ML050590085
| ML050590085 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 02/24/2005 |
| From: | Swenson C AmerGen Energy Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 2130-05-20051 | |
| Download: ML050590085 (13) | |
Text
10 CFR 50.90 2130-05-2005 1 February 24,2005 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington. DC 20555-0001
SUBJECT:
Oyster Creek Generating Station Facility Operating License No. DPR-16 Docket No. 50-219 Technical Specification Change Request No. 319 - Revision to Table 3.1.1 Notes aa and bb Regarding Reactor Building Closed Cooling Water Pump and Service Water Pump Trip Conditions.
Pursuant to 10 CFR 50.90, AmerGen Energy Company, LLC (AmerGen) hereby requests changes to the Technical Specifications (TS) included in Oyster Creek Operating License No.
DPR-16. These changes clarify the conditions under which the Reactor Building Closed Cooling Water (RBCCW) pumps and the Service Water (SW) pumps will trip during a Loss of Coolant Accident (LOCA). The existing notes aa and bb in Table 3.1.1. Protective Instrumentation Requirements, state that the RBCCW pumps and SW pumps will trip during a LOCA. The proposed changes revise the notes to clarify that the RBCCW and SW pumps will trip during a LOCA only if offsite power is unavailable.
Additionally, an editorial change is being proposed to TS page 3.6-2 regarding the incorrect location of a footnote. The affected footnote was inadvertently relocated under TS Amendment 166, approved on 12/13/93. to this letter provides the evaluation of the proposed changes and the no significant hazards consideration determination. Enclosure 2 provides the existing TS pages marked-up to show the proposed changes. The TS Bases pages are provided for information only and do not require NRC approval.
AmerGen requests approval of the proposed amendment by February 24,2006. Once approved.
the amendment will be implemented within 60 days.
AmerGen has concluded that the proposed changes present no significant hazards consideration under the standards set forth in 10 CFR 50.92.
10 CFR 50.90 2130-05-2005 1 February 24,2005 Page 2 The proposed changes to the Technical Specifications have undergone a review in accordance with Section 6.5 of the Oyster Creek Technical Specifications. No new regulatory commitments are included in this submittal.
We are notifying the State of New Jersey of this application for changes to the Technical Specifications by transmitting a copy of this letter and its attachments to the designated State Official.
If any additional information is needed, please contact Dave Robillard at (610) 765-5952.
I declare under penalty of perjury that the foregoing is true and correct.
Sincerely,
\\
Site Vice President Oyster Creek Generating Station AmerGen Energy Company, LLC
Enclosures:
(1) Oyster Creek Technical Specification Change Request No. 3 19, Evaluation of (2) Oyster Creek Technical Specification Change Request No. 319, Markup of Proposed Changes Proposed Technical Specification Page Changes cc:
S.J. Collins, Administrator, USNRC Region 1 P.S. Tam, USNRC Senior Project Manager, Oyster Creek R. Summers, USNRC Senior Resident Inspector, Oyster Creek K. Tosch, Director, Bureau of Nuclear Engineering, New Jersey Department of File No. 03044 Environmental Protection
Oyster Creek Technical Specification Change Request No. 31 9 Evaluation of Proposed Changes
2130-05-20051 Page 1 of5 1.O INTRODUCTION AmerGen Energy Company, LLC (AmerGen) is requesting a change to Technical Specifications (TS), Appendix A of the facility operating license No. DPR-16, for Oyster Creek Generating Station (OCGS). This proposed change revises Oyster Creek Technical Specifications Table 3.1.1, Protective Instrumentation Requirements, notes aa and bb to provide clarification as to when the Service Water (SW) pumps and the Reactor Building Closed Cooling Water (RBCCW) pumps trip during a Loss of Coolant Accident (LOCA).
These pumps will trip on a LOCA when offsite power is unavailable. There are no plant configuration changes associated with this proposed change.
Additionally, TS pages 3.6-1 and 3.6-2 are revised to correct the location of the footnote that applies to Specification 3.6.A.4.1. This footnote was inadvertently relocated under TS Amendment 166. This change is purely editorial in nature and has no impact on the requirements stated in Specification 3.6.A.4.1.
2.0 DESCRIPTION
OF PROPOSED AMENDMENT The proposed amendment will revise TS Table 3.1.1, note aa, which currently states that Service Water pump circuit breakers will be tripped in 10 seconds 515% during a LOCA by relays SK7A and SK8A, and TS Table 3.1.1, note bb, which states that Reactor Building Closed Cooling Water pump circuit breakers will trip instantaneously during a LOCA. The proposed change adds the phrase with a concurrent Loss of Offsite Power (LOOP) to both notes.
Also, the footnote on existing TS page 3.6-1 is relocated to its correct location on TS page 3.6-2.
3.0 BACKGROUND
The primary functions of the Service Water system are to provide normal cooling for the Reactor Building Closed Cooling Water heat exchangers and alternate cooling for the Turbine Building Closed Cooling Water heat exchangers. Additionally, the S W system maintains the Emergency Service Water side of the Containment Spray heat exchangers full of water.
The primary function of the Reactor Building Closed Cooling Water system is to provide cooling water to various components in the Reactor Building and the Drywell. These components include the Spent Fuel Pool Cooling heat exchangers, Reactor Water Clean-up system heat exchangers, heating and ventilation coolers, drain tank coolers, and Reactor Recirculation pump and motor coolers.
Amendment 42 of the Oyster Creek Licensing Application concluded that the Service Water System and Reactor Building Closed Cooling Water System are not safety related,
2 130-05 -2005 1 Page 2 of 5 and are not required for the safe shutdown of the reactor or to mitigate the consequences of postulated accidents.
The Oyster Creek Updated Final Safety Analysis Report (UFSAR) Table 8.3-1 defines the Emergency Busses Automatic Loading Schedule during Loss of Offsite Power (LOOP),
LOOP and LOCA, and LOOP and LOCA with single failure of Emergency Diesel Generator conditions. The UFSAR states that the SW and RBCCW pumps are automatically loaded onto the emergency busses only on LOOP. These loads are not loaded onto the emergency busses when LOOP and LOCA occur concurrently. These loads will be manually restarted by plant operators as the need arises, and emergency bus loading permits. On a LOCA when offsite power is available, these loads remain connected to their normal power supplies, because sufficient capacity is available on the busses.
4.0 REGULATORY REQUIREMENTS & GUIDANCE Amendment 42 to the Oyster Creek Licensing Application concluded that the SW and RBCCW pumps are not required during the accident conditions. Therefore, tripping of these pumps during LOOP with LOCA is acceptable.
Amendment 60 to the Oyster Creek TS authorized changes in the Diesel Generator Load sequence timers for the SW and RBCCW pumps. This amendment added notes aa and bb to TS Table 3.1.1 describing the time delay conditions under which the SW and RBCCW pumps will trip during a LOCA. Notes aa and bb added in Amendment 60 stated that these pumps trip during a LOCA. However, the controls for the SW and RBCCW pumps do not trip these motors on a LOCA when offsite power is available.
The footnote on TS page 3.6-1, applicable to Specification 3.6.A.4.1, was inadvertently relocated under TS Amendment 166. TS pages 3.6-1 and 3.6-2 are revised to correct the location of this footnote. This change is purely editorial in nature and has no impact on the requirements stated in Specification 3.6.A.4.1.
5.0 TECHNICAL ANALYSIS
The proposed change provides clarification to Table 3.1.1, notes aa and bb for when the SW and RBCCW pumps trip whenever a LOCA condition exists. These pumps will only trip on a LOCA when offsite power is unavailable. This clarification does not require any plant configuration change.
The SW and RBCCW systems are not required for the safe shutdown of the reactor or to mitigate the consequences of postulated accidents. Following a complete loss of station normal and auxiliary power, the Circulating Water system, the Steam, Condensate and Feedwater systems, and the Reactor Recirculation system are shutdown. Thus, very little heat load remains on the RBCCW and Turbine Building Closed Cooling Water systems.
One SW pump, operated on emergency diesel generator power, can provide cooling to both of these closed cooling water systems until power is restored. Loss of function of these systems has no adverse consequences on plant releases of radioactive effluents.
2130-05-2005 1 Page 3 of 5 The proposed changes do not affect existing controls. Following a LOCA with offsite power available, these pumps will help in removing plant heat loads. Consequently, the proposed Technical Specification changes will not adversely affect nuclear safety or safe plant operations.
TS pages 3.6-1 and 3.6-2 are revised to correct the location of the existing footnote on TS page 3.6-1, applicable to Specification 3.6.A.4.1. This change is purely editorial in nature and has no impact on the requirements stated in Specification 3.6.A.4.1.
6.0 REGULATORY ANALYSIS
The technical analysis, provided in section 5, and the regulatory requirements, provided in section 4, to add clarification to the TS Table 3.1.1 and to describe the plant configuration as to when the SW and RBCCW pumps trip meets the criteria of 10 CFR 50.36 for inclusion as a condition for operation.
The footnote being relocated from TS page 3.6-1 to page 3.6-2 is applicable to Specification 3.6.A.4.1, which addresses requirements related to sampling of reactor coolant to verify that the reactor coolant specific activity is within specified limits. The Limiting Condition for Operation associated with reactor coolant specific activity meets the criteria of 10 CFR 50.36 for inclusion as a condition for operation.
7.0 NO SIGNIFICANT HAZARDS CONSIDERATION AmerGen has determined that this License Amendment Request poses no significant hazards considerations as defined by 10 CFR 50.92.
- 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed revision to Technical Specification (TS) Table 3.1.1 to clarify the tripping of the Service Water (SW) and Reactor Building Closed Cooling Water (RBCCW) pumps documents the as-built controls for these loads. Amendment No. 42 to the Oyster Creek Licensing Application concluded that these pumps are not required to perform any functions related to safe plant shutdown. During a loss of coolant accident (LOCA) condition, with offsite power available, the plant electrical busses have enough capacity and capability to supply the SW and RBCCW pumps. This proposed change is an administrative change only, and is being made to align the Oyster Creek Technical Specifications with the design of the plant. No physical changes are being made to the plant. Also, the footnote on TS page 3.6-1 wouId be relocated to TS page 3.6-2 to appear on the same TS page as the Specification to which it applies. The proposed changes do not alter the physical design or operational procedures associated with any plant structure, system, or component.
2130-05-2OO5 1 Page 4 of 5 Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed revision to Technical Specification Table 3.1.1 to clarify the tripping of the SW and RBCCW pumps documents as-built controls for these loads. These pumps provide cooling to various non-safety related plant equipment. Following a LOCA condition, with offsite power available, these pumps will help in removing plant heat loads. This clarification that the SW and RBCCW pumps do not trip during a LOCA, with offsite power available, does not affect the Emergency Diesel Generator time delayed loading sequence. The relocation of the footnote applicable to Specification 3.6.A.4.1 is editorial in nature and has no impact on any accident previously evaluated. Accordingly, the proposed changes do not introduce any new accident initiators, nor do they reduce or adversely affect the capabilities of any plant structure or system in the performance of their safety function.
Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
The proposed revision to Technical Specification Table 3.1.1 to clarify the tripping of the SW and RBCCW pumps documents as-built controls for these loads. The NRC Safety Evaluation Report (SER) for Amendment 42 to the Oyster Creek Licensing Application concluded that it is acceptable to automatically trip the SW and RBCCW pumps during a loss of coolant accident. The NRC SER for Technical Specification Amendment 60 concluded that the immediate tripping of the RBCCW pump and the time delayed tripping of the SW pumps during a LOCA was also acceptable. The clarification that the SW and RBCCW pumps do not trip during a loss of coolant accident when offsite power is available does not reduce any margin of safety because these pumps are not required to mitigate the consequences of any postulated accident. The relocation of the footnote applicable to Specification 3.6.A.4.1 is editorial in nature and has no impact on any accident margin of safety.
Therefore, the proposed changes do not involve a significant reduction in a margin of safety.
Based on the above, AmerGen concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92, and accordingly, a finding of "no significant hazards consideration is justified.
2130-05-20051 Page 5 of 5
8.0 ENVIRONMENTAL CONSIDERATION
A review has determined that the proposed amendment would change a requirement with respect to the installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement.
However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibiIity criterion for categorical exclusion set forth in 10 CFR 51.22(~)(9).
Therefore, pursuant to 10 CFR 51.22(c), no environmental impact statement or environmental assessment need be prepared in connection with the proposed license amendment.
9.0 PRECEDENT The proposed changes to provide clarification to Technical Specification Table 3.1.1 for the conditions under which the RBCCW and SW pumps will trip on a LOCA and to relocate a footnote to its correct location are unique to OCGS.
10.0 REFERENCES
- 1) Oyster Creek Nuclear Generating Station Licensing Application, Amendment No. 42
- 2) Oyster Creek Nuclear Generating Station Technical Specification, Amendment No. 60
- 3) Oyster Creek Nuclear Generating Station Technical Specification, Amendment No. 166
Oyster Creek Technical Specification Change Request No. 31 9 Proposed Technical Specification Changes (Mark-up)
The pages included in this enclosure are:
3.1 -7 3.1 -1 8 3.6.1 3.6.2
Specification 3.1.B. 1 dei!:ies the ~ ? m r i ? ? ~ ! ~ +iLI:nbei ~i APRb1 char;nei lrpsis require6 to pzrmit acciiraia average core p o w. i-iG-iiXXng. Specifications 2.;,B,2 and 3.1.C.
I furthe: deiice tha distribution o! :ns OPERABLE chanibsrs io pro\\,,:de rnoniro:ing of local power manges yhai migh: be caused by a single cod withdraviai. Any nearby, OPERABLE LPRM cnainber can Drovide the rzqtiired input for average core moniioring. A Travelling incore Pro& or Probes can be used rempcraiiiy lo provide APKA4 input(s) until LPRM replacement is possibia.
Since APRM rod block protection is nor require0 below 61 % cf rated power, as discussed in Section 2.3. Limitina Saietv Svslem Settinus. operation may continue below 61% as long as Specification 3.1.B.1 and :he requirements c.:
Table 3.:. I are met. In order to maintain reliability of core moniioring in that quadrant where an APRM is inoperable. ii is permitted to remove the OPERABLE APRM from service for caiibration andior test provided that the same core protection is maintained by slternats means.
!n the rare event that Traveliing In-core Probes (TIPS) are used to meet tne requiremenis 3.1.B or 3.1.C, the licensee may periorrr, an anaiysis of subsiitijle iPRM inputs to the APRM system using spare ;non-APRi\\/l inpuij LPRM detectors and change rne APRM system as permitted by 10 CFR 50.59.
Under assumed ioss-of-coolant accidenl conditions and certain loss of otfsite pciwer conditions with no assumed loss-of-coolant accident, it is inadvisable 10 allow the simultaneous starting of emergency core cooling and heavy load auxiliary systems in order to minimize the voltage drop across the emergency buses and to protect against a potential diesel generator overload. The diesei generator load sequence time delay relays provide this protective function ai id are set accordingly. The repetitive accuracy rating of the timer mechanism as well as parametric analyses to evaluate The maximum acceptable iolerances for the diesel loading sequence timers were considered in the estabiishment of the appropriate load sequencing.
Fvlariua! actuation can De accomplished ~9 the operator and is consiaered apcropriate oniy when the aLjiomatlc load seqJencing has beeri completed This will prevent simultaneous starling OT heavy load auxiliary systenis anu proiect ~igainst the poterriai for diesel generator overlcad Also the Reactor Biiiiding Closed Coolilg Water anc Servlcc Water Dump circuit breakers will :rip yvnciever a loss-ot-coolant accident condition exis-s This is jilstified k~
Amendment 42 of the Licensing 4pplication which derermirnci thai these dumps were not required durtrg this acciaerii condltron
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With t h e reactor mode switch i n Run o r S t a r t u p position, with:
- 1.
Thermal power changed by more than 15% of r a t e d thermal p o w e r i n one hour*, o r
- 2.
The off-gas l e v e l, a t t h e SJAE, increased by more than 10,000 microcuries p e r second i n one hour during steady state operation a t release rates l e e s than 75,000 microcuries per second, o r
- 3.
The off-gas l e v e l, a t t h e SJAE, increased by more than 159 i n one hour during steady state operation a t release rates g r e a t e r than 75,000 microcuries per
- second, t a k e sample and analyze a t least one sample, between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following t h e change i n thermal power or off-gas l e v e l and a t least once per four hours t h e r e a f t e r, u n t i l t h e specific a c t i v i t y of t h e primary coolant is restored t o within limits.
3.6.8 Liauid Radwaste Treatment - RELOCATED TO THE ODCH 3.6. C Radioactive Liauid Storaae Applicability:
- 1.
The q u a n t i t y of radioactive material, excluding tritium, noble gases, and radionuclides having h a l f - l i v e s s h o r t e r than t h r e e days, contained i n any of t h e following outdoor tanks s h a l l not exceed 10.0 c u r i e s :
- a.
Waste Surge Tank, HP-T-3
- b.
Condensate Storage Tank I n t h e event t h e quantity of r a d i o a c t i v e material i n any of t h e tanks named exceeds 10.0 c u r i e s, begin treatment as soon as reasonably achievable, continue it u n t i l t h e total q u a n t i t y of radioactive material i n t h e tank is 10 c u r i e s or less, and deecrfbe t h e reason f o r exceeding t h e l i m i t i n t h e next Annurl Effluent Release Report.
Applies a t a l l times t o s p e c i f i e d outdoor tanks used t o s t o r e r a d i o a c t i v e l i q u i d s.
- 2.
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- 3.
S p e c i f i c a t i o n s 3.0.A and 3.0.8 do not apply.
3.6.D Condenser Offaas Treatment - RELOCATED TO THE ODCM 3.6.E Main Con denser Offaas Radioactivity
- 1.
The g r o s s r a d i o a c t i v i t y i n noble gaseo discharged from t h e main condenser air elector s h a l l not exceed 0.21/E C i / S e C a f t e r t h e holdup l i n e where E is t h e average gamma energy (MeV per atomic transformation).
- 2.
In t h e event Specification 3.6.E.1 is exceeded, reduce t h e discharge rate below t h e l i m i t within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> o r be Ln a t least SHUTDOWN CONDITION within t h e following 1 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
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