE LOCA DOSE EXPOSURES FOR A MAXIMUM MSIV LEAK RATE OF 200 scfh HOPE CREEK GENERATING STATinN Whole Body Thyroid Beta IIM1 .._( rf m) iram). 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
- 0) New Calculated Doses 0.7 73.3 Low Population A) , 'CFR 100 Limit 25 300
- Zone (30-0ay)
B) Previous Calculated 0.08 7.7 Doses ** C) Contribution from 0.39 87.8 MSIV at 200 ccfh D) New Calculated Doses 0.47 95.5 Control Room A) GDC-19 5 30 30/15*** (30-Day) B) Previous Calculated 0.04 0.26 .91 Doses ** C) Contribution From 0.17 4.50 2.08 MS!Vs at 200 scfh D) New Calculated Doses 0.21 4.76 3.79 No limit specified. UFSAR Section 15.6.5.5 and 6.4 (includes MSIV. 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.
HOPE CREEK GENERATING STATION
- 4 Enclosure 3 f.ublic Service Electric and Gas ?
LICENSE NO. NPF-57 DOCKET NO. SQ-J11
~
NO SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS b:. 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 feetiper hour _ (scfh) to 200 scfh -per main steam line..
3
;- 2 ) > . Delete- Sections 3/4.6.1.4 and B3/4,6.1.4 to reflect the deletion- '
of the-MSIV Sealing System from the Technical Specifications, d
;3)) ReviseL index- pages xi and Lxii, Section 3.6.1.2(c) and Sections - !
3.6.1.4, 4.6.1.4-and B3/4.4.6.1.4.
~
l< _ Nm 1
~
- PursuantStoL10CFR50.12(a)LPSE&G has appliedcfor an exemption request. ,This
-;applicafion>provides a: detailed.Justifidation Lfor exempting the downstream ti
-; mainfsteam t piping:andicondenser from tha' seismic requirements specified in .
LAppendixl ALof 10CFR100,iand demonstrates that the proposed exemption will not
- presentLan - ' undue risk.to the public health'and safety.
< l
, a iPSE&Gihas,rpursuantitol10CFR Part.50.92, reviewed-the proposed amendment toi !
i determine ;whetheriou'r. request- Involves ~a significant hazards considerations.s m
, JThe operathn of Hape Crey Generatina- Stalionu.10 accordance with thet py_qposed' amendment; willHnot' involve a sianificant increase in the orobability 7 or 'conseauences of an accident previousiv evalgte.d. '
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-HOPECREEKGENERATINGSTNTION 1
1 I The prcposed amendment to Section 3.6.1.2(c) does not involve a change to structures, components, or. systenis that would affect the probability of an -; accident- previously evaluated in the Updated Final Safety Analysis Report (UFSAR).- This proposed change involves increasing the allowable MSIV leakage rate from 46.0 scfh total to 200 scfh per main steam line, The proposed amendment to delete Sections 3/4.6.1.4 and Bases Sections B . 3/4.6.1.4 involves eliminating the MSIV Scaling System requirements from the
-_ Technical - Speci fication s . 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 af ter an accident has occurred, this proposed amendment has no effect on the probability of an accident.
Since MSIV leakages and operation of the MSIV. Sealing System are included in the radiological analysis.for the design-basis Loss-of-Coolant Accident (LOCA) as: described ;in: Section: 15.6.5.5 of the UFSAR, the proposed amendments will q not-' affect the precursors of other analyzed accidents. The proposed Jamendments result -in acceptable radiological consequences of the design basis LOCA previously evalu'ated in Section 15,6.5.5 of the UFSAR. The. Hope Creek Generating . Station has an inherent MSIV leakage treatment capability.- PSE&G proposes to.use the main steam lines and condenser as an
' alternate'to Regulatory Cuide 1._96 " Design of Main Steam isolation' Valve
'_ Leakage Control _ System f_or--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-1 evaluated the availability of main' steam system piping and condenser alternate- r
.treatmerit pathways for processing MSIV leakage. We have determined that:the
. probability ofia'near coincident LOCA and a seismic event is much smaller than for--.other plant safety 1 risks. The-BWROG- has also determined that main steam-
- piping and con' denser designs: are cxtremely- 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.
2-
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HOPE CREEK GENERATING STATION steam piping and condenser alternate treatment pathway,.we have reviewed slimited earthquake experience data on the performance of non seismically designed piping and condensers (in post earthquakes). This study :oncluded 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 eartnquake is highly _ unlikely, ar,i that such a failure would also be contrary to a large body of historical- etrthquake experience data, and thus unprecedented. 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 control 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, this-equipment is expected'to remain intact following a design basis earthquake.
-The-radiological analysis uses standard conservative assumptions for the release of! source terms consistent with Regulatory Guide (RG) 1.3= Revision 2, " Assumptions Used for Evaluating the Potential Radiological Consequences of a LLoss-Of-Coolant Accident for Boiling Water Reactors", dated April 15/4.
The analysis demonstrates that dose contributions from the proposed MSIV leakage l 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-.controinroom doses as contained in-10CFR100 and 10CFR50, Appendix A E(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 doses are the-sum of the LOCA doses previously evaluated in the UFSAR and the Ladditional MSIV_~ doses calculated using the alternate treatment method. Th i_ s method. of _ calec1ating the -revised doses: is very conservative since the LOCA doses' previous evaluated already included dose contributions from MSIV at the. maximum; leakage Lrate-permittef in the current Technical- Specifications. The attached table- shows the- previous calculatei doses and the new -calculated
' doses.,
3- 4 l
l HOPE CREEK GENERATING STATION CONTRIBUTION TO THE LOCA DOSE EXPOSURES FOR A MAXIMUM MSIV LEAK RATE OF 200 scfh HOPE CREEK GENERATING STATION Whole Body Thyroid Beta (rem) (rem) Irgal 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 7.7 Doses ** C) Contribution from 0.39 87.8 MSIV at 200 scfh D) New Calculated Doses 0.47 95.5 Control Room A) GDC-19 5 30 30/75*** (30-Day). B) Previous Calculated 0.04 0.26 .91 Doses ** C) Contribution from 0.17 4.50 2.88 MSIVs at 200 scfh D) New Calculated Doses 0.21 4.76 3.79
- No limit specified.
UFSAR Section 15.6.5.5 and 6.4 (includes MSIV 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 het been made to use protective clothing. 1
HOPE! CREEK GENERATING STATION l i
-The whole body dose at the Low Population Zone _ (LPZ) and the control room is !
Lincreased from 0.08 to 0.47 rem and from 0.04 to 0.21 rem respectively. These -; increases are -acceptable because the revised doses are well within the-regul'atory guidelines :(0.47 versus 25 rem for the offsite and 0.21 versus 5 rem;for the control room). The associated whole body dose at the exclusion area boundary (EAB) increased insignificantly from 0.6 to 0.7 rom. Theithyroid-dose at the LPZ increased from 7.7 to 95.5 rem. This increase is Jacceptable because the revised dose )f 95.5 rem is significantly less than of the_ regulatory' guideline (300 rem). The EAB _ thyroid dose increesed slightly.- from.76.7 to .79.3 rem,- whereas the control room thyroid dose increased from. , 0.26--to 4.76 rem. The increase in control room thyroid dose is acceptable-ibecause the revised dose remains a small fraction (16%) of the limit. The control' room: beta dose :is increased from 0,91- to 3.79 rem, which remains insignificant relative to-the regulatory guideline of 30 rem. It~ is importantLto note-that the resulting doses are dominated by the organic - > 11odine fractions which occur'because of the ultraconservative source term
.assumptionsusedinthisanalyiis, for 200 scfh per steam line, more than 80%. y
-of the 'off-site iodine and control room doses are -due to organic iodine from l the RG 1.3 source term and -organic: iodine converted' from the elemental iodine depositedLin. main steam piping systems. If the actual iodine composition from N Ltheifuel releaseh(cesium iodide) is used -in the calculations, essentially-all
.:of this organic' iodine dose would be eliminated. ;
Infsummary s the ' proposed changes result in an acceptable increase to the: , radiological consequences of a.LOCA proviously evaluated in the'UFSAR.' The revised LOCA dosesiare well'within the regulatory guidelines, q 1 > LThei proposed ; amendment to Table 3.6.,3-1 of the Technical _ Specifications'. involves (the deleti'on of'MSIV1 Scaling System valvesafrom the list of-containment; isolation valves.= This proposed change is consistent with the:
~
proposed ~ deletion of the MSIV Sealing System.- The-inboard system lines +hich j are connected to;the main stsam piping will be disabled -with blind flanges to _ l assure tha't containment integrity is' maintained. 5
~
HOPE CREEK GENERATING STATION j l This proposed change does not involve an increase in the probability of an I 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 i containment integrity, isolation, and leak test capability are not compromised for the postulated 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 Hope Creek Generating Station._in accordance with the ptoposed amendmant._will nqt create the possibility qLa new or different kind af accident from any accident _ oreviously evaluated. < 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 evalusted.- The BWR Owners' Group evaluated MSIV leakage performance and concluded that MSIV 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 which could impact the MSIV operability The 1.0CA 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 jroposed amendment to delete Sections 3/4.6.1.4 and Bases Sections B 3/4.6.1.4 =does not create the possibility of a new or different kind of accident from any accident previously evaluated because the removal of the HOPE CREEK GENERATING STATION 4
. MSly Scaling System does not affect any of the remaining _ systems at- HCGS and the LOCA has been analyzed using the alternate method to process MSly leakages.
-The proposed amendment to delete the MSIV Sealing System isolation valves from Table 3.6.3'-l does not create- the possibility of a new or different kind of j accident. The inboard MSIV Scaling 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 add a footnote to Section 3.6.1.2a 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 in4 nature, and do not create the possibility of a new or 3 different kind of accident from any_ accident previously evaluated.
-The ooeration of Hope Creek Generatina ' Station. in accordance wit'h the
'oronosed amendment. will not involve a sinnificant reduction in the maroin of safety.
The- proposediamendment to Section 3.6.1.2' does not involve a significant-
- reductioniin the margin of safety- - As discussed in the Bases of the Technical-1 Specification 3/4.6.1.2, the- allowable leak rate limit specified for. the~ MSIVs ismused to quantify a' maximumfamount of bypass-leakage assumed in the LOCA- -
l'
-i ra_diol_ogicali analysis Results ofitho. analysis 'are? evaluated = against-the dose-guidelines contained in 10CFR100 for the offsite dosesLand 10CFR50,; Appendix-
'A(General; Design: Criteria 119) for the contro11 room doses. The marginLof safety is considered ~to be the~ difference between the > calculated doses and the j [guidelinestas: contained.in.10CFR100andGDC19.
\
l
. _ . _ - - - - _ _ - . _ - - - _ _ . _ _ = .
HOPE CREEK GENERATING STATION Results of the radiological analysis demonstrate that the proposed (hange does not involve a significant edut tion in the margin of safety. The margin of safety with respect to whole body doses, are insignificantly reduced by 1.6% at the LPZ, 3.4% in the control room, and 0.4% at the EAB. The thyroid dose margin of safety is reduced by 29.2% at the LPZ,14.1% in the control room, and 0.9% at the EAB. The beta dose margin of safety is insignificant 1y reduced by 3.0% in the control room. The margins of safety are not significantly adversely affected because the absolute margins of safety remain well below the guidelines (lowest whole body margin of safety is 95.8% in control room, lowest thyroid margin of safety is 68.2% at LPZ). , Therefore, the propcsed amendment does not involve 2 significant reduction in the margin of safety at_HCGS. The proposed amer.dment to delete Sections 3.6.1.4, 3/4.6.1.4, and B3/4.6.1.4 does not involve a significant reduction in the margin of safety. The intended function of the MSIV Sealing System for treatment of MSIV leakage will be performed by using the alternate path via the main steam lines and condenser. The radiological effects on the margin of safety are discussed above for Section 3.6.1.2. In addition, the MSIV Scaling System will not function at MSIV leakage rates greater than 100 scfh, thus the deletion of the Sealing System, and using the alternate path will eliminate the safety concerns of Generic issue C-8. The safety significance of the LCS in terms of public risk was addressed in NUREG CR-4330. This NUREG evaluated the possibility of regulatory modification concerning eliminatlon of the leakage control system requirements and disabling the systems currently installed at BWRs. The NUREG conclud2d that the increased public risk is less than one percant. Therefore, the proposed amendment does not involve a significant reduction in the margin of safety at HCGS. The proposed amendment to delete MSIV Sealing System valves from Table 3.6.3-1 dces not involve a significant reduction in the margin of safety. Isolation 8 L- J
HOPE CREEK GENERATING STATION 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 reduction in the margin of safety. The proposed change to add a footnote to Section 3.6.1.2a 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. 9
i
, HOPE CREEK GENERATING STATION ENCLOSURE 4 ~
Public Service-Electric and Gas Litf,NSE NO. NPF-57 D_0_CKET NO 50 354 , APPLICATION FOR EXEMPTION TO APPENDIX A 0F 10CFR100 Pursuant .to Section 50.12(a)- of the Regulations of the Nuclear Regulatory
- Commission, Public Service Electric and Gas (PSE&G), holder of f acility
_0perating: License No. NPF-57,'hereby requests an-exemption of the downstream main steam piping.and: condenser from the seismic requirements specified -in j
- Appendix A of :10CFR100.
Iniconjunction with this application for exemption request, PSE&G has
; transmitted to the Nuclear Regulatory Commission an application _ for a license amendment pursuant _to.10CFR 50.90. 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 (scfh) total- to- 200 scfh per main steam-line, and v
a proposed change to.Section.3.6.1.4 for eliminating the requirements for the , MSIV Sealing System. Ne safety analysis:has been revised to-assess the radiological-- effects -of MSIV l leakages followingla postulated design basis ;
* ~LOCA.
PSE&G has" demonstrated .that? the proposed change does;not involve a ; significant hazards consideration. _ a , a
~
- T_his proposed exemption is- a. result of the _ extensive work performed by the BWR
'0wners' Group _-(BWR0Gj;insupport'oftheresolutionofGenericIssueC-8."MSIV N' -leakage.and LCS Failure".=
4 x
< LThe fo_llowing discussion provides;a -detailea justification and evaluation of o
Jthe proposed exemption. - The proposed exemption ?is: found to be ' authorized by-law, willi not present an.-undue risk to the publicLhealth and safety, and.is -, consistent-with the-common defense and_-security. Furthermore,<special circumstances are present that warrant the granting of_this exemption. 1
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; HOPE CREEK-GENERATING STATION The' proposed exemption will not -cause additional operational activities that may significantly affect the environment. It does not result in a significant 2 increase in any adverse environmental _ impact previously evaluated in the final Environmental Impact Statement 0perating License Stage, a significant change
'.in effluents or power levels, or effect any matter not previously reviewed by-the Nuclear Regulatory Commission that may have a significant adverse environmental impact.
1 Therefore, pursuant- to -10CFR 50.12(a), PSE&G hereby requests an exemption for
. Hope-Creek Generating Station for the downstream main _ steam piping and
- condenser from 10CFR100 Appendix A requirements.
A? lustification 4
-The BWROGthas thoroughly; studied the availability of a main steam system
= piping and condenser alternate treatment pathway for processing MSIV' leakage.
.We have. reviewed the potential combinations of loss-of coolant- accidents and~
l seismic events of interest:- (1)- LOSS-0F-COOLANT-ACCIDENT-(LOCA) WITHOUT NEAR COINCIDENT SEISMIC EVENT. For_ this occurrence the pressure in the_ piping system a
. -downstream of-the MSIVs is' rapidly reduced _to stmospheric-
- pressure, and.since there is' no seismic event the alternate flow pathithrough ,ain' 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.HSIVsleakage is:of:no1 radiological! significance. - i
- (3)_ LOS_S -OF-C00L ANT- ACC IDENT - ( LOC A) WITH NEAR COINCIDENT SEISMIC 3 EHE. .Forithis. occurrence (alsoLassuming significanticore' ^
damage) the consequences are of' interest because a-seismic induced
-failurelin the-. main steam or condenser system could allow MSIV-leakage to bypass' the alternate _ treatment pathway. .It.has-been previously well documented that the probability of .a -near 2
n , ij-
i I HOPE CREEK GENERATING STATION coincident LOCA and seismic event-is extremely small (design basis , earthqu'ake' probability approximately 0.001 per reactor per year;
-core me_lt probability is plant specific and typically ranges from 0.00001 to 0.0001 per reactor per year), 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 l' 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 and/or condenser j
-damage is' of_ little significance. Nevertheless,' because main steam piping and ,
" condenser systems designs are extremely rugged, this equipment is expected to'-
- resain; intact'followi_ng-design basis seismic-events. The evolution of design-codes and regulatory requirements is documented in Appendix 0 of NEDC 31858P_.
Itsis = notedithatL B31.1 design requirements have been extensively used for nuclear l power plant- system design and that this code contains a good' deal of margin)- In : addition, specific-seismic design-nrovisions-have been
. incorporated into :some newer BWR main steam and condenser : systems.
lIn'orderitofurtherjustifythecapabilityof'themainsteamsystempipingand; condenser alternate treatment pathway, we. have reviewed limited earthquake-experience 1 data on the performance of- non-seismically; designed piping and -) icondensers: (-i_n- past earthquakes) . The' study summarizes data on the performance ofimaint steam piping and-condensers in past strong' motion! -
; earthquakes 7 and; compares _ these _ piping and condensers with those inLtypical U'.S.fGE Mark :1, ill, and _III' 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-LOCAl occurring lJust prior to.or after the seismic event.
3
HOPE CREEK GENERATING STATION The earthquake experience data are derived frcm an extensive database on the performance of power plants and industrial facilities, compiled by EQE for the Seismic Qualification Utility' Group, the Electric Power Research Institute, and many other EQE 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 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 i 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 exc tons, normal industrial construction and equipment typically have substantial inherent seismic ruggedness, even when they are not aesigned 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. We conclude that (1) the possibility of significant failure in GE BWR main steam piping or condensors in the event of an eastern U.S. design basis earthquake is highly unlikely and that (2) any such failura would also be contrary to a large body of historical earthquake ex.oerience data, and thus unprecedented. B. Authorized By law If the criteria established in 10CFR50.-12(a) are satisfied, as they are in this case, and if no other-prohibition of the law exists to preclude the activities which would be authorized by the requested exemption, and there is no such prohibition, then the Commission is authorized by law to grant this exemption request, (See: U.S. vs Alleshenv-Ludlum Stes] Coro2, 406 U.S. 742, 755 (1972)). l 4 l i
HOPE CREEK GENERATING STATION i C. No Undue Risk to Public Health and Safety The treatment method for HSIV leakages is recommended by the BWROG in support of the resolution to Generic Issue C-8 "HSIV Leakage and LCS failure" The BWROG has addressed the capability of the non seismic main steam piping and condenser to withstand a design basis earthquake event. The BWROG study first concludes that the probability of a near coincident LOCA and seismic event is infinitely small. Then the study shows 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, snd that any such failure would also be contrary to a large body of historical earthquake experience, and thus unprecedented. Therefore, the proposed exemption presents no undue risk to public health and safety. D. Consistent with Common Defense and Security With regard to the " common defense and security" standard, the grant 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 Liohtina Compry (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 Circumsignces 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): 5
HOPE CREEK GENERATING STATION (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 since it has been evaluated that the downstream piping and condenser will remain functional during and after a safe shutdown earthquake. This approach is consistent with the current resolution of the seismic and equipment qualification issues. Earthquake experiences data have applied in seismic equipment qualification issues associated with Vnresolved Safety issues A 46 (Seismic Qualification of Equipment in Operating Plants). Piping performance data have been presented in NUREG 1061, a 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. -Compliance with Appendix A of 10CFR100 requires that equipment used to mitigate releases to the public be designed to withstand seismic events. The proposed increase in the MSIV allowable leak rate will not be possible if strict compliance to Appendix A of 10CFR100 is required. This will result in excessive cost to needlessly repair the MSIVs for the remaining life of the plant. 6 l
, . .. -. - - - - _ . - . - - - - - - . _ . -- . - . ~ .
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HOPE CREEK GENERATING STATION s
.(iv) JThe exemption-would resultiin benefit to the public health and safety that compensates for any decrease in safety that may result from the grant of the exemption.
PSE&G has transmitted'to the Nuclear-Regulatory Commission an application for a license _ amendment which involves a proposed change to Section 3.6.1.2 of-the , Technical- Specifications to increase the allowable MSIVs leak rate from 46,0 scfh total to1200 scfh per steam line. 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 t' urn result in higher leakages. The proposed limit will benefit the public
-healthfand_ safety by reducing the potential for HSivs f ailures, and thus
. keeping the' MSIVs leakages within the radiological analysis values. The ,
exemption from Appendix A requirements for the downstream main steam piping ; and condenser is-required so that Hope Creek Generating Station can operate _
.withithe proposed Technical' Specifications value of 200- scfh per steam line.
This Lbenefit 'will compensate for any decrease in safety that may result from the grant'of.the exemption. Thus, special~ circumstances exist warranting the grant of the exemption. :W
~
f.- Environmental Irroact The4 proposed exemption-has been analyzed and< determined not-to cause
-addit.ional constructica:or operational activities which may significantly j affect the environment'1 .It does not: result in a significant zincreasn in any . adverse environmental , impact previously evaluated .in the Final Environmental
- Impact: Statement-0perating License Stage, a significant change in effluents or power; le'vels', or 'a matter not previously- reviewed by tho' Nuclear Regulatory Commission:which may' have a significant adverse environmental impact.
1 The proposed' exemption _does not alterithe land use forLthe plant, any water uses or Limpacts? on' water quality, air or ambient air quality. _The proposed 7
- m. , _. _ __ _ _ . _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ .
HOPE CREEK GENERATING STATION l 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 af fect thc- analysis of environmental impacts described in the environmental report. 8
f HOPE CREEK GENERATING STATION ENCLOSURE 5
. _ _PUBLIC_ SERVICE ELECTRIC AND CAS ~
LICENSE N0. NPF-57
.-DOCKET NO. 50-354 APPLICATION FOR EXEMPTION T0 APPENDIX J OF 10CFR50 Pursuant .to Section 50.23(a) of the-Regulations of the Nuclear Regulatory Commission, .PSE&G,Jholder of Facility 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". l Specifically,; PSE&G requests .that leakages from the main steam isolation valves :(MSIVs)1be exempted from ;th.e acceptance criteria for the -overall
-integrated leak-rate test- (Type A), .as defined in the regulations of -10CFR50, LAppendix J, Pa'ragraphs.Ill.A.5(b)(1) and Ill.A 5(b)(2). ;
iThe purpose of.the test acceptance-criteria is to-ensure that the: measured leak rate from the containment volume will not exceed:the dnsigned containment-Jeak rate assumed in1the: safety analysis for_ a postulated: design basis
. Loss-Of-Cool ant . Accident 1(LOCA) .
<ln: con 3 unction'_withthis.applicat.ionfor"exemptionrequest,tPSE&Ghas transmitted to the Nuclear Regulatory Commission"an application for a license amendment : pursuant to :100FR 50.90. - This license. amendment involves a proposed fchange to:Section' 3;6.1.2. of the :T2chnical . Specifications to' permit. an
~
increase in the- allowable leak rate' for the MSIVrWrom the- current' 46.0 total - [
~
standard cubic: feet Lper hour. (scfh); toe'00 4 scfh per main steam..line, and a i W propo' sed Jchange -to~ Section 3.6.1.4 for elidinating .the requirements for the - MSIV Scaling System. 1The safety ; analysis has_-been : revised toLassess the radiological effects ~ of MSIV leakages. followingsa postulated design basis - LOCA. PSE&G=has. demonstrated.that_the-proposed change does-not involve'a
- significant-hazards consideration.
L 1 L; .._ . .. . . - _ _ .. - ..-. _ ._ a..-.-.. . . . _ . . _ _ . - . _ _ _ - . _ . _ . _ , _ . . . _ . _ . . . . . . . _ . - , _ , _ . . _ _
- - - -.._ _ _ 4 .
HOPE CREEK-GENERATING STATION- s This proposed exemption is a result of the extensive work performed by the BWR
'0wnerst Group (BWROG) in support of the resolution of Generic Issue C-8 "MSIV-Leakage andLLCS. 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. LThe 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 Environmentals impact Statement-0perating License Stage, result i_n a significant chango 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 exet tion for-
- HCGS.for MSIV le'akages-from the'. acceptance test criteria spectfled in~ Appendix J ofL10CFR50.1 A;- Justification--
LTheiregul'ation .of?10CFR50, Appendix J, Paragraphs' III . A.5(b)(1) requires the-overal1; integrated l leakage rate,'as measured during containment.-pressure tests '
-(Type- A)n toi meet).the' acceptance criterion"of less - than or ~ equal to: 0.75 of:
othe maximum allowable containment-leak rate.
'AsJdescribed.inothe: Base s _ Sect _ t ons B! 3/4.6.'l .'2 of: the Technical
,~ -
JSpe61fications, the; limitations on primary containment leakage-rates ensure- ,
# that-theltotal containment leakage volume will not exceed .the value assumed in _ _
.the'. accident analyses; at the peak e cident' pressure. As an a'dded 1
2 l i _a ,, .. ~ . a . . , .= - , , - . - -. -
. -i
n ,
]
HOPE CREEK GENERATING STATION ! Econservatism,Lthe measured leak rate is further limited to less than or equal J 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 i 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 offsite doses, which were evaluated against the dose [ guidelines:of 10CFR50,: Appendix A (General Design Criteria 19) and 10CFR100, L respectively.- Leakages -from the~ containment volus were contained in ~ the :
. - reactor building (secondary containment), filtered by _ the Standby Gas Treatment System,- and then; released to the environment. The maximum containment-= leakage rate includes' leakages-through structures, - all
= penetrations. identified;as Type B, and- all isolation valves -identified -as Type: 4
~C, 1
'The safety _ analysis has;been revised to account for .the radiological effect from MSIV71eakages and from those of =other containment leakages followingia. ,
; postulated 0dcsign: basis LOCA.: Unlike the; treatment path for;other containment
-: leakages,. thest'eatment -'of MSIV11eakages employs the main steam drain _-piping oand_ theJcondenser. fissioniproducts are removed by! plate-out and hold up in ethe1relativel&large ; volumes- of the' main steam piping. and; condenser. -
]
fTheitreatment method 'for' MSIV leakages is -recommended by the BWROG in support
< :ofMtheJresolution to Generic Issue C-8. The BWROG has evaluated the-avai. lability- of main ~ steam system piping .and. condenser alternate treatment pathways for processing MSIV leakage. We have determined that the prob' ability-d
-ofz a near coincident.LOCA= and a seismic event is much smaller than for other
. plant safety: risks. Tlie BWROG has also determined that main' steam piping and-condenser desigr.s are-extremely rugged, and that the B31.1. design requi.rements u
- typically .used iforin_uclear plant system design .contain a good deal of margin'.
3 l l c . :- . -.2..-_. , . , ,. .,
. . .; . -; N
I HOPE CREEK GENERATING STATION l 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 data, and thus unprecedented. Leakages from the HSIVs are not part of the Type A overall integrated containment leak rate and should not be included in the Type A acceptance criterion because the treatment path for MSIV leakages is different from that of containment leakages. Potential leakages from the containment are contained in the reactor building (secondary containment), treated by the SGTS, and released to the environment. MSIV leakages are crentained, platea-out, and delayed -in the main steam piping and the conJenser, and released via the turbine building. The deletion of the LCS is proposed partly in response to the safuy concern identified by Generic Issue C 8 that the LCS would not function at high MSIV leakage rates since the process capability of the LCS at HCGS is designed for MSIV leakage rate of no more than 100 scfh. MSIV leakages are 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 Trately from the maximum containment leak rate. MSIV leakages will be t .s part of the local leak rate test in accordance with the requirements cion 3.4.6.1.2 of the Technical Specifications. This test ensures that easured MSIV leak rate will not exceed the maximum allowable leak rate
. .aed in the' safety analysis.
4
HOPE CREEK GENERATING STATION There is sufficient conservatism in the maximum allowable MSly 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 MSIV leakage rates up to approximately 500 scfh per main steam line will not result in dose exposures in excess of the regulatory limits, 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 exemption for MSIV leakages from the Type A limits is consistert 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 exemption has been granted to the following plants: Fermi 2, Hatch 1 & 2,, Limerick 1, Shoreham, t.aSalle 1 and 2, Hanford, Clinton, Grand Gulf 1, Perry, Dresden 2 and 3, Menticello, Quad Cities 1 and 2,_ Brunswick 1 and 2 and Nine Mile Point 2. Therefore, the proposed >!xemption 11, authorized by law. C. [Lo Undue Risk to Public Health ansLSafstt_y The proposed exemption presents no undue risk to public health and safety. The revised MSly leakage rate has been inc:,rporated in the radiological 5
HOPE CREEK GENERAllNG STATION analysis for a postulated-LOCA as an addition to the designed containment leak rate. The analysis demonstrates an acceptable increase to the dose exposures previously-' calculated for; the-control room and offsite, i.'ie revised LOCA doses remain-well- within the guidelines of 10CFR100 for offsite doses and 10CFR50, Appendix A, (General Design Criteria 19) for the control room Joses, in addition, Section 3.6.1.2 of. the Technical Specification has provided for allowable MSly leak rates, which assure that the MSIVs isolation functinn is not compromised, Finally, potential MSIV leakages are subjected to plate-out, and~ hold up in _the main steam piping and condenser, thus minimizing their ef fect on the total dose released. As discussed in Section F of this application,_.the proposed change will not adversely affect the conclus%s of the previously issued FES OL. Therefore, the proposed exemption presents no undue riskLto public health and safety. 0,- Consistent with Common Defense and Security
'With regard to the " common defense and security" standard, the grant 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 -Liahtina como.tn_y n (Shoreham Nuclear Power Station, Unit
- l), LBP-84-45,- 20 NRC.1343, 1400 (October 29,-1984). There, the-term " common defense and security" refers principal-ly _to the safeguarding of special '
nuclear material,- the- absence of foreign control _ over the applicant, the
~
protection of Restricted Data,<ano the availability of special nuclear material for defense needs. The grantirq of- the requested exemption will not _ affect any of these matters. and,- thus, such g' rants are_ consistent with' tile > common' defense"and. security. E. Soecial Circumstances Are Present Special; circumstances are present which warrant issuance of this requested E exemption. These special circumstances are discussed in accordance with the classificationcontainedin10CFR50.12(a)(2): i 6
t HOPf CRffK GENlRAllNG $1A110N (ib Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to schieve the underlying purpv e of the rule. Compliance with Apper.div J of 10Cf R50 for Type A tests is not necessary to achieve the underlying parpose of the rule since HSIV leakages are not directed into the reactor primary contaituent. Instead, tht M51V's leakage is directed through the main steam drain pining into the condenser. Since Type A tests are intended to measure the primary containment overall integrated leak rat 9 (IlhT), the MSIV's leakage rate should not be included in the measurement of the '.LRT. (iii) Compliance would result in undue hardship or other costs that are significantly in excess of thoce 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 HSly allowable leak rate will not be possible if the HSIV leak rate results are included in the Type A test acceptance criteria. 3 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 MSIVs disassembly and refurbishin0, which is required to wet the low 1:akage limits contributes to repeated failures. Examples of those maintenance induced defects include machining in fuced 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, HCG 1 avoids 7 l
HOPE Cet((K CENERAllNG STA110N 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) Th9 exemption would result in benefit to the public health and safety that compensates for any decrease in safety that may result from the grant of the exemption. PSE&G has transmitted to the Nuclear Regulatory Commission an application for a license amendment which involves a proposed change to Section 3.6.1.2 of the Technical Specifications to increase the allwable MSIVs leak rate from 46,0 total scfh to 200 scfh per main steam line. This application is partly based on tb f act that the current limit is too restrictive, and results in excessive HSIV maintenance and repair, Icading to additional MSIV f ailures, which in turn result in higher leakages. The proposed limit will benef:t the public health and safety by reducing the potential for MSIVs failures, and thus keeping the MSIVs leakages within the radiological analysis values.. The exemption from Appendix 1 requirements for MSIVs leakage rates is required so ( that HCGS can operate with the proposed Technical Specifications value of 200 scfh, This benefit will compensate for any decrease in safety that may result from the grant of-the exemption. Thus, special circumstances exist warranting the grant of the exemption. F. Environmental ILnpm1 The proposed exemption has been analyzed and determined not to cause additional construction or operat ,ial activities which may significantly 8
I r ; H0pE CRllK GCfGRA11NG STATION l l affect the environment. It does not result in a significant iacrease in any adverse environmental impact previously evaluated in the final Environmental impact Statement Operating License Stage, result in a significant change in 1 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 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 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 i environmental report. , i i f 9 L 4}}