ML20247K730
| ML20247K730 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 05/13/1998 |
| From: | Eric Simpson Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20247K735 | List: |
| References | |
| LCR-H98-01, LCR-H98-1, LR-N98219, NUDOCS 9805220193 | |
| Download: ML20247K730 (12) | |
Text
[" ll1 MAY 131998 company E. C. simpson,
PutAc Service Electoc and Gas Company Po. Box 236. Hancocks Bndge, NJ 08038 609 339-1700 knn W e hemannt Noctem Engw LR-N98219 LCR H98-01 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 REQUEST FOR CHANGE TO TECHNICAL SPECIFICATIONS INSERVICE LEAK AND HYDROSTATIC TESTING REQUIREMENTS 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.
On April 18,1994, the NRC issued a Safety Evaluation for Amendment No. 69 to the Hope Creek Generating Station'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 not to exceed 212 F) provided that certain Operational Condition 3 Limiting Conditions l
for Operation (LCO) for secondary containment 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, " Improved 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 overly 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 that this change is also required at Hope Creek since the overly restrictive LCO requirements impose unnecessary challenges to plant operations. The proposed l
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 appmved by the NRC in an SER dated Much 31,1998 for IES Utilities' Duane Arnold Energy Center facility.
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9805220193 990513 PDR ADOCK 05000354
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MAY 131998 Document Control Desk' Lft-N98219
- 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)(1), 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 pagas affected by the proposed changes are provided in
- Attachment 3.
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 j
any questions regarding this request, we will be pleased to discuss them with you.
l Sincerely, i
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 I
l 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 l
1 95 4933
A REF: LR-N98219 LCR H98-01 STATE OF NEW JERSEY )
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COUNTY OF SALEM
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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.
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Subscribed and Sworn o before me this
/6 day of 1998 f
AI Notary PSic of Nedersey EUZABETH J.MDD Nomy pVBUC OF lWEW JERSEY My Commission expires on
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Dscument Control D:sk LR-N98219 LCR H98-01 HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354 ULT 1 MATE HEAT SINK REVISIONS TO THE TECHNICAL SPECIFICATIONS (TS)
BASIS FOR REQUESTED CHANGE:
On April 18, 1994, the NRC issued a Safety Evaluation for Amendment No. 69 to the Hope Crtok Generating Station's Technical Specifications (TS).
The TS amendment provided a new 3pecial Test Exception, 3/4.10.8, which allowed the performance of pressure 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.
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 requirements for secondary containment isolation, secondary containment integrity and filtration, recirculation and ventilation (FRVS) operability be met.
One of these Operational Condition 3 Limiting Conditions for Operation (LCO) requirements included the "High Drywell Pressure" Secondary Containment Isolation trip function (TS Table 3.3.2-1, Trip Function 2.b).
However, Special Test Exception 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 inspections.
With the primary containment not established, the drywell pressure switches can not be considered operable since the finite volume, for which their trip setpoints were selected, does not exist.
Therefore, regardless of the surveillance status of the High Drywell Pressure trip
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function, the containment conditions would most likely prevent i
this function's initiation of a secondary containment isolation signal.
l The above condition results in a conflict between the Hope Creek l
TS definition of OPERABLE and the TS 3.10.8.a requirement for i
operable drywell pressure switches.
The current Hope Creek TS l
requirements impose a requirement that cannot be met without 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 E---
l Document C:ntrol D3ck LR-N98219 s
Ahtachment 1 LCR H98-01 I
c'onfli'ct by removing the requirement for an operable High Drywell Pressure trip function during performance of the hydrostatic test.
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 OPERABLE 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.
BACKGROUND:
The current TS 3/4.10.8 was implemented at Hope Creek through the NRC 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-1433, dated September 28, 1992.
On April 7,
1995, Revision 1 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 Utilitiee' 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 make the Hope Creek TS consistent with the requirements contained in the Duane Arnold TS and NUREG-1433, Revision 1.
JUSTIFICATION OF REQUESTED CHANGES:
The Hope Creek TS define plant Operational Conditions.
Operational Condition 4, COLD SHUTDOWN, requires that the average reactor coolant temperature be less than or equal to 200 F, and if the average reactor coolant temperature exceeds 2 %"r, then Operational Condition 3, HOT SHUTDOWN, must be eumered.
The HC TS Special Test Exception 3.10.8 permits a relaxation of some of the requirements of Page 2 of 4
Dscument'C:ntrol Deck LR-N98219 LCR H98-01 d'perat'ional Condition'3 only for the period during which the required hydrostatic and leak tests are being conducted.
Specifically, the primary containment is allowed to be opened for frequent unobstructed access to perform the required inspections.
The Operational Condition 3 requirements for maintaining 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 P 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 restrictions provided in LCO 3.10.8 require second ry containment integrity, operable FRVS and operable isolation actuation instrumentation for this equipment.
However, the requirement 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 function provides no additional protection against the events of concern during the inservice leak and hydrostatic tests.
Since RCS temperature is limited to 212'F, there would te 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 performance does not allow, primary containment to be established.
Based on the low temperature requirement, and the fact that primary containment will not be set, the "High Drywell Pressure" Secondary Containment Isolation 1
trip function provides no 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 release to the atmosphere.
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e Document Csntrol Dack LR-N98219 LCR H98-01 fn thd event of a large loss-of-coolant ~ accident during a leak:or hydrostatic test, the RCS would rapidly depressurize, thereby permitting the. low pressure ECCS 3
i equipment, required by TS.3.5.2, to actuate and thereby l
keep the core flooded.
This action would prevent the fuel l
from overheating and releasing radioactive materials. The d
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 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.
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CONCLUSIONS:
i 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.
Page 4 of 4
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D2cument Centrol Dock LR-N98219 Attcchment 2 LCR H98-01 1
HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354 REVISIONS TO THE TECHNICAL SPECIFICATIONS (TS) l 10CFR50.92 EVALUATION
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Public Service Electric & Gas (PSE&G) has concluded that the l
proposed changes to the Hope Creek Generating Station (HC)
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 10CFR50.92 is provided below.
i 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.
l BASIS 1.
The proposed changes do not involve a significant increase in the 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 l
Containment Isolation trip function during a leak or I
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 i
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.D cument Centrol.Derk LR-N98219
. Attachment.2 LCR M98-01 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 pro, posed change does not create the possibility of a new-or different kind of accident from any accident previously evalua ted.
The proposed 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 that could initiate an event.
Therefore, these changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.
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-Dscument Centrol Dsck LR-N98219 lkttcchment 2 LCR E98-01 3.
The proposed change does not involve a significant reduction in a margin of safety.
The proposed TS revisions will still 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.
Since the reactor vessel head will remain in place, secondary containment will continue to be maintained, sufficient isolation actuation instrumentation 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 impact on any design basis accident or safety limit.
Since 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, PSE&G has determined that_the proposed changes do not involve a significant hazards consideration.
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D:cument Csntrol Deck LR-N98219 I
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LCR H98-01 HOPE CREEK GENERATING STATION
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FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354 REVISIONS TO THE TECHNICAL SPECIFICATIONS (TS)
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 3.10.8 3/4 10-8 1
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