ML20236F549
| ML20236F549 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 06/18/1998 |
| From: | Tosch K NEW JERSEY, STATE OF |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9807020205 | |
| Download: ML20236F549 (2) | |
Text
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$faic of picfu 3creeg Christine Todd Whitman Department of Environmental Protection Robert C. Shinn, Jr.
Governor Commissioner Division of Environmental Safety, Health,
)
Analytical Programs Radiation Protection Programs Bureau of Nuclear Engineering P.O.
Box 415 Trenton, NJ 08625-0415 Tel: (609) 984-7700 FAX:
(609) 984-7513 June 18, 1998 U.S. Nuclear Regulatory Commission Document Control Desk Washington, D.C.
20555 Ladies & Gentlemen:
SUBJECT:
Hope Creek Generating Station (H.C.)
Facility Operating License: NPF-57, Docket No. 50-354 Request for Change to Technical Specifications (TS)
Inservice Leak and Hydrostatic Testing Requirements LCR H98-01 On May 13, 1998, PSE&G submitted LCR H98-01 to the Nuclear Regulatory Commission (NRC).
The proposed change is seeking to relax restrictions for access to the drywell to conduct leak and hydrostatic
- testing, and to also result in H.C.
TSs with requirements similar to the ones contained in NUREG-1433 " Standard Technical Specifications - General Electric Plants, BWR/4", Rev 1".
)
Amendment # 69 to the H.C.
Operating License provided a l
Special Test Exception, which permits the unit to remain in j
Operational condition 4
(while the average reactor coolant
)
temperature remains above 200 degrees F but not in excess of 212 j
degrees F) but under certain restrictions, such as maintaining
,l Secondary Containment integrity and meeting operability requirements for the Filtration, Recirculation, and Ventilation /
System.
i The proposed amendment is intended to allow implementation of the pressure testing requirements under Cold Shutdown operational condition (at temperatures not exceeding 212 degrees F), while removing the operability requirement for the Secondary Containment Isolation trip function.
9007020205 900618 PDR ADOCK 05000354 PW P
New Jersey as an liqual Opportutury I mployer l
Recyded Paper I
The New Jersey Department of Environmental Protection (NJ DEP)
Bureau of Nuclear Engineering (BNE) has reviewed the change request in accordance with the requirements of 10 CFR 50.91(b) and has no comments.
If you have any questions, please contact Richard Pinney at (609) 984-7558.
Sincerely, L
D Kent Tosch, Manager Bureau of Nuclear Engineering c:
H.
Miller, Administrator - Region I, U.S. NRC-R.
Ennis,. Licensing Project Manager - H.C.
U.S. NRC J.
Lipoti,_Ph.D., Ass't Director, Radiation Protection
.NJ DEP D. Powell, Director Licensing, Regulation, anf Fuels
- PSE&G l
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LR N98219 LCR H98-01 United States Nuclear Regulatory Commission -
Document Control Desk Washington, DC 20555 REQUEST FOR CHANGE TO TECHNICAL SPECIFICATIONS L
_ INSERVICE LEAK AND HYDROSTATIC TESTING REQUIREMENTS L
.' HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354 Gentlemen:
In accordance with 10CFR50.90, Public Service Electric & Gas (PSE&G) Company hereby requests a revision to the Technical Specifications (TS) for the Hope Creek Generating Station (HC). In accordance with 10CFR50.91(b)(1), a copy of this submittal has been sent to the State of New Jersey.
L O_n April 18,1994, the NRC issued a Safety Evaluation for Amendment No. 69 to the Hope Creek Generating Statior 's Technical Specifications (TS). The TS amendment provided a new Special Test Exception, 3/4.10.8, which permits the unit to. remain in Operational Condition 4 with the average reactor coolant temperature above 200*F (but L
not to exceed 212'F),provided that.certain Operational Condition 3 Limiting Conditions L
for Operation (LCO) for secondary con'.ainment isolation, secondary containment integrity _ and filtration, recirculation and ventilation (FRVS) operability are met.
PSE&G's justification for the changes was based, in part, on their consistency with the provisions contained in GE BWR/4, *lmproved Technical Specifications," NUREG-1433, dated September 28,1992.
. As indicated in the attachments to this letter, Revision 1 to NUREG-1433, issued on April 7,1995,' removed overty restrictive LCO requirements for the inservice leak and -
hydrostatic test. _ Specifically, the operability requirement for the "High Drywell Pressure" Secondary Containment Isolation trip function was deleted. PSE&G believes p
that this change is also required at Hope Creek since the overly restrictive LCO requirements impose unnecessary challenges to plant operations. The proposed changes in this submittal will result in Hope Creek TS similar to requirements contained in Revision 1 to NUREG-1433 and similar to a TS amendment approved by the NRC in an SER dated March 31,1998 for IES Utilities' Duane Arnold Energy Center facility.
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NRC review of the changes contained in this submittal.is requested by December 20, 1998 to support the next refueling outage (RFO8) at Hope Creek. The proposed changes have been evaluated in accordance with 10CFR50.91(a)(11 using the criteria in 10CFR50.92(c), and a determination has been made that this request involves no significant hazards considerations. The basis for the requested change is provided in to this letter, A 10CFR50.92 evaluation, with a determination of no significant hazards consideration, is provided in Attachment 2. The marked up Technical Specification pages affected by the proposed changes are provided in.
Upon NRC approval of this proposed change, PSE&G requests that the amendment be
. made effective on the date of issuance, but allow an implementation period of sixty days to provide sufficient time for associated administrative activities. Should you have any questions regarding this request, we will be pleased to discuss them with you.
Sincerely, Affidavit Attachments (3)
C-Mr. H. Miller, Administrator - Region i U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. R. Ennis Licensing Project Manager - Hope Creek (Acting)
U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 14E21 11555 Rockville Pike Rockville, MD 20852 Mr. S. Pindale (X24)
USNRC Senior Resident inspector-HC
. Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering P. O. Box 415 Trenton, NJ 08625
REF: LR-N98219 LCR H98-01 STATE OF NEW JERSEY )
) SS.
COUNTY OF SALEM
)
E. C. Simpson, being duly sworn according to law deposes and says:
I am Senior Vice President - Nuclear Engineering of Public Service Electric and Gas Company, and as such, I find the matters set forth in the above referenced letter, concerning Hope Creek Generating Station, Unit 1, are true to the best of my knowledge, information and belief.
f 8 1 M 1,
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Subscribed and Sworn o before me this
/3 day of
.,1998 f
l Notary PSic of Nedersey' EllZABETH J. 200
Mm PUBUC OF W JERSEY My Commission expires on
.. s.nn y_ g,37
LR-N98219 D cument Centrol Dock LCR 898-01 HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354 ULTIMATE HEAT SINK REVISIONS TO THE TECHNICAL SPECIFICATIONS (TS)
BASIS FOR REQUESTED CHANGE:
On April 18, 1994, the NRC issued a Safet' Evaluation for h 69 to the Hope Creek Generatini Static 5' M chnical
~
A pr s; specifications (TS).
The TS amendment provided a new Special Test Exception, 3/4.10.8, which allowed the performance of a reactor coolant temperature of greater tnan pressur'E Esiting at 200*F but 1_ess,than Requal. to 212'F, while considering the plant to remain in Operational Condition 4.
This Special Test Exception relaxed the Primary Containment Integrity requirements normally associated with reactor coolant temperatures greater than 200*F, thus allowing less restricted access to the Reactor Pressure Vessel (RPV) head area of primary containment for the performance of the required inspections.
The Special Test Exception, however, required that the Operational _ Condition 3,TS isolation, secondary requirements.for se_co.ndar.y containment containment ihte_grity and filtration, recirculation and ventilation (FRVS) o,pe rabi li t y,b_e_ met.
One of these Operational Condition 3 Limiting Conditions for Operation (LCO) requirements included the "Hi_gh Dryw_ ell. Pressure" Secondary Containment Isolation trip function (T5 Table 3.3.2-1, Trip Function 2.b).
However, Special Test ExcepTi~on~3.10.8 does not require, and test performance does not allow, primary containment to be established in order to allow for the.
aforementioned leak inspection's.
With the primary containment not established, the drywell pressure switches can not be considered operable since the finite volume, for which their trip does not exist.
Therefore, regardless
-setpoints were selected, of the surveillance status of the High Drywell Pressure trip function, the containment conditions would most likely prevent this function's initiation of a secondary containment isolation signal.
The above condition results_in a_c_onfligt_be,tyeen the Hope Creek for
_IE_ definition of. OPERABLE and the TS 3.10.8.a requirement operable drywell pressure switches.
The current Hope Creek TS be met without requirements impose a requirement that,cannot defeating the purpose of having access to the drywell during the
' hydrostatic test.
As discussed in the following sections, Revision 1 to NUREG-1433, issued April 7, 1995, addressed this Page 1 of 4
Document Centrol Deck LR-N98219 LCR E98-01 conflict jg/ remqving the requirement for an operable High Drywell
.Ptessure. trip. function ~ddring' performance of the. hydrostatic
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t e s t...
REQUESTED CHANGE AND PURPOSE:
As shown in Attachment 3 of this letter, TS 3/4.10.8, Inservice Leak and Hydrostatic Testing, is being. revised to delete the requirement for an operable High Drywell Pressure trip function.
Specifically, TS 3.10.8.a is being revised to remove the reference to the. Secondary Containment Isolation Actuation instrumentation trip function 2.b.
These changes resolve the current conflict betWeen the Hope Creek TS definition of OPERASLE and the TS 3.10.8.a requirement for operable drywell pressure switches, and enable access to the drywell during performance of the hydrostatic test.
RACKGROUND:
The current TS 3/4.10.8 was implemented at Hope Creek through the NR6 approval of TS Amendment No. 69.and its associated SER dated April 18, 1994.
In a letter, dated March 4, 1994, PSE&G provided justification for these TS based, in part, on their consistency with the provisions contained in GE BWR/4, " Improved Technical Specifications," NUREG 143.3, dated September 28, 1992.
On April 7,
1995, _Reyision_.l to NUREG-1433, was issued, which removed the overly restrictive LCO requirements for the high drywell pressure trip function during the hydrostatic test.
Subsequently, other utilities have adopted the revised NUREG-1433 requirements for the hydrostatic test, including IES Utilities' Duane Arnold Energy Center facility.
The Duane Arnold TS amendment was approved by the NRC in an SER dated March 31,_1998.
The changes
~ ~
proposed in this submittal mak's the Hope Creek TS qonsistent with j
the requirements contained in the Duane Arnold TS and NUREG-1433, l
Revision 1.
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JUSTIFICATION OF REQUESTED CHANGES:
The Hope Creek TS define plant Operational Conditions.
1 Operational Condition 4, COLD SHUTDOWN, requires that the average reactor coolant temperature be less than or equal l
to 200*F, and if the average reactor coolant temperature exceeds 200*F, then Operational Condition 3, HOT SHUTDOWN, must be entered.
The HC TS Special Test Exception 3.10.8 permits a relaxation of some of the requirements of Page 2 of 4 f
- s LR-N98219 D:cument Centrol Dock LCR 898-01 Operational Condition 3 only for the period during which the required hydrostatic and leik Yests are being
~
conducted.
Specifi~c~ ally, the primary containment is
, allowed to be opened for f requent unobstructed Acc.ess lo perform..the_ required inspections.
The Operational Condition 3 requirement's for miTEtaining secondary
~
containment integrity as well as filtration, Recirculation and Ventilation System (FRVS) operability are imposed during the conduct of the-testing.
As stated in the NRC SER for Hope Creek TS Amendment No.
69, permitting the average reactor coolant temperature to be increased above 200*F and limiting the maximum reactor coolant temperature to -212*F while performing leak, or hydrostatic tests will not substantially affect the consequences of potential accidents which might occur with the increased average reactor coolant temperature since these tests are performed with the reactor coolant system (RCS)'near water solid and with all control rods fully inserted.
Therefore, the stored energy in the reactor core would be very low and the potential for causing fuel failures with a subsequent increase in coolant activity is minimal.
The ~ restri_ct. ions provide _d in LCO_3.102 (e_ quip _q syndary cpqtainment Mtegrity, pperable FRVS and pperable isolation agtuation instrumentation for this equipment.
However, the re'qu'i remen't for the "High Drywell Pressure" Secondary
~
Containment Isolation trip function (TS Table 3.3.2-1, Trip Function 2.b) is unnecessarily restrictive since the trip the f
prpvides no additional protection against 7 unct, ion,f concern during the inservice leak and hy~drostatic la events o
~
5jsts.
Since RCS temperature is limited to 212*F, there would be no flashing of coolant to steam.
Even with a large line break, drywell pressure will not likely reach the isolation actuation setpoint.
Furthermore, Special
[ Test Exception 3.10.8 does not require, and test allow, primary containment to be l performance does.notBased on the low temperature requirement, and established.
\\ the fact that primary containment will not be set, the Isolation "High Drywell Pressure" Secondary Containment trip function provides 333. additional protection during a vessel drain down event.
Even without an operable, "High Drywell Pressure" Secondary Containment Isolation trip function, leakage of the radioactive materials from the RCS would be still be filtered by the FRVS prior to releas'e~t'o the atmosphere.
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Document Control D:ck LR-N98219 LCR E98-01 In the event of a large loss-of-coolant accident during a leak or hydrostatic test, the..RCS would rapidly d.,ep re s s u r i z e, thereby permitting the low pressure ECCS 4
equipment, required by TS 3.5.2, to actuate and thereby keep the core flooded.
This action would prevent the fuel Trom~ overheating and releasing radioactive materials. The RCS inspections required to be performed as part of the leak or hydrostatic tests continue to be expected to detect small leaks before a significant inventory of coolant was lost.
Based upon the above, PSE&G concludes that the deletion of the operability requirements for the "High Drywell Pressure" Secondary Containment' Isolation trip function 1
during a leak or hydrostatic test will have no impact on plant safety.
The proposed changes will ensure acceptable consequences of any' postulated accidents, are_ enveloped by the previously_ accepted justification for Hope Creek TS Amendment No. 69, and are, therefore, acceptable.
l CONCLUSIONS:
PSE&G concludes that these proposed changes are adequately justified and result in No..Significant Hazards Consideration as described in' Attachment ~2 of this letter.
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LR-N98219 Decument Control Deck LCR H98-01 HOPE CREEK GENERATING S M ION FACILITY OPERATING LICENSL MF-57 DOCKET NO. 50-354 REVISIONS TO THE TECHNICAL SPECIFICATIONS (TS) 10CFR50.92 EVALUATION Public Service Electric & Gas (PSE&G) has concluded that the (HC) proposed changes to the Hope Creek Generating Station Technical Specifications do not involve a significant hazards consideration.
In support of this determination, an evaluation of each of the three standards set forth in 10CER50.92 is provided below.
REQUESTED CHANGE As shown in Attachment 3 of this letter, TS 3/4.10.8, Inservice Leak and ' Hydrostatic Testing, is being revised to delete the requirement for an operable High Drywell Pressure trip function.
Specifically, TS 3.10.8.a is being revised to remove the reference to the Secondary Containment Isolation Actuation Instrumentation trip function 2.b.
BASIS 1.
The proposed changes do not involve a significant increase in che probability or consequences of an accident previously evaluated.
The proposed TS revisions will continue to allow the performance-of inservice leak and hydrostatic testing at a
reactor coolant temperature of greater than 200*F but less than or equal to 212'F while considering the plant to remain in Operational Condition 4; however, the requirement to have an operable "High Drywell Pressure" Secondary Containment Isolation trip function during a leak or hydrostatic test is being deleted.
This change will not have.an impact on the consequences of an accident previously evaluated since the tests will continue to be performed nearly water solid and with all control rods fully' inserted.
The stored energy in the reactor core and coolant will continue to be very low and the potential for causing fuel failures with a subsequent increase in coolant activity will continue.to be minimal.
The remaining restrictions provided in special Test Exception 3.10.8 requiring Secondary Containment Integrity and Filtration, Recirculation and Ventilation System (FRVS) operability Page 1 of 3
.e D cument Centrol Dack LR-N98219 LCR E98-01 will continue to provide assurance that potential releases into secondary containment will be, restricted from direct release to the environment.
With the reactor coolant continued to be limited to 212*F, there will be little or no flashing of coolant to steam, and any release of radioactive materials will be minimized.
In the event of a large primary system leak, the reactor vessel will rapidly depressurize, allowing the low pressure Emergency Core Cooling Systems (ECCS) to operate.
The capability of the required ECCS in Operational Condition 4 remains adequate to maintain the core flooded under these conditions.
Small system leaks will continue to be detected by leakage inspections, which are an integral part of the inservice leak and hydrostatic testing programs, before any significant inventory loss can occur.
In addition, the "High Drywell Pressure" Secondary Containment Isolation trip function (TS Table 3.3.2-1, Trip Function 2.b) provides no additional protection against the events of concern during the inservice leak and hydrostatic tests.
As a result, these changes will not increase the probability of an accident previously evaluated nor significantly increase the consequences of an accident previously evaluated.
2.
The proposed change does not create the possibility of a new or d1fferent kind of accident from any accident previously evaluated.
The pro (osed changes to Special Test Exception 3.10.8 contained in this submittal will not adversely impact the operation of any safety related component or equipment.
Since the proposed changes involve no hardware changes and no changes to existing structures, systems or components, there can be no impact on the potential occurrence of any accident due to new equipment failure modes.
The remaining restrictions provided in proposed Special Test Exception 3.10.8 requiring Secondary containment Integrity and Filtration, Recirculation and Ventilation System (FRVS) operability will continue to function as required, which will provide assurance that potential releases into secondary containment will be restricted from direct release to the environment.
Furthermore, there is no change in plant testing proposed in this change request I
that could initiate an event.
Therefore, these changes will not create the possibility of a new or different kind i
of accident from any accident previously evaluated.
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LR-N98219 D cument Centrol Dack LCR 898-01 4
3.
The proposed change does not involve a significant reduccion in a margin of safety.
r TheproposedTSrevisionswillstillallowtheperformahce of inservice leak and hydrostatic testing at a reactor coolant temperature of greater than 200*F but less than or equal to 212'T while considering the plant to remain in
-Operational Condition 4; however, the requirement to have an operable "High Drywell Pressure" Secondary Containment Isolation trip function during a leak or hydrostatic test is being deleted.
Since the reactor vessel head will remain in place, secondary containment will continue to be sufficient. isolation actuation instrumentation maintained, will be maintained and all-systems required in Operational Condition 4 will continue to be operable.in accordance with the TS, the proposed changes will not have any significant Since impact on any design basis accident or safety limit.
Hope Creek will still remain capable of meeting all applicable design basis requirements and retaining the capability to mitigate the consequences of accidents described in the UFSAR, the proposed changes contained in this submittal were determined to not result in a significant reduction in a margin of safety.
CONCLUSION Based on the above, PSEGG has determined that the proposed changes do not involve a significant hazards consideration.
Page 3 of 3
_ _ _ _ _ _ _ _ _ _ ~ _ _ _ _ _
i Document Centrol Dock LR-N98219 i
LCR E98-01 HOPE CREEK GENERATING STATION I
FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354 l
REVISIONS TO THE TECHNICAL SPECIFICATIONS (TS) i 1
TECHNICAL SPECIFICATION PAGES WITH PROPOSED CHANGES The following Technical Specifications for Facility Operating License No. NPF-57 are affected by this change request:
Technical Specification Page l
3.10.8 3/4.10-8 I
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SPrefAL TYST EXCEPTIONS 3/4.10.8 IN5ERVICE LEAK AND NYDPOSTATIC TESTING LIMITING CONDITION FOR OPERATION when conducting inservice leak or hydrostatic testing, the average l
3.10.8 reactor coolant temperature specified in Table 1.2 for OPERATIONAL CONDITION 4 may be increased to 212'F, and operation censidered not to be in OPERATIONAL CONDITION 3, to allow performance of an inservice leak or hydrostatic test provided the following OPERATIONAL CONDITION 3 LCO's are mets Functions 2.a, 2.b,
- 3. 3.2, " ISOLATION ACTUATION INSTRUMENTATION",
a.
2.c, 2.d and 2.e of Table 3.3.2-1; b.
3.6.5.1, ' SECONDARY CONTAINMENT INTEGRITY" and
- 52CONDARY CONTAINMENT AUTOKATIC ISOLATION DAMPERS";
c.
3.6.5.2,
" FILTRATION, RECIRCULATION AND VENTILATION SYSTEM."
d.
3.6.5.3, OPERATIONAL CONDITION 4, with average reactor coolant APPLICABILITY:
temperature > 200'F.
ACTION:
immediately With the requirements of the above specification not satisfied, enter the applicable condition of the af fected specification or immediately suspend activities that could increase the average reactor coolant temperature or pressure and reduce the average reactor coolant temperature to :s 200'F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SURVEILLANCE REQUIREMENTS
...........................................................=....=====...===
verify applicable OPERATIONAL CONDITION 3 surveillance for 4.10.8 specifications listed in 3.10.8 are met.
Amendment No. 69 3/4 10 3 HOPE CREEK
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