LIC-12-0136, Responses to Request for Additional Information License Amendment Request to Establish the Reactor Protective System Actuation Circuits Limiting Condition for Operation
| ML12276A043 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 10/01/2012 |
| From: | Prospero M Omaha Public Power District |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LIC-12-0136, TAC ME8038 | |
| Download: ML12276A043 (7) | |
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iJpjju Omaha Public Power District 111 444 South 16 Street Mall Omaha, NE 68102-2247 October 1, 2012 LlC-12-0136 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 References : 1. Docket No. 50-285
- 2. Letter from OPPD (D. J. Bannister) to NRC (Document Control Desk), Fort Calhoun Station (FCS) License Amendment Request (LAR) 12-01, Proposed Change to Establish the Reactor Protective System (RPS) Actuation Circuits Limiting Condition for Operation (LCD), dated February 10, 2012 (LlC-12-0006)
- 3. Letter from the NRC (L. E. Wilkins) to OPPD (David J. Bannister), Fort Calhoun Station, Unit No.1 - Request for Additional Information Regarding License Amendment Request to Establish the Reactor Protective System Actuation Circuits Limiting Condition for Operation (TAC ME8038), dated August 31, 2012 (NRC-12 0084) (ML12236A243)
SUBJECT:
Responses to Request for Additional Information Re: License Amendment Request for Fort Calhoun Station to Establish the Reactor Protective System Actuation Circuits Limiting Condition for Operation This letter provides the Omaha Public Power District's (OPPD's) responses to the Nuclear Regulatory Commission's (NRC's) requests for additional information (,RAls) transmitted in Reference 3. The RAI responses are provided in the enclosure to this letter.
In Reference 2, OPPD requested an amendment to Renewed Facility Operating License No. DPR-40 for Fort Calhoun Station (FCS), Unit No.1 to establish the l.imiting condition for operation requirements for the reactor protective system actuation circuits ,in Technical Specification (TS) 2.15, "Instrumentation and Control Systems." The NRC staff reviewed the information provided in OPPD's application and determined that additional information was required in order to compl'ete its review.
If you should have any questions, please contact Mr. Bill R. Hansher, Supervisor-Nuclear Licensing, at 402-533-6894.
I declare under penalty of perjury that the forgoing is true and correct. Executed on October 1, 2012.
Michael J. Prospero Plant Manager
Enclosure:
Responses to Request for Additional Information Emplovment with Equal Opportunity
LIC-12-0136 Enclosure Page 1 of 6 RESPONSES TO REQUEST FOR ADDITIONAL INFORMATION LICENSE AMENDMENT REQUEST FOR FORT CALHOUN STATION TO ESTABLISH THE REACTOR PROTECTIVE SYSTEM ACTUATION CIRCUITS LIMITING CONDITION FOR OPERATION OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION, UNIT NO.1 DOCKET NO. 50-285 By letter d ated February 10, 201 2 (Agencywide Documents Access and Management Sy stem Accession No. ML12046A838), Omaha Public Po wer District (OPPD, the licen see) requested an amendment to Renewed Facility Operating License No. DPR-40 for Fort Calhoun Station, Unit 1 (FCS). The proposed amendment w ould establish the limiting condition for ope ration requirements for the reactor protective system (RPS) actu ation circuits in T echnical Specification (TS) 2.15, "Instrumentation and Control Systems."
The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the information provided b y the licensee and has determined that the fo llowing information is needed in order to complete its review.
NRC RAI #1:
The RPS contactors p rovide the means of powering the control element drive mechanism (CEDM) clutches or removing the clutch power. The contactors consist of an M co il, four main contacts (Normally Open), and one auxiliary contac t (Normally Close d). When the M coil is energized, its asso ciated main contacts are clo sed to conn ect the cir cuit breaker output to th e clutch power supplies. The same sequence occurs in t he three other trip paths, simultaneously energizing all clutch po wer supplies. The main contact s are open to interrupt th e power to t he clutch power supplies a nd the auxiliary contact is closed t o illuminate the trip light on the RPS clutch power distribution panel (AI-3) when the M coil is de-energized.
The RPS cl utch power supply trip contactor (AI-3-M1/M2/ M3/M4) is an Allen Bradley contactor, model No. 500-DOD940-91. This is a National Electrical Manufacturers Association (NEMA) size 3, 4-pole contactor with a coil rated for 120 volts (VAC) - 60 Hertz (Hz). The contactor is complete with one normally closed auxiliary contact (curr ent rating of 10A at 120 VAC). The contactor s were purchased as op en type contactors an d were inst alled in the existing NEMA T ype 1 contactor enclosures.
- 2. The current TS for FCS identifies the following operating modes:
a) Power Operation Condition (Operating Mode 1) b) Hot Standby Condition (Operating Mode 2) c) Hot Shutdown Condition (Operating Mode 3) d) Cold Shutdown Condition (Operating Mode 4) e) Refueling Shutdown Condition (Operating Mode 5)
To support the NRC staff review, please provide the following information:
U. S. Nuclear Regulatory Commission LIC-12-0136 Page 2 of 6 a) Please explain ho w the FCS Op erating Modes line u p with the Operating Modes in NURG-1432, "Standard Technical Specifications [STSs] Combustion Engineering Plants." In particular, please explain how FCS Operating Mode 2 c ompares to NUREG-1432, LCO 3.3.3, "Reactor Protective System (RPS) Logic and Trip Initiation (Analog)."
The attached diagram provides a detailed description of the corr espondence between operating modes defined in the STS and those defined in the FCS TS. (See Figure 1.)
While it is true that the FCS T S do not include an operating mode d escribed as Startup, (STS Operating Mode 2), the FCS Operating Mode 2 whi ch is identif ied as Hot Standb y Condition, is essentially the same as STS Operating Mode 2. Both STS and FCS Operating Mode 2 are defined as reactor critical at low po wer. In the STS, the reactor is crit ical at less than 5% rat ed thermal power (RTP) whereas the FCS TS Operating Mode 2 definition is reactor critical at less than 2% RTP.
In a comparison betwee n each of th e six STS-defined oper ating modes and the five FCS T S operating modes, it can be seen that each of the STS modes has a corre sponding FCS mode with one exception. That exception is STS Op erating Mode 4 which is defined as reactor subcritical and the RCS temperature betw een 200° and 350° Fahrenheit (F). FCS Operating Mode 3 (Hot Shutdown) is de fined as reactor subcritical and reactor coolant system (RCS) temperature greater tha n 515° F while FCS M ode 4 (Cold Shutdown) is defined as reactor subcritical and RCS te mperature less than 21 0° F. The FCS TS do not define a mode f or RCS temperatures between 210° F and 515° F whereas STS operating mode definitions comprehensively define the entire range of possible temperatures between cold shutdown and normal operating conditions.
The lack of total correlation between STS and FCS TS op erating mode definitions results in the need to describe the undefined mode in the Applicability section of the pr oposed TS 2.15.2. Since FCS TS has no operating mode corresponding to the STS Operatin g Mode 4, the gap is a ddressed by requiring a pplicability when RCS t emperature is greater than 210°F.
This requirement corresponds to ST S Operating Modes 3 a nd 4 and en sures that RPS Logic and Trip In itiation channels are o perable throughout the same range of plant operating conditions as those defined in STS.
b) License Amendment Request (LAR), Enclo sure 2, Section 3, "Applicability ," describes the applicability for when the RPS logic and trip initiation channels are required to be operable. Please identify the FCS operating modes when the prop osed LCO 2.15.2 is applicable.
As described above, ap plicability for the proposed TS 2.15.2 is intende d to cover t he same range of plant operating conditions as those required in STS. Therefore, for FCS, the modes of applicability would be Mode 1, Mode 2, Mode 3 and Mode 4 as well as RCS temperatures above 210° F which are not included in a specific mode definition. By requiring applicability above an RCS tempera ture of 210° F with mor e than one CEA capable of being withdrawn and RCS boron concentration le ss than R efueling Boron Concentration, Mode 3 is adequately addressed. Therefore, it is only necessary to specifically list Mode 1, Mode 2, and Mode 4 in the Applicability section.
U. S. Nuclear Regulatory Commission LIC-12-0136 Page 3 of 6 Figure 1 Comparison of Plant Mode Definitions as Defined in Standard Technical Specifications and Fort Calhoun Station Technical Specifications Standard Technical Specifications FCS Technical Specifications Mode 1: Power Operations Corresponds to Mode 1: Power Operations RX Power > 5% RTP RX Power > 2% RTP Mode 2: Startup Corresponds to Mode 2: Hot Standby RX Power < 5% RTP RX Power < 2% RTP Mode 3: Hot Standby Mode 3: Hot Shutdown Corresponds to (Reactor Subcritical) (Reactor Subcritical)
RCS Temp > 350o RCS Temp > 515o FCS Technical Mode 4: Hot Shutdown No Specifications do not (Reactor Subcritical)
Corresponding define a mode for RCS 350o > RCS Temp > 200o Mode at FCS temperatures above 210o and below 515o Mode 5: Cold Shutdown Mode 4: Cold Shutdown Corresponds to (Reactor Subcritical)
(Reactor Subcritical) o RCS Temp < 210o RCS Temp < 200 RCS Boron Concentration > Shutdown Boron Concentration but < Refueling Boron Concentration Mode 6: Refueling Mode 5: Refueling (Reactor Subcritical) Corresponds to (Reactor Subcritical)
RX Vessel Head Bolts not fully RCS at or > Refueling Boron tensioned Concentration Note: Refueling boron concentration corresponds to a shutdown margin of not less than 5% with all CEAs withdrawn.
U. S. Nuclear Regulatory Commission LIC-12-0136 Page 4 of 6
- 3. LCO 2.15.2 states that it is applicable fo r FCS Operating Modes 1 and 2; and when reactor coolant temperature (T cold) is greater than 210° F or FCS Operatin g Mode 4 with more than one control element assembly (CEA) rod capable of being withdrawn and RCS boron concentration less than REFUELING BORON CONCENTRATION.
a) Please clarify the ope rating modes applicable for LCO 2.15.2 using the definitions provided in FCS TS.
The operating modes a pplicable for TS 2.15.2 are Mode 1, Mode 2, Mode 3, and Mode 4 as well as the operating condition during which the RCS temperature is between 210° F and 515° F which is not included in any FCS-defined operating mode. As explained in 2b above, by specifying applicability above 210° F, the undefined condition is addressed as well as Mode 3.
Consequently, it is necessary to only call out Mode 1, Mo de 2, and Mode 4 as well as the RCS temperature range that is not specifically addressed in any FCS mode definition.
b) One of the applicable modes identified in the proposed LCO 2.15 .2 is "w hen reactor coolant temperature (Tcold) is greater than 210° F." Pl ease explain if this de scription corresponds to FCS operating Mode 3.
FCS operating Mode 3 (Hot Shutdown) is defin ed as reactor subcritical and RCS temperature above 515° F. Operating Mode 4 (Cold Shutdown) is defined as reactor subcritica l and RCS temperature below 210° F. The de scription when the reactor coolant temperature (T cold) is greater than 210° F does not correspond to any FCS-defined mode. Since there is no specific FCS mode def inition that addresses RCS temperature in the range of 210° F to 515° F, it is necessary to include an applicability statement to address t his possible operating condition.
c) NUREG-1432, STSs, identifies the applicability for LCO 3.3.3 to be STS operating Modes 1 and 2; or Modes 3, 4, and 5 with any reactor trip circuit breakers closed and any CEAs capable of being withdrawn. Please clarify how the operating modes for the proposed LCO 2.15.2 compare with this description.
STS-defined Mode 1 an d Mode 2 c orrespond to FCS defi nitions for Mode 1 an d Mode 2 ,
respectively. STS-defined Mode 3 and Mode 4 are addressed by the requirement, When the reactor coolant temperature (T cold) is greater than 210° F. STS-defined Mode 5 cor responds to FCS Mode 4. Consequently, the plant conditions required in the Applicability section for the proposed TS 2.15.2 correspond to the same range of plant operating conditions as defined in NUREG 1432.
U. S. Nuclear Regulatory Commission LIC-12-0136 Page 5 of 6
- 4. LAR, Enclosure, Section 2.0, states that the RPS man ual trip fun ctional unit included i n Table 2-2, "Instrument Operating Requirements for Reactor Protective Sy stem," will be removed, and that thi s function will be included in th e proposed LCO 2.1 5.2. The LAR Enclosure Section 3.0, Technical Evaluation, does not state that the manual function would be moved to the ne w LCO 2.15.2, instead this section s tates that requiring restoration of the inoperable channel is consistent with the current TS requirement contained in TS Table 2-2, item 1. Please clarify.
TS Table 2-2 is primarily focused on operability of channels or logic subsystems that make up the initiation portion of auto matic functions. That is, when the tables refe r to Minimum Operabl e Channels or Minimum Degree of Redundancy, they are referring to the inputs that are logically combined in order to develop an actuation log ic for the associated aut omatic function. In this context, it makes sense to define requirements for minimum operable channels or a minimu m degree of r edundancy that could be generically addressed by the p aragraphs of TS 2.15(1 )
through 2.15(4). These paragraphs list various possibilitie s of failed input channels and provide guidance as to the appropriate action. The paragraphs of TS 2.15(1), 2.15(2), 2.15(3), and 2.15(4) address initiat ing channels as opposed to actuatio n trains, which are addressed in th e new section 2.15.2. (See Reference 2)
Manual trip channels are functionally implemented at the actuation portion of the RPS circuitry as opposed to the initiating channel logic portion of the system. Therefore, the manual trip channel operability requirements are broken out from T able 2-2, which is i ntended to focu s on initiat ing channels, and are placed as a separate line item in Section 2.15.2.
The current TS requirement for man ual trip channel operability, as stated in Table 2-2, item 1, is that one of the two channels is required to be operable (i.e., the minimu m number of channels is specified as one). If o ne channel is inopera ble, TS 2.15(2) would apply since this se ction addresses a condition in which the minimum number of channels is reached. TS 2.15(2) requires that the inoperable cha nnel be placed in a tripped condition which would result in a shutdown o f the reactor. Since ma nual trip ch annels are not tested in modes 1 and 2 when the reactor is critical, appropriate corrective action would be taken whe n the plant is in a mo de where t he inoperability would be discovered. This is the approach r equired in the TS for t he Palisades Nuclear Plant (Palisades). Therefore, the proposed TS 2.15.2 would ali gn operability for manual trip channels with the Palisades TS. (The Palisades TS is referenced because the Palisades RPS design is similar to the FCS design in that contactors are used for trip actuation rather tha n reactor trip breakers.) The proposed change is consistent with the current TS req uirement in that it is unlikely that the cur rent required action would ever be applied due to the fact t hat testing is not performed when the reactor is critical.
- 5. Current TS requirement in LCO 2.15 (2) states that if a channel of a particular s ystem (Table 2-2) is inoperable, this channel must be placed in the tripped position. Further, if the inoperable channel has not been restored to operable status after 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the reac tor should be placed in a Hot Shutd own Condition (mode 3) w ithin the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> based on this information, it is not clear w hy the proposed LCO 2.15.2 (3) is requiring tha t with one inoperable RPS manual trip channel to restore the status to OPERABLE prior to entering Mode 2 from Mode 3. Please clarify.
U. S. Nuclear Regulatory Commission LIC-12-0136 Page 6 of 6 OPPDs Response to RAI #5:
The proposed change is based on t he fact that the manual trip channels are only tested when the reactor is n ot critical, (i.e., operatin g Modes 3, 4 or 5), which leads to the conclusion that a n inoperable manual trip channel would not be discovered while the plant is in operating Mode 1 or
- 2. Therefore, it is more appropriate to specify that correctiv e action be based on the mode of the plant when the inopera bility is di scovered (i.e., prohibit tra nsition from Mode 3 to Mode 2 until operability is restored). Further, the proposed change would align FC S TS to those currently in place for Palisades which has an RPS design as FCS.
Alternate Shutdown and Auxiliary Feedwater Panel, w ill be incorporated into a ne w LCO 2.15.3 with the list of c omponents being included into a new Table 2-6. Further, Section 3 states that no changes are prop osed for th e requirements, other than form atting. Please clarify the relationship between required actions 1 and 2 in the proposed TS 2.1.5.3.
The current TS 2.15(5) states that, when the number of operable channels falls below the required number of channels, either restore the required number of channels to OPERABLE status within seven (7) days, or be in hot shutdown (Mode 3) within the next twelve hours The proposed TS 2.15.3 required action is stated as follows:
(1) With the number of OPERABLE channels or control circuits less than the required number of channels, restore the required number of channels to OPERABLE within seven (7) days.
(2) With the required actions of (1) not met, be in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
RAI question #6 asks fo r a clarification of the fact that the current TS 2. 15(5) contains the logical conjunction or between the two re quired actions whereas the proposed TS 2.15.3 does not. In the proposed required action (2), the phrase With the required actions of (1) no t met takes the place of the or in the current TS. In other words, the proposed wording implies the logical connector without actually stating it. This is considered to be an improvement o ver the current wording in that the logical connect or or suggests that t he second required action could be performed without making an effort to restore inoperable channels to operability. The intent of the current TS is to allow seven days to restore required operability before requiring a plant shutdown.
Therefore, the proposed wording more precisely aligns with the intent of the current TS.
References:
- 1. Letter from OPPD (Jeffrey A. Reinhart) to NRC (Document Control Desk), NRC Inspection Report 05000285/2011007, Reply to a Noti ce of Violation (NOV); EA-11-0 25, dated August 17, 2011 (LIC-11-0084)
- 2. Letter from OPPD (R. P. Clemens) t o NRC (Document Cont rol Desk), Response to Request for Additional Information Concerning License Amendment Request (LAR) 09-01, Stea m Generator Blowdown Isolation Operability and Testing Requirem ents, dated June 30, 200 9 (LIC-09-0043)(ML091830041)